^ i-FS DOC. E 1.20: 0026-FS/V.2 - r :^ / (y .:l Volume 2 Appendices FINAL SUPPLEMENT ENVIRONMENTAL IMPACT STATEMENT Waste Isolation Pilot Plant Volume 2 of 13 January 1990 U.S. DEPARTMENT OF ENERGY Office of Environmental Restoration and Waste Management UNIVERSITY OF ILLliSiCiS LIBRARY URBANA-CHAMPAIQN BOOKSTACKS DOE/EIS-0026-FS FINAL SUPPLEMENT ENVIRONMENTAL IMPACT STATEMENT Waste Isolation Pilot Plant Volume 2 of 13 January 1990 U.S. DEPARTMENT OF ENERGY Office of Environmental Restoration and Waste Management Washington, D.C. 20585 ym^^: COVER SHEET RESPONSIBLE AGENCIES: Lead Agency: U.S. Department of Energy (DOE) Cooperating Agency: U.S. Department of the Interior, Bureau of Land Management (BLM) TITLE: Final Supplement, Environmental Impact Statement, (SEIS), Waste Isolation Pilot Plant (WIPP) CONTACT: For further information, contact: 1) Project Manager WIPP Supplemental Environmental Impact Statement Office P. O. Box 3090 Carlsbad, New Mexico 88220 (800) 274-0585 2) Carol Borgstrom, Director Office of NEPA Project Assistance Office of the Assistant Secretary for Environment, Safety and Health U.S. Department of Energy (EH-25) 1000 Independence Avenue, SW Washington, D.C. 20585 (202) 586-4600 3) William Dennison Acting Assistant General Counsel for Environment U.S. Department of Energy (GC-11) 1000 Independence Avenue, SW Washington, D.C. 20585 (202) 586-6947 ABSTRACT: In 1980, the DOE published the Final Environmental Impact Statement (FEIS) for the WIPP. This FEIS analyzed and compared the environmental impacts of various alternatives for demonstrating the safe disposal of transuranic (TRU) radioactive waste resulting from DOE national defense related activities. Based on the environmental analyses in the FEIS, the DOE published a Record of Decision in 1981 to proceed with the phased development of the WIPP in southeastern New Mexico as authorized by the Congress in Public Law 96-164. Ill Since publication of the FEIS, new geological and hydrological information has led to changes in the understanding of the hydrogeological characteristics of the WIPP site as they relate to the long-term performance of the underground waste repository. In addition, there have been changes in the information and assumptions used to analyze the environmental impacts in the FEIS. These changes include: 1) changes in the composition of the TRU waste inventory, 2) consideration of the hazardous chemical constituents in TRU waste, 3) modification and refinement of the system for the transportation of TRU waste to the WIPP, and 4) modification of the Test Phase. The purpose of this SEIS is to update the environmental record established in 1 980 by evaluating the environmental impacts associated with new information, new circumstances, and proposal modifications. This SEIS evaluates and compares the Proposed Action and two alternatives. The Proposed Action is to proceed with a phased approach to the development of the WIPP. Full operation of the WIPP would be preceded by a Test Phase of approximately 5 years during which time certain tests and operational demonstrations would be carried out. The elements of the Test Phase, tests and operations demonstration, continue to evolve. These elements are currently under evaluation by the DOE based on comments from independent groups such as the Blue Ribbon Panel, the National Academy of Sciences, the Environmental Evaluation Group, and the Advisory Committee on Nuclear Facility Safety. At this time, the Performance Assessment tests would be comprised of laboratory-scale, bin-scale, and alcove-scale tests. The DOE, in December 1989, issued a revised draft final Test Phase plan that focuses on the Performance Assessment tests to remove uncertainties regarding compliance with long-term disposal standards (40 CFR 1 91 Subpart B) and to provide confirming data that there would be no migration of hazardous constituents (details are available in Subsection 3.1.1.4 and Appendix O). The tests would be conducted to reduce uncertainties associated with the prediction of natural processes that might affect long-term performance of the underground waste repository. Results of these tests would be used to assess the ability of the WIPP to meet applicable Federal standards for the long-term protection of the public and the environment. The operational demonstrations would be conducted to show the ability of the TRU waste management system to certify, package, transport, and emplace TRU waste in the WIPP safely and efficiently. Waste requirements for the Integration Operations Demonstration remain uncertain. A separate document would be developed to describe in detail the Integration Operations Demonstration following the DOE's decision as to the scope and timing of the demonstration. During the Test Phase, National Environmental Policy Act (NEPA) requirements would be reviewed in light of the new information developed and appropriate documentation would be prepared. In addition, the DOE will issue another SEIS at the conclusion of the Test Phase and prior to a decision to proceed to the Disposal Phase. This SEIS will analyze in more detail the system-wide impacts of processing and handling at each of the generator/storage facilities and will consider the system-wide impacts of potential waste treatments. IV Upon completion of the Test Phase, the DOE would determine whether the WIPP would comply with U.S. Environmental Protection Agency (EPA) standards for the long-term disposal of TRU waste (i.e., 40 CFR Part 191, Subpart B; 40 CFR Part 268). The WIPP would enter the Disposal Phase if there was a favorable Record of Decision based on the new SEIS to be prepared prior to the Disposal Phase and if there was a determination of compliance with the EPA standards and other regulatory requirements. During this phase, defense TRU waste generated since 1970 would be shipped to and disposed of at the WIPP. After completion of waste emplacement, the surface facilities would be decommissioned, and the WIPP underground facilities would serve as a permanent TRU waste repository. The first alternative. No Action, is similar to the No Action Alternative discussed in the 1980 FEIS. Under this alternative, there would be no research and development facility to demonstrate the safe disposal of TRU waste, and TRU waste would continue to be stored. Storage of newly generated TRU mixed waste would be in conflict with the Resource Conservation and Recovery Act (RCRA) Land Disposal Restrictions; treatment would be required to avoid such conflict. The WIPP would be decommissioned as a waste disposal facility and potentially put to other uses. The second alternative to the Proposed Action is to conduct the bin-scale tests at a facility other than the WIPP and to delay emplacement of TRU waste in the WIPP underground until a determination has been made of compliance with the EPA standards for TRU waste disposal (i.e., 40 CFR Part 191, Subpart B). The bin-scale tests could be conducted outside the WIPP underground facilities in a specially designed, aboveground facility. The implications of this alternative include delays in both the operational demonstrations and alcove-scale tests, the lack of alcove-scale test data for the compliance demonstration, and placing the WIPP facilities in a "standby" mode. The specialized facility for aboveground bin- scale tests could be constructed at any one of the DOE facilities. In order to analyze the environmental impacts of this alternative in the final SEIS, the DOE has evaluated the Idaho National Engineering Laboratory in Idaho as a representative facility for the aboveground bin-scale tests. ADDITIONAL INFORMATION: The 1980 FEIS was reprinted and provided to the public with the draft SEIS which was published April 21, 1989. Public comments on the draft SEIS were accepted for a period of 90 days after publication. During that time, public hearings were conducted in Atlanta, Georgia; Pocatello, Idaho; Denver, Colorado; Pendleton, Oregon; Albuquerque, Santa Fe and Artesia, New Mexico; Odessa, Texas; and Ogden, Utah. This final SEIS for the WIPP project is a revision of the draft SEIS published in April 1989. It includes responses to the public comments received in writing and at the public hearings and revisions of the draft SEIS in response to the public comments. Revisions of importance have been identified in this final SEIS by vertical lines in the margins to highlight changes made in response to comments. Volumes 1 through 3 of the final SEIS contain the text, appendices, and the summary comments and responses, respectively. Volumes 6 through 1 3 of the final SEIS contain reproductions of all of the comments received on the draft SEIS, and Volumes 4 and 5 contain the indices to Volumes 6 through 13. An Executive Summary and/or Volumes 1 through 5 of the final SEIS have been distributed to those who received the draft SEIS or requested a copy of the final SEIS. Although not distributed to all who commented on the draft SEIS, Volumes 1 through 1 3 of the final SEIS have been placed In the reading rooms and libraries listed in Appendix K; these volumes will be mailed to the general public upon request. A notice of availability of the final SEIS has been published by the EPA in the Federal Register . The DOE will make a decision on implementation of the Proposed Action or the alternatives no earlier than 30 days after publication of the EPA notice of availability. The DOE's decision will be documented in a publicly available Record of Decision to be published in the Federal Register and distributed to all who receive this final SEIS. VI Foreword In October 1989, the Secretary of Energy issued a draft Decision Plan for the Waste Isolation Pilot Plant (WIPP). The Decision Plan listed all key technical milestones and institutional activities for which Departmental, Congressional, or State actions are required prior to receipt of waste for the proposed Test Phase, which is the next step in the phased development of the WIPP. The Plan was issued for review to States, Congressional representatives, other Federal agencies (including the Environmental Protection Agency and the Department of the Interior), and oversight groups (e.g., the Advisory Council for Nuclear Facility Safety, the Blue Ribbon Panel, the National Academy of Sciences, and the Environmental Evaluation Group). Revision 1 of the Plan was issued in December 1989. Departmental activities required prior to receipt of waste at the WIPP include completion of the "as-built" drawings for the facility, the Energy Systems Acquisition Advisory Board review process, waste-hoist repairs, preoperational appraisal and operational readiness review, mining and outfitting of the alcoves for the proposed Test Phase, and completion of this Supplement to the Environmental Impact Statement. Other Departmental activities include completion of the Final Safety Analysis Report (FSAR) and issuance of the FSAR addenda to address the proposed Test Phase and associated waste retrieval (if necessary). Future Departmental activities include the planned issuance of the EPA Standards Compliance Summary Report and the evaluation of waste form treatments and design modifications that may be required to meet the EPA Subpart B disposal standards. Key activities involving oversight groups include final development of an acceptable retrievability program to demonstrate that waste emplaced during the first five years of the facility operation are fully retrievable, and an integrated waste handling demonstration using simulated wastes to ensure system-wide readiness for receipt of wastes for the Test Phase. Institutional activities include concurrent pursuance of legislative and administrative land withdrawal (legislative withdrawal is the process preferred by the Department); the EPA's ruling on the DOE'S No-Migration Variance Petition in compliance with the Land Disposal Restrictions under the Resource Conservation and Recovery Act (RCRA); resolution of regulatory issues, including the State of New Mexico's authority to regulate mixed waste under the RCRA and the designation of routes to be used for transport of transuranic waste; Departmental resolution of any mineral lease at the WIPP; and completion of appropriate agreements with the Western Governors Association and Southern States Energy Board. This Supplemental Environmental Impact Statement (SEIS) is one of a number of milestones which are critical to the opening of the Waste Isolation Pilot Plant. This SEIS provides an upper bound of the potential impacts of the Proposed Action and alternatives. Based on this final SEIS, the Department will issue a Record of Decision no sooner than 30 days after the EPA publishes a notice of availability in the Federal Register. vii/viii TABLE OF CONTENTS VOLUME 2 COVER SHEET FOREWORD TABLE OF CONTENTS Appendix A WASTE ISOLATION PILOT PLANT WASTE ACCEPTANCE CRITERIA WASTE CHARACTERISTICS TRANSPORTATION EMERGENCY PLANNING TRANSPORTATION AND TRANSPORTATION-RELATED RISK ASSESSMENTS HYDRAULIC AND GEOTECHNICAL MEASUREMENTS AT THE WIPP HORIZON | B C D E F RADIOLOGICAL RELEASE AND DOSE MODELING FOR PERMANENT DISPOSAL OPERATIONS G TOXICITY PROFILES, RISK ASSESSMENT METHODOLOGY, AND MODELS FOR CHEMICAL HAZARDS H PUBLIC INFORMATION AND INTERGOVERNMENTAL AFFAIRS I METHODS AND DATA USED IN LONG-TERM CONSEQUENCE ANALYSES J BIBLIOGRAPHY K DOE READING ROOMS AND PUBLIC LIBRARIES L CONTAINERS AND CASKS FOR SHIPPING TRU WASTE M SUMMARY OF THE MANAGEMENT PLAN FOR THE TRUCKING CONTRACTOR N RE-EVALUATION OF RADIATION RISKS FROM WIPP OPERATIONS TEST PLAN SUMMARY P TRU WASTE RETRIEVAL, HANDLING, AND PROCESSING ix/x APPENDIX A WASTE ISOLATION PILOT PLANT WASTE ACCEPTANCE CRITERIA A-l/ii TABLE OF CONTENTS Section Page A.1 INTRODUCTION A-1 REFERENCES FOR APPENDIX A A-9 LIST OF TABLES Table Page A.1 .1 Summary of WIPP Waste Acceptance Criteria A-3 A-iii/iv A.1 INTRODUCTION The DOE has established Waste Acceptance Criteria (WAC) for the safe handling and long-term disposal of TRU radioactive waste at the WIPP (DOE, 1989). These criteria establish conditions governing the physical, radiological, and chemical composition of the waste to be emplaced in the WIPP, in addition to specifications for waste packaging to provide for the health and safety of workers and the public. Prior to any waste shipment departing any generator or storage facility, the shipment will be certified to meet the WAC. Similarly, the certification of shipments received at the WIPP will be verified prior to emplacement. The changes to the WAC since 1 980 are summarized in Subsection 2.3.1. The WAC were developed by a DOE-wide committee of experts on the handling and transportation of radioactive material. The basic concepts and limits chosen as WAC requirements are based on personnel safety, handling and storage restrictions at the WIPP facilities, methods of handling equipment, and procedures. Technical justification for the selection of the various requirements is provided in the WAC support documents 1 Revisions have been incorporated into the WAC as the WIPP project has evolved. These revisions have been reviewed and commented on by the storage/generator facilities, and others. The WAC is being modified as necessary to ensure compatibility with regulatory requirements such as the TRUPACT-II Certificate of Compliance issued by the Nuclear Regulatory Commission (NRC), the Resource Conservation and Recovery Act (RCRA), and the Department of Transportation (DOT) regulations. Modifications may also result from the Test Phase. The WAC were established with the assumption that the radiological hazards of TRU mixed waste containing hazardous materials listed in 40 CFR Part 261 , Subparts C and D, are much greater than any hazards from associated chemical constituents (Appendix B). Therefore, the WAC focus on the radiological properties of the waste, and the chemical criteria of the WAC are primarily for the prevention of immediate hazards such as fire and explosion. The labeling and data packaging criteria of the WAC also provide for the identification of hazardous waste. To ensure compliance with the WAC, the DOE has established the WIPP Waste Acceptance Criteria Certification Committee (WACCC) and requires that each facility certify that the WIPP-bound waste meets the WAC. Certification will be directed by the following documents as revised: DOE 5820.2A, "Radioactive Waste Management" ^ Vertical lines in the margins denote changes to the draft SEIS made in response to comments. A-1 WIPP-DOE-069, 'TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant" WIPP-DOE-114, 'TRU Waste Certification Compliance Requirements for Acceptance of Newly Generated Contact-Handled Wastes to be Shipped to the WIPP" WIPP-DOE-120, "Quality Assurance Requirements for Certification of the TRU Waste for Shipment to WIPP" WIPP-DOE-1 37, 'TRU Waste Certification Compliance for Acceptance of Contact-Handled Wastes Retrieved from Storage to be Shipped to the WIPP" WIPP-DOE-1 57, "Data Package Format for Certified Transuranic Waste for the Waste Isolation Pilot Plant (WIPP)" WIPP-DOE-1 58, 'TRU Waste Certification Compliance Requirements for Remote-Handled Wastes for Shipment to the WIPP" SOP 6.6, "Quality Assurance Audit Program" These documents may be reviewed in the DOE WIPP Project Office, Carlsbad, New Mexico and all DOE reading rooms. Each waste generating or storage facility will prepare a TRU Waste Certification Plan that describes the Site Certification Program and how that program meets the WAC and the requirements of the documents listed above. Each facility will also prepare a IRU Waste Qualitv Assurance Plan that describes their QA program designed to meet the requirements of WIPP-DOE-1 20. Both of these plans must be approved by the WACCC. Following the formal approval of Certification and Quality Assurance Plans for the waste generator or storage facility, a compliance verification audit will be performed by the WACCC. Subsequent periodic audits will be performed to verify that the facility is following the approved plans. Audit frequency will be determined by the Chairperson of the WACCC, in consideration of systematic requirements and facility certification status, but will generally be conducted on an annual basis at all facilities. The management of the generator or storage facility is expected to respond to findings and recommendations noted in the audit report, indicating the corrective action taken (or to be taken) to preclude recurrence. If subsequent facility audits determine that corrective action has not been satisfactorily implemented, the WACCC will decertify the waste so that it cannot be accepted at the WIPP. Since publication of the FEIS, the WAC have been modified twice, and these modifications are summarized in Subsection 2.3.1. A detailed discussion of the WAC and the basis for these criteria are provided in the TRU Waste Acceptance Criteria for the WIPP (DOE, 1989); a summary of the current WAC is provided in Table A.1.1. A-2 (0 'C 0) .t:! o 0) o c CO s. o o < Q. Q. $ ^^ CO E E 3 CO < 0) I- c (0 E 0) CC c o o E CO x: (0 a. Q. CO 0) < CD CO Its -Q o ® ?? CO g S.S *- .a a c = CO . ^ « £ 0) E (]) o Q) CO to Q. 5 E $ 0) 3 r ^ - 0) i T3 "CO i H 1 8 tj CO 2 o O 1 Q. ^< *- « m C x: >» o *- I- £ *" o «> ^ s > S Q. <«: EO® 8 0> CO c '^ E o ». o c o u -Q. >' « JC ^ o Q. CD ^ CO >,£: CO P H *' ® ■c tr 2 ♦- SO © « ^ CO O « CD ^ o = CO "D CD ^ 3 CO^ « c « ® o c o o> 3; -^ C -^ T- CO {H £ to .£ o ^ CO O CO c 2 C3) 9 n iz c CO 2 < xj E CO -^ c - o o c -^ If CO CO CD C CO o o CD CO $ D cr a> X cc z (O ^ ^ 5€ CO 5 CO 0) c _ CO $ o o (D 00 c ■" ^ c iSi! E O 0) jC c 'c5 o O sf S S 3 CC h- X o i (0 3 0) 0) 0) $« o O CD to CO c c CO X t o 0) (0 $ A-3 0) to T3 T3 C CO sz I e CC c 03 0) o E o c 5 CO sz v> VI CD O) CD O CO Q. 0) to CO 5 o CC 00 3 C3) nr h- 0) X Cfl c/> CC O CO ^ C\J c •g iS il CD a> -o ^ 0) CO o CO =11 c: O o _ ^ iE § CO CO O CO ~ 0) CO CO > D O r: Q) o n E — $ "CO SI CO « "O CO c .> CO D in Q> So, . en — CO _ C\J c ^ Q_ CO 1 > CO CO (T< ^ CO W CU CO Z £ < « to-i 2 2 3 Pl b: ^ -ti O ^ ^ LL CO O in 0) t^ CO CC I O CO CO o E CO CO 16 (0 o •a o ro t D CO (0 V sz C C "CO Q) J2 2 2 s Q) p H 5 " O Q) „ Z Z ^ ^ .CO -^ 9 £ w CO CO CO Q. O 0) CO s: (0 5« CD c "cO c o o CD to CO D CC X CC o E o °o C\i c 55 p 2 E 2^ to -^ 2 i C3) — O ® C CO c CO t.i |i CO ♦i 4- O CO O *- CO x: 2 c o o c CM o Q. C 0) CO "- SZ 0) *- 13 0) CO to -c Q CO CO ©■^ n o o CC 3 I- -Q -r -■°. CC > ^ S o £ ■d!I 2 2 o-o 3 CO g O" CO Q. O CD Q. w a5 w CO c i3 o Q) 2^ fe « "J Q. CO p « 0) ^ i_ CO i:: OOP CL ^ S g1°- T3 O 3 C O O < n CO I- to CO CC ■a 0) T3 c CO tj CO c o o CO SI CO CO s E CO o o> CO o CO Q. co" oS O o 8 m T— ^ C\J o m c Q. CO sz (0 o 5 E CC ^ H x: O) X 5 O 5 D O) CC .£ xl O o c o o u o o 3 C g CO 2 (0 ■c ?s SIS *= 3 to £ 2 2 O 0) 0) Oi CO CO CO CO *- ~ 0) 5 si c _ o CO o CO ^ d c (^ o CO CO D ^i5 E 3 E E C3) o »_ *^ O CO c o CO *- E E 3 3 ■o -o a. 3 CO •= CD CO CO uS 6 in CO O) Ob o o o o CO o X o oif^ o ^ o in in en ■D c CO in CO CM CO 00 . T- Q. A « r: I 5 09 . CO CO OT ct Z p 2 o) I CO .S .2 0) CO Q._ C C "CO 2 CO x: J3 ^ CO o g xi ■c 'x: = o $ « ^ CO «> CO CO CO « E 0) ill O ir C7 U. 3 0) CO ,d- 3 " UJ ^ -ft te "D O O ^ 2 S ro ^ i S »- -g *- |1 Z^ ^ 2 x: |1 E £ E 0) o •g 3 CO E 3 E o •— o S CM O J2 CM E c rti ro c ro IT CO 0, *- x: ro o ._ ^ to 2 0) CO CO to ■o IZ CO a CO t: o CL ^E CO sz CO CO c 'co O C CO B CO ro *i ro S > CO o c o o o CO x: CO 0) to k- CO o •o c c 'co "co c g o CD CO JiC CO Q. 4-^ CI) x: (0 O) CO ^^ CO o O k. CO o 3 CS ar to ;> CC 2? ^ CO << CM ^ CO il Q c ro -^ CO ^ ^ 1z = o- 3 CL m CO A-4 ® > 0) E T3 C (0 I 0) S E (0 0) cc © ■o c (0 I (0 o O c g o 0) cr I- X o M (0 0) E OJ (0 W § c o ■« c 0) 1 8 CO D o CO O M 2 « x: CC 0) « (0 I'll ® o c ♦- »_ _ ^ 5 P $ S' CO I jt "D "• (0 3) 5 Q-o E (D 0) 0) (0 ^ tr "o ® T- 0) E ^ Its o ■= Q-n c _, 111- I - c Q- O) g o £ c t3 2> c > (A O O Q-_ o "5 ro © E £ ros- ace Eo).2 ® ffl ^ &.y i Q.^ O (0 Q. O c o c o (0 O A-5 0) % D cr I- 1 ■o c CO SI I E 0) •o 0) D C *^ c o o < (Q H to CO ^ ID (T I- T3 ® C (0 jC « c o o 1 ^ © pi? 2. ® c S9 C3) c M C O i5 2LS -co $ B c g '16 o !E 0) § ^ 0) 5 S 3 1 2 DC CO ^E CO >» lu n tn 3 O 3 O ■q. (0 c »? P CO c o CD CO B -2 5 « E^ c\j .P to^ CO a a> -0) CO j5 CO % ^ . CO «) CD S E E^ i_ -^ -c CO ^ o .Q to .£ CD tii ■D <« to O t« CO O go .^Cto c I {« T=> o C 5 I- "J 1-^ Jo 2g o •a CO CD 0) if CD O Ql O 3 l1 CD CD ^ E o II c o .ti o c "53 13 a> Q. (S to A-6 a> 5i. 0) (0 £ sii E c (0 E el! f 0) 88 E (0 CO T3 ^ (0 (0 Q. < E E o 3 C 8 c C 3 0) o «J c II ® 13 E E 2 o ® 0) 55 P .12 o •^ 3 O) c CO o "^^ 3 (0 Q. CC (0 3 o- T3 C (Q « 5 ^ ^ > (0 0) S 0) (0 ^ w O _ ■^ > (D 6 U ® (0 5 O) Q. w, 2 3 CD d orOC x: O (0 •Si ^§ 15 o o > I- £ "o (0 0) ?^ $5 u aQ Z w a> £1 n $ ^ > o ■£ 5. ;= ^ 0) a> E E 3 C a Z H (0 •t E o Q. 0) s « CO o 0) E O < U CL O CC > I- .9>^ ?c3 pi w o c 8 §> (0 a) E o P -ft •^ O 0) - 2* k- 0) (0 ® C -i^ Z d. c o •c •E o o z ^^ w > to "5 3 O <5 « E. ® (0 = 2E O CD 0) 8 ^ « ^ S.E o ^^ 2.1 « o i§E ® £ w c ^ o O ^- .£ a o 8|^ w o c «^ .£2 E?.« 3 ^ .£ (Q 2 CO . c d CO E " il •fc £ c'2 0"° . C C Sl ^ CO O) o ■5 c o « c CO ® (S § E CO 5 J S w I 'co o o £ 8 c o o 0) c Q I O 1 'S o E E A-7 0) 0) V) o T3 a> o (0 •c CO (S i (1) to cr \- ■a ■o c CO o E tr ■D 0) T3 D O c o o < 0) CT5 0) ♦-» CO CO tr •o ■D C CO x: I CO ^^ c o o c o .ti o to CO D I o (0 CO o E CO CD W X3 CO CO E n ^ .2 CO ■o o 2 E o c ^ CO o ^ *- CD C 0) ^ E > ^£ O it-'^ E y 2 So CD ^ O O Q. *' 0) 2 T3 Q. >, CD Q. ^7 Q) O CO tl *- CO Q-cD E ■a CD CO k. (D a CO ■•5 CO CD C CD O) CD jO S Q ),- CD O c O t c CD = (D E /I) CO H- CO CO CD CO CD ^^ c CO CO » CO n CD CD C CO CO (D ■D CO CO CO (D CO CO £ $ O O) GC I- I O E (0 CO CO CO CO $ CO O ■D k— CO CO .9 CO .t= •^ CD T3 ■D -0 CD CO CD C T3 15 ii CO C CO 0) .9 E c = CO CO w ^ c w 2 .55 (D to E ^S^^ sz gco^ iS .g CO c o . o -o (D CC CO CD c CD t« CO CO CO CO 8 CO CC O £ o CD •9 <" CO ^ O CO '-' CO o CD CO II (D CO °l c E C *: CO 0) E = c CO .5^ CO •- -D ^ x: CD *^ ^^ •i ^ CO o si "s- ^ CO cr 2 o 8 "^ c E o o c o t; CO Q. CC u. o o ■D CD •c o a 2 d V T3 13 C CO 2o W CO 53 Q. E CO CO 0) CO ■D O ■D CD .>< C S CO T3 - CD CD CO CO > osive Ipres: CO o CD Expl Com ■•o CO CC CO CO c CO o to to i D CC c O E CO CO O • O U- x X T3 s C 5 . Q. Q. Z t?) < *. CO C o § B 3 ■o CO CC i co^- k. X o CO SZ CC 4-^ ■D sz o 4-^ .CO i S o li 03 03 c h— o S.2 o o o i^^ O *" O O UJ eS Q^ CO o CO S c 'k— O CD rj ■ss o O CD iz *-» 3 Q. c CO c in CO > (S2 cr §4 LU ^ E E ^ _3 Eo c o o T- *i »_ D ^ )r Q. < ^ CD O ^ .c »_ 1- h- o A-8 REFERENCES FOR APPENDIX A DOE (U.S. Department of Energy), 1989. 'TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant," WIPP-DOE-069, Rev. 3, U.S. Department of Energy, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1987. The DOE Evaluation Document for DOT Type 7A. Type A Packaging . MIL 3245/DOE/DP 0053-HI, Mound Applied Technology, Dayton, Ohio. A-9/10 APPENDIX B WASTE CHARACTERISTICS B-i/il •'<.•.»■•• •>y.. TABLE OF CONTENTS Section Page B.1 INTRODUCTION B-1 B.2 RADIONUCLIDE INVENTORY OF TRU WASTE B-4 B.2.1 Waste Acceptance Criteria B-4 B.2.2 Waste Volumes B-4 B.2.3 Radionuclide Characteristics B-5 B.2.3.1 General Radiation and Radioactivity Characteristics B-5 B.2.3.2 High-Curie Waste B-10 B.2.3.3 Neutron-Emitting Waste B-10 B.2.4 Calculation of Source Terms for Various Release Scenarios ... B-1 2 B.2. 4.1 Source Terms for Transportation Analyses B-1 2 B.2.4.2 Source Term for WIPP Operational Analysis B-1 9 B.2. 4. 3 Source Term for Long-Term Performance Analyses B-1 9 B.3 HAZARDOUS CHEMICAL CONSTITUENTS B-25 B.3.1 CH TRU Mixed Waste B-25 B.3.2 RH TRU Mixed Waste B-27 REFERENCES FOR APPENDIX B B-31 B-iii .'■/.• ■ LIST OF TABLES Table B.2.1 B.2.2 B.2.3 B.2.4 B.2.5 B.2.6 B.2.7 B.2.8 B.2.9 B.2.10 B.2.1 1 B.2.12 B.2.13 B.2.14 B.3.1 B.3.2 Page Currently projected total radionuclide inventories by facility for CH and RH TRU waste B-6 Estimated volumes of CH TRU waste in retrievable storage or projected to be generated through the year 201 3 B-7 Estimated volumes of RH TRU waste in retrievable storage or projected to be generated through the year 2013 B-8 Volumes of stored and newly generated TRU waste, scaled up to equal the design capacity of WIPP b-9 Summary of average TRU waste characteristics B-1 1 Average radioactivity in a shipment of CH TRU waste B-1 4 Average radioactivity in a shipment of RH TRU waste B-1 5 Quantities used in estimating the average radioactivity in a shipment of CH TRU waste from Rocky Flats Plant B-1 6 Quantities used in estimating the average radioactivity in a shipment of RH TRU waste from Los Alamos National Laboratory B-1 7 Mass and radioactivity of the radionuclides in an average drum of CH TRU waste B-20 Mass and radioactivity of the radionuclides in an average standard waste box of CH TRU waste B-21 Radioactivity of the radionuclides in an average canister of RH TRU waste B-22 Initial radionuclide inventory in CH TRU waste for the assessment of long-term performance B-23 Modified radionuclide inventory in CH TRU waste for the assessment of long-term performance B-24 Estimated quantities of TRU mixed waste (by waste form) from Rocky Flats Plant B-28 Estimated maximum concentrations of hazardous chemical constituents in CH TRU mixed waste from the Rocky Flats Plant .... B-29 B-iv B1 INTRODUCTION This appendix provides information on the characteristics and quantities of the TRU waste that may be emplaced at the WIPP. This information is necessary for assessing the potential impacts of transportation and WIPP operations, as well as the performance of the WIPP over the long term. Current information and assumptions regarding TRU waste have changed substantially since the WIPP FEIS (DOE, 1980) was published. As explained below, these changes have resulted from changes in the definition of TRU waste, changes in the sources of the waste (i.e., the DOE facilities at which TRU waste is generated or stored), the elimination of experiments with defense high-level waste from the plans for the WIPP, the addition of high-curie radioactive waste and neutron-emitting waste, the decision to evaluate the potential impacts of the hazardous chemicals that are contained in the TRU waste, and an extensive effort to accurately characterize the waste at each of the generator or storage facilities. The characterization effort has provided information about the radionuclide inventory (i.e., the radioactivity, the mass, and the longevity [the half-life] of radionuclides in the waste) and the hazardous chemicals that are present in the waste. Between 1970, when the category of TRU waste was established, and 1982, TRU waste was defined as waste containing long-lived alpha-emitting radionuclides at a concentration greater than 10 nCi (i.e., 10 one-billionths of a Ci) per g of waste. In 1982, the DOE, having evaluated the potential hazards of TRU waste, decided to change its definition. This new definition was accepted by the EPA (1982) and TRU waste is now defined as waste containing alpha-emitting transuranic radionuclides that have half-lives of 20 years or more and that occur in concentrations exceeding 100 nCi per g of waste. ("Transuranic" in this case means uranium and several radionuclides that are heavier than uranium.) As a result, some waste formerly classified as TRU waste is now classified as low-level radioactive waste, and therefore it is not eligible for disposal in the WIPP. In general, as a result of this change, the average radioactivity of TRU waste has increased. As in the FEIS, a distinction is made between TRU waste known as contact-handled (OH) waste and TRU waste known as remote-handled (RH) TRU waste (DOE, 1989a). For the CH TRU waste, the radiation-dose rate at the external surface of a waste container (drum or box) must be below 200 mrem (200 one-thousandths of a rem) per hour. This waste can be handled directly by personnel without excessive radiation exposure. The RH TRU waste has surface-radiation-dose rates between 0.2 and 1 ,000 rem per hour, but only 5 percent of this waste can exceed 100 rem per hour. In general, the FEIS analyses were based on waste from only two sources: the Idaho National Engineering Laboratory, which was expected to send both CH and RH stored TRU waste, and the Rocky Flats Plant in Colorado, which was expected to send newly generated CH TRU waste. The DOE now expects that post-1970 TRU waste would B-1 eventually come from 10 generator and/or storage facilities as discussed in Subsection 3.1.1. Thus, in order to establish the upper limit for the potential impacts, the analyses in this SEIS, like those in the draft Final Safety Analysis Report (FSAR-DOE, 1989b), are based on waste from 1 facilities, and 5 of these facilities have both CH and RH TRU waste. The consideration of 10 facilities significantly affected assumptions about the contents of average containers of TRU waste, which vary from facility to facility (see Tables B.2.5, B.2.10, B.2.11, and B.2.12). For example, a facility not previously considered, the Savannah River Site, will contribute 92 percent of the plutonium-238 that may be disposed of at the WIPP, and plutonium-238 accounts for nearly half (46 percent) of the total radioactivity of the CH TRU waste that may be emplaced at the WIPP. Similarly, the combined waste from three of the new facilities-Savannah River Site, Los Alamos National Laboratory, and Hanford Reservation-account for 73 percent of the plutonium- 241. Although waste may be received from more facilities, the change in the definition of TRU waste has decreased estimates of waste volumes. The WIPP was designed to receive 6.2 million cubic ft of CH TRU waste and 250,000 cubic ft of RH TRU waste, and the analyses in the FEIS (DOE, 1980) were based on those volumes. However, the DOE'S Integrated Data Base, which contains information on the various types of radioactive waste in the United States and is revised annually, shows a decreasing trend. In 1987, the Integrated Data Base (DOE, 1987) reported 5.6 million cubic ft as the estimate for CH TRU waste, both retrievably stored and to be produced from 1987 through 2013 ("newly generated"), whereas the 1988 edition (DOE, 1988) reported a volume of 4.8 million cubic ft, and the 1989 document (DOE, 1989d) estimated a total volume of 4.2 million cubic ft. To provide conservative (i.e., pessimistic) upper limits for the estimated potential impacts of the WIPP, the DOE decided to base the SEIS analyses on the design capacity of the WIPP. Therefore, for the purposes of this SEIS, the volumes given for each generator or storage facility in the 1987 integrated Data Base were proportionately scaled up to the total design capacity of the WIPP. Since the publication of the FEIS in 1980, the DOE has attempted to better define the characteristics of the waste. These efforts have included improved sampling of the waste, examination by x-raying, assays of the radioactive-material content, and implementation of improved methods for tracking and recordkeeping. In the FEIS, the information on the RH TRU waste was based on the data available for defense high- level waste, which contains significant amounts of short-lived fission products and therefore has more radioactivity than does the RH TRU waste. The information in the SEIS is based on data collected specifically for RH TRU waste. The rest of this appendix is divided into two parts: Section B.2, which discusses the radionuclide inventory of the TRU waste, and Section B.3, which covers the hazardous chemical constituents of the TRU waste. The section on the radionuclide inventory includes information on waste volumes and the radioactivities, half-lives, and masses of the radionuclides in the waste. In addition, it explains the procedure used in calculating the following quantities used in various impact analyses: the average radioactivity per shipment of waste, which was used in the analyses of transportation impacts; the average radioactivity per container of waste, which was used in analyzing B-2 the safety of WIPP operations; and the radionuclide inventory for the assessment of long-term performance. Section B.2 also discusses two types of TRU waste that were not considered in the FEIS analyses: high-curie and neutron-emitting waste. Section B.3. discusses the hazardous chemical constituents in both CH and RH TRU waste. The comments on the draft SEIS and continued discussions with personnel at the various waste generating and storage facilities led to the following revisions in this appendix: • This introduction was rewritten to explain why there are differences in the radionuclide inventory of the FEIS and this final SEIS. • Tables B.2.2 and B.2.3 were revised to use the correct number of significant digits for waste volumes and to reflect minor redistribution of volume projections for Argonne National Laboratory-East for RH TRU waste. • The waste volumes in Table B.2.4 were scaled up for all waste facilities in proportion to the volume given for each facility in the 1987 Integrated Data Base (DOE, 1987). • The text in Subsection B.2.4. 1 was modified to more clearly explain how the values given in Table B.2.6 for the radioactivity per waste shipment were calculated. The values were corrected to account for the misapplication of various data. • Tables B.2.8 and B.2.9 were rearranged to more clearly demonstrate the calculations made to determine the radioactivity per waste shipment. • The discussion of the transport index in Subsection B.2.4. 1 was revised to more clearly explain the source of the radiation that determines the transport index. • The assumption that the drums of CH TRU waste are filled to 80 percent of their capacity was eliminated because the calculations based on this assumption greatly overestimated the volume of waste to be emplaced in the WIPP. • Tables B.2.13 and B.2.14, which show the radionuclide inventory used in assessing the long-term performance of the WIPP, were revised by increasing the inventory to represent a volume equal to the design capacity of the WIPP. In addition, the radionuclides in the latter inventory were assumed to have undergone radioactive decay for 100 years to account for the period of institutional control. • The text on high-activity waste. Subsection B.2.3. 2, was modified to more clearly discuss the radioactivity of plutonium-238. B-3 B.2 RADIONUCLIDE INVENTORY OF TRU WASTE This section discusses the radionuclide inventory of TRU waste and explains how the initial amounts of material needed for assessing environmental impacts were calculated. These quantities serve as the basis for the estimation of the amounts of radioactive material that would be released in a given situation, such as transportation, operation under normal conditions, various accident scenarios that may occur during operations, or unintentional human intrusion after the WIPP has been permanently closed. B.2.1 WASTE ACCEPTANCE CRITERIA All waste must be certified to meet the WIPP Waste Acceptance Criteria (DOE, 1 989a) before it is transported to the WIPP. The Waste Acceptance Criteria have been refined to reflect the requirements of regulations issued by the U.S. Nuclear Regulatory Commission (NRC) and the Department of Transportation for the transportation of waste and to enhance the safety of long-term isolation. The original criteria were described in Chapter 5 of the FEIS (DOE, 1980); the current criteria are summarized in Subsection 2.3.1 and Appendix A, Table A.1.1. The Waste Acceptance Criteria that are relevant to the radionuclide source term include the following: • The surface contamination on containers of CH or RH TRU waste may not exceed 50 percent of the limits specified in Department of Transportation regulations in 49 CFR 173.442. • The thermal power (the heat-generating capacity) of a package of CH TRU waste must be labeled if it exceeds 0.1 W per cubic ft. The thermal power of RH TRU waste may not exceed 300 W per canister. In addition, the total plutonium-equivalent curies (PE-Ci) are limited to 1,000 per container. (The PE-Ci concept is discussed in Appendix F). In order to ensure that nuclear criticality (i.e., a self-sustaining nuclear chain reaction) will not occur, the total quantity of fissile material is limited to 200 g per drum. Fissile-material concentrations in boxes (e.g., the standard waste box that may be shipped to the WIPP~see Appendix D) are restricted to a maximum of 5 g per cubic ft, up to a maximum of 350 g per box. B.2.2 WASTE VOLUMES The WIPP was designed to receive about 6.2 million cubic ft of CH TRU waste and about 250,000 cubic ft of RH TRU waste, or a total of about 6.45 million cubic ft. These quantities were used in designing the WIPP and in estimating radionuclide inventories for the analyses in the FEIS (DOE, 1980). However, as explained in the B-4 introduction to this appendix, tiie estimated volumes of waste that may be sent to the WIPP have decreased over the years. When the preparations for the SEIS analyses began, the recent information available on waste volumes was the information given in the 1 987 edition of the DOE's Integrated Data Base (DOE, 1987), which is revised annually. This data base showed that the volumes of TRU waste that had been stored since 1970 or were projected to be generated between 1 987 and the year 201 3 were lower than those estimated for the design of the WIPP: the 1987 estimates were 5.6 million cubic ft for the CH TRU waste and about 95,000 cubic ft for the RH TRU waste, or a total of about 5.7 million cubic ft. The radionuclide inventory for these waste volumes is shown in Table B.2.1, and the waste volumes reported in the 1987 Integrated Data Base are given for each generator or storage facility in Tables B.2.2 and B.2.3 for CH and RH TRU waste, respectively. The data-base reports issued since 1987 continue to show a decrease in waste volumes. The 1988 Integrated Data Base (DOE, 1988) and the report for 1989 (DOE, 1989d) cite 4.8 and 4.5 million cubic ft, respectively, for the total volume of the TRU waste. However, in order to establish conservative (i.e., pessimistic) upper limits for the potential impacts of the WIPP, the DOE decided to base the analyses in this SEIS on the maximum assumed volume of 6.45 million cubic ft of TRU waste. This was done by scaling up, for each waste generating or storage facility, the volume given in the 1 987 data base for CH and RH TRU waste to correspond with the design capacity of the WIPP, with the scaling up being in proportion to the volumes reported in 1 987. For CH TRU waste, the 1987 volume was multiplied by 10.7 percent. The scaling-up factor (10.7 percent) was determined by subtracting the volume in the 1987 data base report from the design capacity of the WIPP and dividing this difference by the volume in the 1 987 data base report. For RH TRU waste, the volume at each waste facility that may ship RH TRU waste to the WIPP was increased by 1 63 percent. The scaled-up volumes for each facility are given in Table B.2.4. B.2.3 RADIONUCLIDE CHARACTERISTICS B.2.3.1 General Radiation and Radioactivitv Characteristics In addition to waste volumes, the SEIS analyses of potential impacts from waste transportation and WiPP operations and the assessment of long-term performance required information on the radionuclide composition of the TRU waste (radionuclides and weight fractions) and radioactivity (i.e., number of curies from plutonium and other alpha-emitting TRU radionuclides). These data were obtained from the 1987 Integrated Data Base (DOE, 1 987) and additional information that was obtained from each of the waste facilities on fission-product fractions, the total quantities of radionuclides (in curies), and the numbers of actual waste containers in storage and projected through the year 2013. This additional information has been published as a report that documents the waste-characterization data base for the WIPP (DOE, 1989c). Together with the 1 987 data base, this report constitutes the basis for the radiological analyses reported in this SEIS and in the WIPP draft FSAR (DOE, 1989b). The 1987 Integrated Data Base (DOE, 1 987) was consistently used to establish the volume of waste from B-5 TABLE B.2.1 Currently projected total radionuclide inventories by facility for CH and RH TRU waste Radionuclide inventory (curies)® Retrievably Newly stored generated Waste facility'' waste*' waste^ Total CH TRU waste Idaho National Engineering X 10^ Laboratory 3.74 X 10^ 7.61 X 10^ 3.75 Rocky Flats Plant® 1.05 X 10^ 1.05 X 10® Hanford Reservation 6.85 X 10^ 1.10 X 10^ 1.78 X 10® Savannah River Site 8.59 X 10^ 3.70 X 10® 4.56 X 10® Los Alamos National Laboratory 5.96 X 10^ 1.61 X 10® 2.21 X 10® XIO" X 10^ Oak Ridge National Laboratory 2.80 X 10" 3.51 X 10" 6.31 Nevada Test Site^ 4.73 X 10^ 4.73 Argonne National Laboratory-East® 7.13 X 10^ 7.13 X 10^ Lawrence Livermore National Laboratory® 8.45 X 10" 8.45 X 10" X 10^ Mound Laboratory® 1.87 X 10^ 1.87 Subtotal 2.54 X 10^ 7.58 X 10® 1.01 X 10^ RH TRU waste Idaho National Engineering X 10" X 10" Laboratory 1.51 X 10^ 2.28 X 10" 2.43 Hanford Reservation 4.04 X 10^ 1.93 X 10" 2.33 Los Alamos National Laboratory 3.64 X 10^ 2.42 X 10^ X 10^ X 10^ 3.88 X 10^ •3 Oak Ridge National Laboratory 2.71 X 10^ 1.84 2.89 X 10^ _-5 Argonne National Laboratory-East 1.03 1.03 X 10^ Subtotal 1.19 X 10" 4.36 X 10" 5.54 X 10" GRAND TOTAL 2.58 X 10^ 7.62 X 10® 1.02 X 10^ Radionuclide inventories for the waste volumes estimated in the 1987 Integrated Data Base (DOE, 1987)-that is, 5.6 million ft^ of CH TRU waste and 95,000 ft^ of RH TRU waste. Unless indicated othenwise, these facilities both generate TRU waste and are designated as a TRU waste storage facilities. Stored as of December 31 , 1 986. Generated between 1987 and 2013. Facility that generates but does not store TRU waste. Facility that does not generate TRU waste, but is designated a TRU waste storage facility. B-6 TABLE B.2.2 Estimated volumes of CH TRU waste in retrievable storage or projected to be generated through the year 201 3 Estimated volume (ft^)^ Waste facility'^ Retrievably stored waste'^ Newly generated waste*^ Total Idaho National Engineering Laboratory 1,073,710 9,920 1 ,083,630 Rocky Flats Plant® 2,037,600 2,037,600 Hanford Reservation 293,250 537,800 831 ,050 Savannah River Site 91 ,465 615,700 707,165 Los Alamos National Laboratory 250,910 302,300 553,210 Oak Ridge National Laboratory 19,160 42,000 61,160 Nevada Test Site^ 21 ,290 21 ,290 Argonne National Laboratory-East® 3,800 3,800 Lawrence Livermore National Laboratory® 259,400 259,400 Mound Laboratory® 40,100 40,100 TOTAL 1 ,749,785 3,848,620 5,598,405 ^ Estimated volumes correspond to the Integrated Data Base for 1987 (DOE, 1987). The volumes of waste used for the environmental analyses in this SEIS are higher and are based on the design capacity of the WIPP. Unless otherwise indicated, these facilities both generate TRU waste and are designated TRU waste storage facilities. ^ Stored as of December 31, 1986. From Table 3.5 in the Integrated Data Base for 1987 (DOE, 1987). Generated from 1987 through 2013. From Table 3.16 in the Integrated Data Base for 1987 (DOE, 1987). ® Facility that generates but does not store CH TRU waste (except limited quantities pursuant to RCRA regulations). Facility that does not generate TRU waste, but is a designated TRU waste storage facility. a B-7 TABLE B.2.3 Estimated volumes of RH TRU waste in retrievable storage or projected to be generated through the year 201 3 Estimated volume (ft^)^ Waste facility^ Retrievably stored waste^ Newly generated waste*^ Total Idaho National Engineering Laboratory 985 4,820 5,805 Hanford Reservation 848 28,600 29,448 Los Alamos National Laboratory 1,020 191 1,211 Oak Ridge National Laboratory 45,478 9,540 55,018 Argonne National Laboratory-East® 3,500 3,500 TOTAL 48,331 46,651 94,982 ^Estimated volumes correspond to the Integrated Data Base for 1987 (DOE, 1987). The volumes of waste used for the environmental analyses in this SEIS are higher and are based on the design capacity of the WIPP. j ^Unless otherwise indicated, these facilities both generate RH TRU waste and are ] designated TRU waste storage facilities. I ^Stored as of December 31, 1986. From Table 3.5 in the Integrated Data Base for 1987 (DOE, 1987). [ ^Generated from 1987 through 2013. From Table 3.16 in the Integrated Data Base for 1987 (DOE, 1987). Facility that generates but does not store RH TRU waste. B-8 mm w/. ■g «0 ^ 3 (0 ~ UJ to CO o (0 a (0 o c « T3 to (0 D T3 ■5 «> C3) ■o c a 0) to o E _3 O > cvi CD UJ _) m < if I S O CO CO Q E o o E LU (0 o o 0> <0 Z i 5 T3 a> i: v> o a 35 5 xxxxxxxxxx »-CMO>r«.(0(OCsl'*»-'.-55ino i-^cviodNiiritDcvicocvJTt ^ o ____J)00___ M a S X u xxxxxxxxxx c\J5oo«>ojpQoa)T- otDrti-ocQoooioo o) c\j 10 c\i»-'c\iddc) I o CD o UJ - — ^ « QL o z If _i « |z':|^ ir o "S ■^ o i8 -Q w (H UJ _J I _ £> O CO -s o « JQ Z roi- T3 o c O 0) cc I c rf OC Tl § 8 I S CO -J O z ixi _ O £« « E o 2 * S '■S ■-, 2 ^ CO « o — ' C C 73 C X M a 3 CC X X X X X CM «0 ■* Q »- 00 CO in in a> "S^ cvi o) CO >-' O I o Si (8 O) _c o o D) o UJ "S -5 ^ CD m o « O O -s ^ « (0 5 I a >>UJ £' 205 o 2-P CO o _I Si — <^ « <8 -I c o «-2 ■^0-5 oz 8 ^^ i C JO < ■5 cc O < _J X CM '^^o^o'^'^ ■^ ^^ ^- T- T- *— X X X X X X m g h- to N. ^ CO O) f^ Oi 10 (3> tT 00 10 »- »- '^^o^o'^'^ ^0 T— ^ ^ *— T— T- X X X X X X -a- Q '- 00 cj> in in CM c» CM in CM in CO T- cn ^0 "W^^cTb ^0 ■^ *— T— T— T- T— ^" X X X X X X X in in 00 m Q CM 00 t in CO CO ■^ cj) 00 -s- T- 'T o I- CO 5 2 cc 1^ tr CO •^ CO CO =3 OC \- ■o O 0) CO ' m" !S .S 2 l« - £ o CO w 0) 5 to w g- E '" O c OC $ «►- S (S i > (S o 1 o o> S (O cvi CD lU _l CD < ^ .^ tr O I c 3 o z d z 3 LU z z < X 111 —I z < "o 3 C O « cc o o o <3,' fT><3,T "^ 8 8 8 8 d d d d X S CVI ■r^ c\i X X X ° 5 5 - 00 «> f^ a> _ o 5 P T^ O o o d -^ O0000'o000oo°° X X X X X 8 8 8 8 8 d d d d d .- f^ XXX CM lO O) 00 I- CM ^ 'T »-^ 8 d Q^jQQ*-l»—»^-'*— '«—'*—'*— 'V-'V^ X '^ X (O g lO CM "*? •- ^r " - X X X X X 8 8 8 8 8 d d d d d in t r^ CM in ^ 'o o o o o o o 'q o o o o o X X X X X X ■^ 5? S f2 * 2 '° 2 2 ° ° ° ° ° ° " " " XxXX^XXXXyXXX 8g[;558;q8SS8^88 o<^o6nO'-®^'°'r^°° 'o X NO I o X CM N. ai X X X X X X 8 S § 5 S 8 d »- lo 1- ^ d X X CJ) o 00 o CO d o o "o 'o o X X X X X X 8 8 8 8 8 S d d d d d c6 O O O CJ o o o o o X o X 8 S S § X X X X X X 8 8 8 8 }R ^ o o o o 0) m ^ ?? ^ s 1 ?5 CM O <0 a CO a\ b 5 to »- o '*• o o CM ^ o 1- ? CM 1- S ^ CM <^ CM CM TT E 3 E E E E b 3 ^ -) 3 3 3 E c r C C C O o o o o o ■s S B B E 3 m a. Q. 0. Q. < o O X CM CD 0) -^ X in in in < o X V 4) CJ) m oq n Q. CO ^ o , ^" o X 3^ a o 03 Q- O O) CD r CO 4) CM r r 0) 4) ^^ b V) n U) ji >. (0 ffl r 4; CO u. en C/) 4> 111 c CO CO o c £ bH o (- ■D 0) < h Q. 3 -) cc CO 1- 4) V 4) w r ? ~' ro ■5 E CO rr 4) F CO 3 c 4) 4) E ^ a ^ £ c Q. c 4) ^ 3 E n 4) CO e _l 2 r 3 en a > SI (0 c .0 15 to in \- Z ^ 4> 4> <^ 3 ■^ C c _l u C\J CO c CD CJ> c 4) < ■55 Z ^ <0 1- uJ > _i z CO C < E E T5 3 >. 4> (0 0) > DC •^ _1 B-14 TABLE B.2.7 Average radioactivity in a shipment of RH TRU waste^ Waste facility' Radionuclide ANLE HANF Cobalt-60 Strontium-90 Ruthenium-106 Antimony-125 Cesium-137 Cerium-144 Europium-155 Thorium-232 Uranium-233 Uranium-234 Uranium-235 Uranlum-238 Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Plutonium-242 Americium-241 Curium-244 Californium-252 TOTAL 0.00 X lO"" 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.83 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 1.21 X 10-^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 2.52 X lO'"" 9.27 X 10'^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 2.97 X 10" 6.76 X 10° 1.89 X 10"^ 0.00 X 10° 9.46 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 5.41 X lO'"* 8.11 X 10"^ 2.43 X 10"^ 5.41 X 10"^ 0.00 X 10° 9.73 X 10'^ 1.38 X 10° 4.05 X lO""" 8.11 X 10° 8.65 X 10'^ 5.95 X 10"'' 0.00 X 10° 0.00 X 10° INEL 0.00 X 10° 4.08 X 10° 0.00 X 10° 0.00 X 10° 5.81 X 10° 0.00 X 10° 0.00 X 1 0° 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.68 X 10"2 2.46 X 10'^ 0.00 X 10° 1.63 X 10-2 8.80 X 10^ 3.58 X 10^ 0.00 X 10° 0.00 X 10° 3.27 X 1 0'^ 0.00 X 1 0° 0.00 X 10° LANL 0.00 X 10° 7.99 X 10° 6.31 X 10° 1.95 X 10"^ 6.18 X 10° 6.22 X 10^ 3.13 X 10''' 0.00 X 10° 0.00 X 10° 0.00 X 10° 9.48 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.29 X 10 2.73 X 10''' 1.26 X 10^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° ORNL 0.00 X 10° 1.12 X 10° 0.00 X 10° 0.00 X 10° 4.42 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 4.56 X 10"^ 0.00 X 10° 1.87 X 10"® 1.96 X 10-^ 0.00 X 10° 1.18 X 10"^ 3.67 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 1.88 X 10'^ 1.69 X 10 2.91 X 10 9.18x10° 2.98x10^ 1.34x10^ 9.68x10^ 1.68x10^ Radioactivity In curies per shipment for the volumes of waste assumed for the SEIS analyses (i.e., volumes scaled up to correspond to the design capacity of the WIPP--see last column, Table B.2.4). The volume per shipment is 0.89 m^ (one shipping cask per shipment). u Key: ANLE, Argonne National Laboratory-East; HANF, Hanford Reservation; INEL, Idaho National Engineering Laboratory; LANL, Los Alamos National Laboratory; ORNL, Oak Ridge National Laboratory. B-15 TABLE B.2.8 Quantities used in estimating the average radio- activity in a shipment of CH TRU waste from Rocl. ■/ For transportation under normal conditions, the radiological risk depends on the radiation field at the surface of the shipping container or cask. This field is measured in terms of the transport index (Tl), which is the radiation-dose rate (in mrem per hour) at 1 m from the surface of the container or cask and is used in calculating radiation exposures under normal transportation conditions. The radiation field measured by the transport index comes mainly from the gamma radiation released by fission products and other radionuclides (i.e., activation products) in the TRU waste. In CH waste, these products exist in trace amounts and do not contribute sufficient gamma radiation to exceed the limit of 200 mrem per hour for the radiation-dose rate at the surface. These trace amounts are therefore not usually reported in the CH waste inventories. In RH waste, the activation and fission products exist in more significant amounts, as shown in Table B.2.7. The gamma radiation from these products results in radiation-dose rates exceeding 200 millirem per hour and is the reason the waste is assigned to the category of remotely handled, rather than contact-handled, waste. For the Tl used in these SEIS analyses, data from the 1987 Integrated Data Base and the updated radionuclide data (DOE, 1989c) were supplemented with information from the waste facilities. This supplemental information concerned field measurements of the gamma radiation levels around Type A TRU waste containers such as drums and standard waste boxes. The objective of this data-collection effort was to develop a listing of waste containers in terms of the maximum surface dose rates for each facility. From this information, an average for the maximum surface dose rate for the containers from each waste facility was calculated. To ensure that the radiation field was not underestimated, it was assumed that this field resulted entirely from radionuclides emitting photons with an energy of 1 million electron-volts (MeV). In actuality, most of the gamma radiation from CH TRU waste results from the radioactive decay of americium-241 and has an energy of 0.060 MeV. The 0.060 MeV gamma radiation would be significantly attenuated by the TRUPACT-ll, while the 1 MeV gamma radiation would not be. The assumption of 1 MeV gamma radiation resulted in radiation levels that exceeded and bounded the expected radiation levels. Shielding models of the TRUPACT-ll containers and the shipping cask for RH waste were then developed to calculate the transport index from the 1-MeV radiation fields. In some cases, the lack of waste-specific information (as in the case of the RH waste from Hanford Reservation) necessitated an assumption about the radiation field. For I this SEIS, the Hanford RH waste was assumed to produce a field of 100 rem per hour from the 1-MeV photons (100 rem per hour is the upper limit for 95 percent of the RH waste to be received at the WIPP; the remaining 5 percent may have radiation fields of I up to 1,000 rem per hour). This very conservative assumption resulted in a high transport index for RH waste shipments from Hanford Reservation in comparison with the other facilities. I For the CH waste from each waste facility, the number of truck shipments (three I TRUPACT-ll containers per shipment) was estimated by multiplying the volume per drum (0.2 cubic m) by the number of drums per shipment (42 drums) and dividing this I number into the total volume (in cubic meter) of TRU waste (stored and newly generated) at the facility. For rail shipments from facilities with rail access, it was assumed that each shipment carried six TRUPACT-ll containers. For the RH waste, since only one cask will be sent per shipment, the number of shipments was obtained by dividing the volume per shipment (in cubic meters) by the volume per shipping cask (0.89 cubic m). Rail shipments were assumed to carry two B-18 ■r^y. casks per shipment. In all of the shipment calculations, the waste was assumed to be the same as in the above-described calculations of radioactivity per container and the Transport Index. B.2.4.2 Source Term for WIPP Operational Analysis For this SEIS, the analysis of radiation safety during WIPP operations (waste receiving, handling, and emplacement underground) was derived from the WIPP draft FSAR (DOE, 1989b). The safety analyses in the draft FSAR were based on waste inventories reported in Radionuclide Source Term for the WIPP (DOE, 1989c). These safety analyses were scaled up to correspond to the volume design capacity of the WIPP. Scaled-up inventories were used to calculate the number of containers (55-gal drums, standard waste boxes, canisters) that may be processed annually at the WIPp' Average characteristics were also calculated for containers of OH waste (55-gal drums and standard waste boxes) and RH waste (canisters), as shown in Tables B.2.10, B.2.11, and B.2.12. The average radioactivity per container was used in the draft FSAR and the SEIS to analyze the impacts of both normal operations and accidents. Impacts from accidents involving containers with the maximum allowable contents, per the Waste Acceptance Criteria (DOE, 1989a), were also assessed. In assessing occupational safety, the radiation exposures of workers handling waste at the WIPP were based on the same assumptions about radiation fields as those used to calculate the transport index in the transportation-impact analysis. B.2.4.3 Source Term for Long-Term Performance Analyses The source term used in assessing the long-term performance of the WIPP was derived from the scaled-up waste volumes (Table B.2.4) and the radionuclide composition reported in the waste-characterization data base for the WIPP (DOE, 1989c). A discussion of the source term requirements for the long-term performance analyses, including the decay chains, is in Lappin et al. (1989). The total inventory of CH TRU waste of approximately 11.4 million curies (Table B.2.13) was modified to account for the decay of short-lived nuclides and the buildup of daughter products with high radiotoxicity (100 years for institutional controls). In addition, radionuclides with low radiotoxicity were eliminated from the inventory. The modified Inventory (Table B.2.14) is approximately 3.8 million curies. The RH TRU waste is not included in the long-term performance-assessment inventory because RH TRU waste constitutes less than 2 percent by activity. Also, as discussed by Lappin et al. (1989), the procedures for emplacing waste in the WIPP will minimize the interaction of RH waste canisters and CH waste rooms. And many of the short- lived radionuclides (which are typically the reason for the waste being assigned to the RH category) will have minimal consequences over the long term. An analysis has been made of the consequences of RH TRU waste being brought directly to the surface by an intruding borehole (see Subsection 5.4.2.6). B-19 TABLE B.2.10 Mass and radioactivity of \he radionuclides in an average drum of CH TRU waste^ Mass (g) Radioactivity (curies) Radionuclide FEIS'^ FSAR^ FEIS^ FSAR"^ Thorium-232 NP 6.0 X 10° NP 6.6 X 1 0-"^ Uranium-233 NP 1.7 X 10° NP 1.7 X 10-2 Uranium-235 NP 4.0 X lO-"" NP 8.8 X 1 0-"^ Uranium-238 NP 1.0 X 10^ NP 3.5 X 10'^ Neptunium-237 NP 3.1 X 10-2 NP 2.2 X 1 0-^ Plutonium-238 2.5 X 1 0-^ 6.2 X lO'"" 4.2 X 10"^ 1.1 X 10^ Plutonium-239 7.5 X 10° 1.4 X 10^ 4.6 X lO""" 8.5 X lO'"" Plutonium-240 5.0 X lO-"" 8.5 X lO-"" 1.1 X lO-"" 1.9 X lO'"" Plutonium-241 2.7 X 1 0'^ 6.6 X 10-2 2.8 X 10° 6.8 X 10° Piutonium-242 2.4 X 10'^ 7.8 X 10-^ 9.4 X 10'^ 3.1 X 10"^ Annericium-241 1.5 X 10-^ 4.9 X lO-"" 5.2 X 10"^ 1.7x10° Curium-244 NP 4.2 X 1 0''^ NP 3.4 x 10"^ Californium-252 NP 1.0 X 10'^ NP 5.4 X 10'^ TOTAL 8.0 X 10° 3.4 X 10^ 3.4 X 10° 2.1 x 10^ j ^ The reasons for the differences between the 1 980 FEIS and the draft FSAR values I are discussed in Section B.I. j ^ From the WIPP FEIS (DOE, 1980). NP indicates that data were not provided in the FEIS. ] ^ From the WIPP draft FSAR (DOE, 1989b). These values were also used in the SEIS j analyses. B-20 TABLE B.2.11 Mass and radioactivity of the radionuclides in an average standard waste box of CH TRU waste^ Mass (g) Radioactivity (curies) Radionuclide FEIS'' FSAR° FEIS^ FSAR^ Thorium-232 NP 1.2 X 10^ NP 1.3 X 10"^ Uranium-233 NP 6.7 X 10° NP 6.5 X IQ-^ Uranium-235 NP 9.6 X lO-"" NP 2.1 X 10-^ Uranium-238 NP 2.5 X 10^ NP 8.3 X 10'^ Neptunium-237 NP 4.4 X 1 0-"^ NP 3.1 X IQ-'' Plutonium-238 4.0 X 10"^ 4.2 X 10-2 6.8 X 10'2 7.2 X IQ-"" Plutonium-239 1.2 X 10^ 7.9 X 10^ 7.5 X lO-"" 4.9 X 10° Plutonium-240 8.1 X 10"'' 6.5 X 10° 1.8 X lO""" 1.5 X 10° Plutonium-241 4.4 X 10"2 6.7 X lO""" 4.5 X 10° 6.9 X 10^ Plutonium-242 3.9 X 10-2 7.5 X 10-2 1.5 X 10-^ 2.9 X 1 0-"^ Americium-241 2.5 X 1 0-^ 2.1 X lO-"" 8.4 X 10-3 7.3 X 10-^ Curium-244 NP 8.6 X 10-^ NP 7.0 X 10"^ Californium-252 NP 2.1 X 10-^ NP 1.1 X 10'^ TOTAL 1.3 X 10^ 1.3 X 10^ 5.5 X 1 0° 7.7 X 10^ ^ The reasons for the differences between the FEIS and the draft FSAR values are [ discussed in Section B.I. ' ^ From the WIPP FEIS (DOE, 1980). NP indicates that data were not provided in the j FEIS. ^ From the WIPP draft FSAR (DOE, 1989b). These values were also used in the SEIS | analyses. i B-21 TABLE B.2.1 2 Radioactivity of the radionuclides In an average canister of RH TRU waste^ Radioactivity (curies) Radionuclide FEIS b.d FSAR^'^ Cobalt-60 Strontium-90 Ruthenium-1 06 Antimony-125 Cesium-137 Cerium-144 Uranium-233 Uranium-235 Uranium-238 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Plutonium-242 Americium-241 Curium-244 Californium-252 TOTAL 1.6 X 10" 2.5 X 10^ 2.2 X 10° NP 1.2 X 10° NP NP NP NP 6.5 X 10-2 7.5 X lO-"" 1.8 X lO""" 4.6 X 10° NP 1.2x1 0"2 NP NP 2.6 x 1 0^ 1.7 x 5.1 X 3.5 X 1.1 X 4.3 X 3.4 X 5.5 X 3.0 X 1.5 X 5.7 X 6.8 X 2.2 X 10" 10^ 10'' 10-^ 10^ 10"' 10 10"^ 10"^ 10^ 10^ 10^ -3 1.2 X 10 + 1 3.8 X 10' 2.1 X 10' 1.6 X 10- 2.8 X 10' 3.7 X 10^ I ® The reasons for the differences between the FEIS and the draft FSAR values are I discussed in Section B.I. I ^ From the WIPP FEIS (DOE, 1980). NP indicates that data were not provided in the FEIS. ] ^ From the WIPP draft FSAR (DOE, 1989b). These values were also used in the SEIS ] analysis. ^ Daughter products not included. B-22 TABLE B.2.13 Initial radionuclide inventory in CH TRU waste for the assessment of long-term performance^ Radionuclide Thorium-232 Uranium-233 Uranium-235 Uranium-238 Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Plutonium-242 Americium-241 Curium-244 Californium-252 TOTAL Half-life (years) 1.41 X 10 10 1 .59 X 1 0^ 7.04 X 1 0^ 4.47 X 1 0^ 2.14 X 10^ 8.77 X 10^ 2.41 X 10"^ 6.54 X 1 0^ 1.44 X 10^ 3.76 X 10^ 4.32 X 1 0^ 1.81 X 10^ 2.64 X 10° Radioactivity (curies) 3.07 X 10"^ 9.48 X 10^ 4.59 X lO""" 1 .84 X 1 0° 1.08 X 10^ 5.25 X 1 0^ 4.89 X 10^ 1.20 X 10^ 4.70 X 10^ 2.13 X 10^ 7.72 X 10^ 1.57 X 10"* 2.51 X 10^ 1.14 X 10'' ^ This source term is different from that given by Lappin et al. (1989), because it was scaled up to correspond to the design volume of the WIPP. This was done by scaling the source term, by radionuclide, at each waste facility by the volume increment for that facility. B-23 TABLE B.2.14 Modified radionuclide inventory in CH TRU waste for the assessment of long-term performance^ Radionuclide Half-life (years) Radioactivity (curies) Mass (g) Plutonium-238 8.77 X 10^ 2.38 X 10^ 1.39 X 10^ Plutonium-239 2.41 xlO^ 4.89 X 10^ 7.87 X 10^ Plutonium-240 6.54 X 10^ 1.20 X 10^ 5.26 X 10^ Uranium-233 1.59 X 10^ 9.48 X 10^ 9.82 X 10^ Uranium-234 2.44 X 10^ 1.03 X 10^ 1.64 X 10^ Uranium-235 7.04 X 10^ 4.59 X 10-1 2.12 X 10^ Uranium-236 2.34 X 10^ o'^ Americium-241 4.32 X 10^ 7.94 X 10^ 2.31 X 10^ Neptunium-237 2.14 X 10^ 1.08 X 10^ 1.53 X 10^ Thorium-229 7.43 X 10^ 0*^ Thorium-230 7.70 X 10^^ 0^ Radium-226 1.60 X 10^ 0^ Lead-21 2.23 X 10^ 0*^ TOTAL 3.79 X 10^ ^ The radionuclide inventory in Table B.2.13 was modified by assuming that the radioactivity has decayed for 100 years and, therefore, removing the nontransuranic radionuclides, except uranium. ^ The radionuclides with zero activity are listed to establish initial amounts for all radionuclides in the decay chains shown in Table 4-3 of the report by Lappin at al. (1989). B-24 i B.3 HAZARDOUS CHEMICAL CONSTITUENTS The FEIS (DOE, 1 980) addressed only the impacts of the radioactive component of TRU waste. Since that time, it has been determined that TRU waste is subject to dual regulation under the Atomic Energy Act and the Resource Conservation and Recovery Act (RCRA) because it may also contain hazardous chemical constituents; such waste is called TRU mixed waste. TRU mixed waste is defined as waste that is contaminated with transuranic radionuclides at levels exceeding 100 nCi per g of waste and with hazardous chemical constituents. Information provided by the DOE waste generators indicates that 60 percent of the total TRU waste proposed to be sent to the WIPP over 25 years of operation will contain hazardous waste that is subjected to regulation under RCRA. All shipments of mixed waste are required to meet the conditions of RCRA and the U.S. Department of Transportation (WEC, 1989). Until recently, few records were required to document the hazardous chemical constituents in TRU waste. The waste was and currently is not routinely sampled and analyzed, because some of the waste is contained in complex matrices and such sampling activities might expose personnel to unacceptable levels of radiation. However, it was possible to determine the composition and other characteristics of TRU mixed waste from knowledge about the waste and the industrial processes from which it was generated. For example, because of the requirements for strict product quality and concerns for safety in handling radioactive materials, production and research activities are highly structured. The ingredients used in a given process and the process conditions are highly controlled. This precision both requires and generates extensive knowledge of the ingredients and the processes involved; it also facilitates the characterization of TRU mixed waste. This section discusses the hazardous chemical constituents in TRU waste. This information serves as the basis for estimation of the amount of hazardous chemicals that would be released in a given situation. B.3.1 CH TRU MIXED WASTE The DOE facilities that may ship waste to the WIPP have used very conservative approaches characterizing their CH TRU mixed waste (i.e., approaches that are likely to overestimate the hazardous chemical constituents in the waste). The conservative approaches were chosen to facilitate preparation of the permit application to operate the WIPP as an "interim status" facility under the RCRA. The characteristics of the waste were recently reported in the Radioactive Mixed Waste Compliance Manual (WEC, 1989) and represent a conservative upper bound for the concentrations of hazardous chemicals in the waste. In other words, if a chemical is present in the waste, it is identified even though its concentration in the waste may be below the regulatory limit. B-25 The identification of the hazardous chemical constituents in CH TRU mixed waste is based on newly generated waste from the Rocky Flats Plant and waste from the Rocky Flats Plant that is currently in retrievable storage at the Idaho National Engineering Laboratory. It is estimated that this waste represents approximately 86 percent by volume of the total CH TRU mixed waste proposed to be emplaced in the WIPP over the 25-year operating life. Furthermore, the Rocky Flats Plant generates many different forms of waste from a variety of processes. Other DOE facilities generate smaller quantities of TRU mixed waste, fewer categories of waste, and waste that contains a narrower range of hazardous chemical constituents (WEC, 1989). Therefore, data on the stored or newlv generated waste from Rocky Flats Plant represent a conservative upper bound for the potential risks associated with the chemical components of the CH TRU mixed waste. In the WIPP Waste Acceptance Criteria (See Section 2 and Appendix A), CH TRU waste is divided into several categories based on the physical characteristics of the materials in the waste. These categories or forms are used by each DOE waste facility to classify its TRU mixed waste. Before shipment to the WIPP, each waste form must be certified by the DOE for compliance with the WIPP Waste Acceptance Criteria. Waste forms identified by the Rocky Flats Plant as containing hazardous chemical constituents are cemented and uncemented aqueous and organic waste, cemented process and laboratory solids, combustible waste, metal and filter waste, inorganic solids, and leaded rubber waste. Each of these waste forms is briefly described below: • Cemented and uncemented aqueous process waste. This waste consists of a wastewater4reatment sludge that is precipitated at a pH of 10 to 12. The sludge contains alcohols and halogenated organics from the cleaning of equipment and glassware and the degreasing of metal. Some aqueous process waste may also contain metals (e.g., cadmium and lead), although no analyses have been performed to determine specific concentrations. Since 1984, aqueous process waste has been solidified in a process involving neutralization, precipitation, flocculation, clarification, filtration, and solidification with portland cement. Before 1984, this waste was not cemented and it exists today as a damp solid. • Cemented and uncemented organic waste. Organic waste consists of lathe coolants and degreasing solvents used in plutonium fabrication. Organic waste containing oil and halogenated organic solvents is solidified with Envirostone cement and an emulsifier. Before 1984, this waste was not solidified with cement; it is a damp solid. • Cemented (immobilized) process and laboratory solids. This waste consists of ion-exchange resins and incinerator ash that has been neutralized and solidified with portland cement. The solvents in this waste come from plutonium-recovery operations. • Combustible waste. This waste consists of paper and cloth (dry and damp); various plastics, such as polyethylene and polyvinyl chloride; wood; and filters contaminated with trace quantities of halogenated organic solvents. B-26 These materials are generated during plutonium recovery and fabrication and in analytical laboratories. • Metals. The principal constituents of this waste are lead, tantalum, stainless steel, and aluminum. This waste includes equipment, tools, crucibles from laboratories, and molds. Residual halogenated organic solvents may also be present. • Filters. This waste consists of polypropylene filters and high-efficiency particulate air filters as well as processed filter media. Portland cement is added to absorb any residual liquid and to neutralize residual acids. Exhaust-stream filters may be contaminated with volatile organic solvents used in plutonium fabrication and recovery. • Inorganic solid waste. This waste contains materials like firebrick. Oil Dri, concrete, and soil. It is generated from the decontamination and decommissioning of plutonium-recovery areas. Oil Dri, concrete, and soil may be contaminated with residual halogenated organic solvents. • Leaded-rubber waste. This waste consists of the leaded rubber dry-box gloves and aprons that are used throughout plutonium-processing areas. It is considered an RCRA-regulated hazardous waste according to the EPA extraction procedure toxicity test (40 CFR Part 261) for lead, although no analysis has been done to establish the lead concentrations. The EPA toxicity test is used to characterize waste as hazardous under the RCRA. The estimated quantity of each waste form is given in Table B.3.1. The above descriptions indicate that most of the organic solvents are present in residual quantities from the cleaning of equipment, plastics, glassware, and filters. A major constituent in CH TRU mixed waste is lead, which is present mainly in shielding, dry-box parts, and lead-lined gloves and aprons. The types and estimated maximum concentrations of hazardous chemical constituents in the various forms of CH TRU mixed waste are given in Table B.3.2. This information is used to determine the types of hazardous chemicals expected in various waste forms and their relative abundance. The concentrations, estimated by the Rocky Flats Plant (Rockwell International, 1988) from knowledge of the waste-generating processes, are very conservative and do not represent the actual concentrations of these chemicals. Information from Clements and Kudera (1985) indicates that the volatile organic compounds in the headspace of drums are well below saturation values for the various chemicals and that the source is limited. A description of the actual hazardous chemical source term used in the hazardous chemical risk assessment is provided in Subsection 5.2.4. B.3.2 RH TRU MIXED WASTE As discussed in Subsection 2.3, RH TRU waste represents a much smaller portion than CH TRU waste of the total waste proposed for shipment to the WIPP site: the design B-27 capacity for RH TRU waste at the WIPP is 250,000 cubic feet. Oak Ridge National Laboratory reported the following two major waste forms in the Radioactive Mixed Waste Compliance Manual (WEC, 1989): TABLE B.3.1 Estimated quantities of TRU mixed waste (by waste form) from Rocky Flats Plant a,b Description of waste form Cemented and uncemented aqueous waste Cemented and uncemented organic waste Immobilized process and laboratory solids Combustible waste Metal waste Filter waste Inorganic solid waste Leaded rubber waste Quantity (kilogram) 1.35 X 10' 3.27 X 10'^ 3.38 X 10^ 6.66 X 10^ 9.65 X 1 0^ 2.21 X 10' 4.15 X 10- 3.64 X 10- Total 3.64 X 10' ^From the Radioactive Mixed Waste Compliance Manual . (WEC, 1989), Appendix 6.4.1. ^ Quantities include waste projected to be generated through the year 201 3 and waste in retrievable storage at the Idaho National Engineering Laboratory. B-28 m c a> 3 •? (/) c 8 (0 o E 0) x: o c V) (0 D Q. O ■D CT3 CO N LL CO x: ^ o O o (0 CE c g 0) ^— » ^ TO ■*"" -*— • c E o o >♦— c o Q) o CO CO E § :3 E ■o CD X X <0 E £ 3 "O cc 1- CT3 E X o w LU c C\J CO od LU _l OQ < \- CO E CO O) o 0) Q. cn E CO c o *-» CO c 0) o c o o E 13 E X CO E ■D CD -*— » CO E CO LU ■D CD ■D CO CD CD *-« CO 3 o c CO CO -D ^ o O CO c »_ CD CD t) CO iZ § ro o CO CD CO ^ $ Q) XJ ■■Jo CD 3 CO JQ CO E 5 o O T3 CO T3 ;g C CO o CO CO *«.^ CO ^ CD o O o "co o m o L> c 0) CO w O) CO 6 $ CO D J o CD CD CO 3 CO CT § < (0 o E 0) JH "iD <0 ID O ■D c CD 3 CO CO N c CO o I o o o o o o o o o o o o 8 CX) o o o in in o X to o o o o o o 1- ^ T- "^ in o in o ,-^ o o S o o iP in CM in CM CVJ o o o o" in o o o o o o o" o" in in in in >=i h- OJ ° o o o o o CO o ° o o o o o CM X in o o o o ° in o o o "j h^ b^ o o ^ -^ Ul T- ^ o o o o o o o o in o o o o cvj in -.- T- T- - o ;-- o iii _ 7= t:t 7= o CD ,o +- o ^T oh- T-_ n cj T-' CO -r-" ^ O T-* CD CD "D C IZ CO O — CD O -§ CD O c lo CD O ■^ I >, >. 05 *« ~ 2 2 X CQ O _J C CD CD ^ > 0)0 (D ^ I— J*: CD e 5 CD $ O CO CD CD £ £ ^£ o)E 3 .b CO CO ? ^ CD c C c CD ■o ••;= CD (0 CO =5 0) CD CO ^2 CO p 3 CO CO o "^ ^ " -^ o o o fl^ -i "co o CD -t; E i: (D CD o S o ?^ c CO D) — o -o CO N CO sz CO CD E o x: O "S CD CO CO ■o v> 0) .■D o 1 CD CO "o CO CX) CO en CO o CO ■D "co CO T o 9- (J Q) CD ^*"*^ D " O) E lo c — CD to -c >< CO "D 3 CD 3^ o C "^ c CO TO c dous chemical co concentrations for CD O 2 ■D CD •«— • C CD CD 5 c o o cr "co E CO 0) E CD O c •D C CO o E E ■D C CO ■D E o N 1 -1 > o "D 0) c CD CD .N 2 CO CD 2S CO Q) E "3 CO -c 2 CD CD Q 1- E O) (J z (0 n O T3 B-29 • Solid RH TRU mixed waste. This waste contains mixtures of combustible materials (e.g., paper, polyvinyl chloride, polypropylene, polyethylene, and Neoprene) and noncombustible materials (e.g., laboratory equipment, tools, and small electric motors) that were removed from an experimental facility at the Oak Ridge National Laboratory (the Alpha Gamma Hot Cell Facility). This waste does not contain free liquids or particulates. • Sludges. This waste consists of fuel and process sludges that are currently stored in tanks but will be solidified before shipment (with cement or by exposure to microwaves). This waste will be solid packaged in lead- shielded canisters. The primary hazardous chemical constituent of RH TRU mixed waste is lead, which is used to provide shielding against gamma radiation. Trace quantities of mercury, barium, chromium, and nickel have also been reported in some of the sludges. B-30 REFERENCES FOR APPENDIX B Clements, T. L, Jr., and D. E. Kudera, 1985. TRU-Waste Sampling Program. Informal Report . Vols. I and II, EEG-WM-6503, TLC-82-85, EG&G Idaho, Idaho Falls, Idaho. DOE (U.S. Department of Energy), 1989a. TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant . WIPP-DOE-069, Rev. 3, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1989b. Final Safety Analysis Report, Waste Isolation Pilot Plant , draft, DOE/WIPP 89-XXX, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1989c. Radionuclide Source Term for the WIPP . 88- 005, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1989d. Integrated Data Base for 1989: Spent Fuel and Radioactive Waste Inyentorles. Projections, and Characteristics . DOE/RW- 0006, Rev.5. DOE (U.S. Department of Energy), 1988. Integrated Data Base for 1 988: Spent Fuel and Radioactive Waste Inventories. Projections, and Characteristics . DOE/RW- 0006, Rev. 4. DOE (U.S. Department of Energy), 1987. Integrated Data Base for 1 987: Spent Fuel and Radioactive Waste Inventories. Proiections. and Characteristics . DOE/RW- 0006, Rev. 3. DOE (U.S. Department of Energy), 1980. Final Environmental Impact Statement. Waste Isolation Pilot Plant . WIPP-DOE-0026, Vols. 1 and 2, Carlsbad, New Mexico. EPA (U.S. Environmental Protection Agency), 1982. Federal Register . Vol. 47, No. 250 p. 58196. ' I Lappin et al. (A. R. Lappin, R. L Hunter, D. Garber, and P. B. Davies, eds.), 1989. Systems Analysis. Long-Term Radionuclide Transport, and Dose Assessments. Waste Isolation Pilot Plant (WIPP). Southeastern New Mexico . March 1989, SAND89-0462, Sandia National Laboratories, Albuquerque, New Mexico. Rockwell International, 1988. "Hazardous Constituents of Rocky Flats Transuranic Waste," Internal Letter for distribution from J. K. Paynter, May 24, 1988, WCP02- 31, Golden, Colorado. WEC (Westinghouse Electric Corporation), 1989. Radioactive Mixed Waste Compliance Manual . Appendix 6.4.1, WP-02-07, Rev. 0, Waste Isolation Pilot Plant, Carlsbad, New Mexico. B-31/32 ■ ■ -N ■ .. %". ■yy.'-A-'.- APPENDIX C TRANSPORTATION EMERGENCY PLANNING C-i/ii TABLE OF CONTENTS Section Page C.1 INTRODUCTION C-1 C.2 OVERVIEW OF RESPONSIBILITIES AND RESOURCES IN EMERGENCY RESPONSE C-3 C.2.1 Overview of Responsibilities C-3 C.2.2 General Responsibilities and Resources of State, Tribal, and Local Governments C-4 C.2.2.1 Response Plans C-5 C.2.2.2 Evacuation Plans C-5 C.2.2.3 Capabilities C-6 C.2.3 Federal Assistance in Emergency Response 0-7 C. 2.3.1 The Federal Emergency Management Agency and the Framework for Federal Assistance C-7 C.2.3.2 The Emergency-Response Resources of the DOE .... C-9 C.2.3.3 Guidance to State, Tribal, and Local Govern- ments for Emergency Response to Transportation Accidents C-1 4 C.2.3.4 Federal Emergency-Response Training C-1 5 C. 2.3.5 Federal Information Services for Radiological Emergencies C-1 6 C.2.3.6 Financial Responsibility for Transportation Accidents C-1 6 C.3 EMERGENCY-RESPONSE PU\N FOR WASTE TRANSPORTATION TO THE WIPP C-18 C.3.1 Emergency-Response Procedures for the Carrier C-18 C.3.1.1 Procedures for the Drivers of the Vehicles C-21 C.3.1. 2 Procedures for the Dispatcher of the Trucking Contractor C-22 C.3.1. 3 Insurance C-23 C.3.2 Emergency-Response Procedures for the State, Tribal, and Local Governments C-23 C.3.3 Procedures for Responses by the DOE and Its Contractors .... C-23 C.3.3.1 Procedures for the Central Coordination Center at the WIPP C-23 C.3.3.2 Procedures to Be Followed in the TRANSCOM Control Center C-24 C.3.3.3 Emergency-Response Responsibilities of Other DOE and DOE-Contractor Organizations C-25 C-iii Paqe Section — '^ C.3.4 Emergency-Response Training ^"26 C.3.4.1 Introduction ^'^^ C.3.4.2 iVIedical Response Training C-28 C.3.5 Assistance to Medical Facilities ^-30 C.3.5.1 Hospital Planning and Capabilities C-30 C.3.5.2 Specialty Drugs ^'^ C.3.6 Funding ^■^' C.4 EMERGENCY-RESPONSE SCENARIOS C-33 C.4.1 Scenario for a Hypothetical Severe Transportation Emergency C.4.2 Radiological Assistance Response: Burley, Idaho- October 12, 1986 ^"^^ C.4.3 Radiological Assistance Response: Pocatello, Idaho- October 10, 1985 ^'^^ REFERENCES FOR APPENDIX C ^'^^ C-iv y^ m^ii^ LIST OF FIGURES Figure Page C.2.1 U.S. Department of Energy Regional Coordinating Offices C-10 0.3.1 Activities performed by State, Indian Tribal, and local authorities and the DOE in response to a transportation emergency involving TRU waste 0-19 0.3.2 Typical notifications that would be made after a transportation accident involving a shipment to the WIPP 0-20 0-v/vi C.1 INTRODUCTION ■f. y Of 500 billion domestic shipments annually, about 100 million, or 0.02 percent, are shipments of hazardous materials, and 3 million, or 0.0006 percent, are shipments of radioactive materials. The vast majority (95 percent) of the radioactive-material shipments involve small quantities for general users like hospitals, research laboratories, and industries. The remaining 5 percent are large quantity shipments for commercial reactors or shipments related to national defense (Wolff, 1984). The safety record of the radioactive-material shipments is outstanding. No serious injuries or deaths have ever resulted from the radioactive materials carried in these shipments. The main reason for this outstanding safety record is the stringent Federal requirements for the packagings, shipping containers, and shipping casks that must be used for radioactive materials. Accidents that have released radioactive material from limited quantity, or Type A containers, have resulted in insignificant consequences and in each case the material was cleaned up, and no one was injured from the radioactivity. Large quantity, or Type B containers and casks are occasionally involved in transportation accidents; fifty such containers or casks were involved in accidents between 1971 and 1985 (DOE, 1989a). No Type B packages have ever released their radioactive contents because of impact or fire, except for a radiography camera failure. As described in Appendix L, the packagings that will be used for shipping TRU waste to the WIPP are in the Type B category and are designed to withstand severe accidents without releasing their contents. However, as an additional precaution the DOE continues to ensure its emergency-response capabilities and procedures to protect public health and safety after transportation accidents. The current status of those capabilities and the plans for their future development are discussed in this appendix. Planning for radiological emergency preparedness, including transportation activities, began several years ago. State, Tribal, and local governments as well as the DOE and several other Federal agencies have been closely involved in this effort. The Federal effort includes developing transportation-specific planning guidance and reviewing generic State radiological emergency-response plans. This appendix describes the responsibilities and resources available for responding to emergencies in general and transportation accidents in particular. Then it presents a detailed discussion of the emergency-response responsibilities in transportation to the WIPP and presents the procedures to be followed by the carrier of the waste (i.e., the WIPP trucking contractor); the State, Tribal, and local governments; and various organizations in, or employed by, the DOE. The subsection on procedures is followed by a discussion of the training programs that the DOE has conducted in various States. To illustrate how the carrier, the State and local governments, and the DOE would respond in a given accident situation, the last subsection in this appendix describes a hypothetical accident and emergency-response scenario. In addition, it describes the C-1 responses to actual transportation accidents and incidents involving radioactive materials. This appendix has been rewritten in response to the many comments received which requested additional clarification and detail concerning emergency-response capabilities and plans in the event of transportation accidents. C-2 C.2 OVERVIEW OF RESPONSIBILITIES AND RESOURCES IN EMERGENCY RESPONSE In the Civil Defense Act of 1950, the U.S. Congress broadly defined the roles and responsibilities of the Federal Government in responding to nuclear attacks and other emergencies in general. Following a tradition established early in the history of the United States, the Act assigned to State and local governments primary responsibility for implementing measures to protect life and property, whereas Federal agencies were given responsibility for providing assistance when requested by State, Tribal, and local governments. Subsequently, responsibilities were also defined for the shippers and carriers of hazardous materials, including radioactive waste. This subsection reviews emergency-response responsibilities and roles. It also discusses the resources that are available for emergency response. The discussion is not specific to WIPP transportation; emergency response for WIPP transportation is discussed in Subsection C.3. C.2.1 OVERVIEW OF RESPONSIBILITIES The general roles of shippers; carriers; State, Tribal, and local governments; and Federal agencies can be summarized as follows: • Shippers. The shipper Is required to provide to the carrier, at the time of shipment, any special precautions required for each shipment. If called on in case of an accident, the shipper will also provide information that may be necessary for, or helpful in, emergency-response activities. • Carriers. The carrier has the initial responsibility for minimizing radiation hazards to the public and notifying State, Tribal, and local authorities of accidents in their jurisdictions. • States. T ribal, and local governments . These entities have primary responsibility for implementing measures at the scene of the accident in order to protect life, property, and the environment. • Federal agencies. If requested, assistance from Federal agencies is available to support the emergency-response measures taken by State, Tribal, and local governments. In the case of transportation to the WIPP, the DOE has responsibilities in two of the above categories: 1) the DOE is the shipper, and 2) the DOE is a Federal agency that can provide assistance if requested by State, Tribal, or local governments. As shipper and owner, the DOE would respond directly to transportation accidents involving the TRU waste. \ I 1 I C-3 This subsection describes the State, Tribal, and local responsibilities for ernergency response. Although State, Tribal, and local governments have a more important ole in emergency response, and Federal assistance is rarely requi^d ,n a t'«n=P°rt^*°" accident this subsection also presents a comprehensive discussion of Federa eS^rgency-response resources which allows the reader to understand the types of assistance that are available to State, Tribal, and local governments. C2.2 flENERAL RESPONSIBILITIES aNn RESOURCES OF S TAT E. TRIBAL, AND LOCAL GOVERNMENTS In the event of a transportation accident involving radioactive waste, State, Tribal, and ocalgovernments are responsible for taking measures to protect life, PW, and the environment. This might entail direct actions, such as rescuing people from a wreck eXquishing fires, and giving first aid to the injured, as well as protective actons, such as kee^mTpeople away from the area of the accident. These are activities that "sually occur wimin the first 30 minutes of a response and are normally performed by local qove nmenls If the local government determines that its response capabilities have been exceeded, which is often the case in incidents involving <-'"°^^^l'^'';''^^ they would request additional radiological monitoring and assessment he p from a State government organization. In addition, State, Tribal, and local Q^f ~^ ™?' '"'^" that cleanup and decontamination activities, if necessary, meet their standards. In 1980 the Nuclear Regulatory Commission (NRC) published a survey' (t*/litter et al, ^oTofltaTemerge'ncy.e'Iponse capabilities for responding to tran^^^^^^^^^^^^ accidents The NRC Survey reports that the number of requests for State assistance in transportation accidents involving radioactive materials is 275 per year, or a rnean ot 5 eTequests per State per year. Many of the States responding to the sun/ey stressed fhat most of ?hese accidents are not serious, the shipping containers o' casks retain their integrity and there is rarely any release of radioactive material. Some o he responS mentioned that they were more concerned about aoade".s invoking hazardous chemicals. However, knowledge that most transportation accidents involving adioactive materials are not serious does no. diminish the need '-"^J^^^^^^^^^^^ at the scene because hasty decisions or actions by uninformed personnel can lead to unnecessa^' panfc In one accident, for example, a civil-defense volunteer who was amona the first responders used a pocket dosimeter that had not been calibrated fo mTefhan a year The worker's defective dosimeter indicated a near-lethal reading o radiation dose, causing an entire township to panic. The State response team later determined that there had been no radiation leakage. Fortv six States responding to the NRC sun/ey (Mitter et al., 1980) reported that they haSf needed to call on Federal assistance in transportation accidents involving adioaS material. Four o, these States, however, have DOE instal^^ns w.l.r, e hnrders- these installations are routinely notified and respond on behalf of me State, it mey are Ihe nearest source of qualified personnel. Only three States reported having This survey is currently being updated. C-4 called for Federal personnel, and one of these stated that they asked for Federal assistance to verify the integrity of shipping casks that had been involved in a rail accident. In addition, several States mentioned that in some incidents involving shipments from or to Federal installations, the drivers had notified the Federal install- ation, which sent personnel to respond. As discussed in Subsections C.2.3.1 and C.2.3.2, when the DOE is the shipper, the DOE will respond automatically. If DOE receives notification of an accident from its carrier, they will provide this information to the State and coordinate the response. To be prepared to respond, it is necessary to develop and implement emergency- response plans. The rest of this subsection briefly describes planning by State, Tribal, and local governments; guidance for evacuation plans; and capabilities. C.2.2.1 Response Plans State, Tribal, and local governments are generally responsible for providing the first response to a transportation accident. In addition, according to a guidance document issued by the Federal Emergency Management Agency (FEMA, 1988), the local govern- ment must determine the action required in order to prevent further damage to life or property. (State and local statutes should be consulted to determine specific responsi- bilities.) Cleanup and decontamination may be performed by any of a number of organizations, but the carrier and shipper have ultimate financial responsibility. The State does have a responsibility to assure that cleanup is in compliance with State- established levels. In the event State, Tribal, or local governments expend resources for activities needed to mitigate the effects of the accident, these expenses would be reimbursable (see Subsection 0.2.3.6). Under Federal and State regulations, each State, Tribal, and local government is responsible for developing emergency-response plans and for providing the first response to emergencies involving radioactive material. As discussed in the subsequent subsections, assistance is available from the Federal government for planning for emergency preparedness and evaluating the adequacy of the plans. States have generic plans for responding to emergencies involving radioactive materials. These plans include procedures for notifying the organizations that can provide the required assistance and lists of organizations to call in order to initiate the proper response. There is no requirement for State, Tribal, and local governments to develop specific plans for responding to transportation accidents involving radioactive materials. The guidance document issued by the FEMA (FEMA, 1988) suggests that planning for transportation accidents be closely integrated into generic emergency operating plans for all types of disasters and emergencies. C.2.2.2 Evacuation Plans In a transportation accident, the State, Tribal, or local government has the responsibility for taking emergency protective actions, like evacuation, it should be noted, however, that a transportation accident involving radioactive materials, unlike an accident involving explosives or noxious gases, is not likely to require an evacuation in the ordinary sense. At most, in the unlikely event that some radioactive material is released, it I I \ I 0-5 would be necessary to establish a small control zone (with a radius of 150 feet from The Lource) f~m wWch people would be excluded until cleanup was completed. Federal agencies dearly have t^--ponsibinty to coordinat^^ emergency p.^aredness with °t*^^' i"'i=*^'74„7° '''y;„'pJS ,0 provinetioh makers at the State, ^::::Sr::Sl^^<^'^o,:^. . develop wntten procedures for making protective-action decisions, such as evacuations. .« nnp'c c;tfp trainina course presents the recommendations of the FEMA expressed in probabilities of developing cancer. C.2.2.3 Ca pabiHties equipment. Most first responders do not -i-in the capa«^^^^^ rwrer%rCommir o°n UJ OC III 1- <• o CO O z LU O CC 3 O CL n CC III > Ul T OC n z < O <• Ul o O < X Q T' Z) III n CC o CC Z o f) < O u. •t* 13 o T z O iii ~> CO U. <• oc <• <■ _i u. J. n < CD o w < O U UJ •^ c\i CO *" tn (O h- UJ O u. o 5 z Q CC O o o -I < UJ CM Z 6 2 o UJ D = O >. u. (D a, UJ z UJ u. O z UJ cc < Q. UJ Q (/> 3 C-10 For minor incidents, the DOE's response may be limited to advice given by telephone. The point-of-contact at the Regional Coordinating Office in whose area the accident occurred requests from the party reporting the accident essential information, including a telephone number where the first responders can be reached and a description of the accident. This information is then provided to a designated health physicist. The health physicist then calls the first responders at the scene of the accident and provides all advice necessary for mitigation, including recommendations to expand the response, if necessary. The Regional Coordinating Office also coordinates an exchange of information with the appropriate State and Tribal agency or agencies. When the caller asks for assistance in radiological monitoring or assessment, the Regional Coordinating Office coordinates with and receives approval from the State or Indian Tribe prior to dispatching a Radiological Assistance Program (RAP) team to the scene of the accident. This team consists of specialized personnel, such as health physicists, industrial hygienists, and medical specialists, chosen from DOE and contractor personnel. The size and composition of the team will depend on the severity of the accident. The mission of the team is to help State, Tribal, and local authorities identify and mitigate the radiological effects of the accident. Specific activities include identifying vehicles or property that is contaminated with radioactive materials, providing advice on decontamination, and arranging for medical advice on the treatment of personal injuries that may be complicated by exposure to radiation and/or contaminated with radioactive material. A designated spokesperson of the RAP team also coordinates with the local or State authorities to provide prompt information to the public about DOE shipments and the DOE's response assistance. In the event of a major emergency requiring response by several Federal agencies, the FRMAP is activated, and the activities of the RAP team are incorporated into the general Federal response. In such an event, the DOE's management and staff would initiate and maintain effective coordination of their radiological monitoring and assessment efforts with State and local agencies and Tribal governments. The DOE would provide all necessary resources to fully integrate Federal activities with the response efforts of the State, Tribal, and local authorities. It should be noted, however, that an emergency of such severity is not likely in transportation accidents involving radioactive materials. ^•^•^•2-3 Sequence of Events in an Emeroency Response . The basic activities of a DOE Regional Coordinating Office in response to a transportation accident are likely to proceed in the sequence given below. However, because each Regional Coordinating Office has its own response plans and procedures, some variations may occur. 1) The Regional Coordinating Office receives a call for assistance. 2) The appropriate State, Tribal, or local authorities are immediately notified to verify the request. 3) A health physicist may give advice over the telephone and determine the proper level of response. z a 2 A. I C-11 4) If the emergency requires emergency-response personnel or equipment, the Regional Coordinating Office will contact State, Tribal, and local authorities to determine their capabilities. If the State, Tribal, or local resources are adequate, the participation of the DOE is terminated unless additional assistance is specifically requested. However, if the DOE is the owner, shipper, or receiver of the shipment, the Regional Coordinating Office will respond automatically, 5) The Regional Coordinating Office notifies the Emergency Operations Center at DOE Headquarters in Washington, D.C., about the incident and the resources requested. If the Office needs additional support, such as the Atmospheric Release Advisory Capability, it will request DOE Headquarters to facilitate that request. 6) On arriving at the scene of the accident, the RAP team assesses the situation to determine whether additional assistance is needed. If an emergency requires additional resources, the leader of the RAP team contacts the Regional Coordinating Office, which requests the Emergency Operations Center in Washington to activate additional DOE resources. If no other assistance is required, the leader of the RAP team ensures that the response proceeds appropriately until it is terminated. 7) In the unlikely event that the resources needed for radiological monitoring assessment exceed those of the DOE, the Federal Radiological Monitoring and Assessment Plan will be activated. When this happens, the manager of the DOE'S Nevada Operations Office, (responsible for managing DOE resources during responses to major radiological emergencies), will select a director to coordinate monitoring and assessment assistance and to establish the liaison with the cognizant Federal agency (the shipper or owner) and State, Tribal, and local officials. 8) The appointed director selects a site near the incident to establish a Federal Radiological Monitoring and Assessment Center. The appropriate procedures from the Federal Radiological Monitoring and Assessment Plan are then executed until the emergency phase of the accident is over. 9) Once the initial emergency is over, the EPA assumes the DOE's duties of radiological monitoring and assessment. The time for this transfer will be determined by consultation among the DOE, the EPA, and the State or Indian Tribe. The EPA designates who assumes the DOE's responsibilities. C.2.3.2.4 Resources Available to Regional Coordinating Offices . Each of the Regional Coordinating Offices has a wide range of resources for responding to a transportation accident involving radioactive materials, including both personnel and equipment. These resources are drawn from the staffs and facilities of the DOE and the DOE contractors. C-12 The equipment available at most of the Offices includes the following: 1) Radiation monitors a. Alpha detectors b. Beta and gamma detectors c. Neutron detectors d. Tritium detectors 2) Whole-body dosimeters 3) Spectrometers (instruments capable of identifying specific radioisotopes) 4) Sampling equipment a. Air-sampling equipment for particulates and gases b. Environmental sampling equipment (plastic bags, etc.) 5) Decontamination equipment 6) Aerial-survey instruments 7) Protective clothing a. Gloves, boots, etc. b. Anticontamination clothing c. Breathing apparatus, including respirators and self-contained breathing apparatus 8) Dedicated response vehicles 9) Mobile laboratories 10) Electric power generators 11) Communications equipment (RAP radio frequencies). The personnel available for response include experts in health physics, medicine, security, legal counsel, public information, and industrial hygiene. ^•2-3.2.5 Other DOE Resources . In responding to a major radiological emergency, the Regional Coordinating Offices can request assistance from various other DOE resources. The magnitude of resources available is extensive. However, for scenarios considered credible for transportation accidents, only a portion of the DOE's full cadre of resources would be called upon. These resources, which are described in more detail in the above-cited report on the DOE's emergency preparedness (DOE, 1989b), include the following: ' Atmosphe ric Release Advisorv Capabilitv . This resource is operated by the Lawrence Livermore National Laboratory in Livermore, California. It provides estimates, using computer modeling techniques, of atmospheric diffusion, deposition of radioactive material on the ground, and radiation doses. z u C-13 . Aerial Measurement System . This system, based in Las Vegas, Nevada and Washington, D.C., consists of airplanes and helicopters with extensive equipment for radiation detection, data management, location mapping, and photography. It can be used for aerial monitoring to determine the extent of lost or diverted radioactive materials. • Mobile Accident Response Group . This unit consists of two trucks and two trailers designed to support a military response and can be transported by U.S. Air Force C-141 aircraft. One of the trailers is a personnel- decontamination unit equipped with a shower, sink, a 30-gallon hot-water tank, and anticontamination equipment and supplies, while the trucks carry an electric generator, a 250-gallon water tank, and a workshop. • Mobile Manipulator . The mobile manipulator is used as an emergency or standby system for toxic or radioactive environments. It is attached to a control console and can operate at a distance of up to 700 feet from the console. The mechanical hand on the manipulator can lift up to 160 pounds and drag up to 500 pounds. Two television cameras mounted behind the arm transmit pictures to monitors on the control console. This equipment is located at the Oak Ridge National Laboratory in Tennessee. . Radiation Emeraencv Assistance CenteryTraining Site . This facility in Oak Ridge, Tennessee provides the most modern multipurpose facilities available for handling victims of radiological emergencies and is designed to handle any type of incident involving exposure to radiation (see Subsection C.3.4.2). C.2.3.2.6 The TRANSCOM Vehicle-Tracking and Communication Svstem . As described in Appendices D and M, a satellite-based communications system will be used to track vehicles carrying TRU waste. Based in Oak Ridge, Tennessee, it has several features that can be useful during a transportation emergency. For example, the monitoring screens at the TRANSCOM Control Center will indicate the occurrence of an accident to an operator who is on duty 24 hours a day, 7 days a week. In addition, the system can be used to obtain information about the type of radioactive material carried in a shipment, it provides information from the Emeroencv Response Guidebook (DOT, 1987), and it provides a means for communication between the drivers of the vehicle involved in the accident and the Central Coordination Center at the WIPP. C.2.3.3 Guidance to State. Tribal, and Local Governments for Emergency Response to Transportation Accidents The Subcommittee on Transportation Accidents (Subsection C.2.3.1), of which the DOE is a member, has been charged with coordinating activities associated with transportation accidents involving radioactive materials. One of the major activities of this subcommittee has been to prepare emergency planning guidance for State, Tribal, and local governments so that they may safely and appropriately respond to a transportation accident involving radioactive material. The subcommittee has coordinated the development of a document entitled Guidance fo r Developing State and Local Radiological Emergency Response Plans and Prepa redness for Transportation C-14 Accidents (FEMA, 1988). This document, which is referred to as FEMA Rep-5, was initially released in 1983 and was revised in 1988. In addition to general information on transportation systems and casks, the document provides planning objectives and guidance. Included in the revised document is guidance for ensuring that State, Tribal, and local organizations have established procedures for contacting the proper emergency- response personnel, establishing methods for communicating to the general public when an accident occurs, ensuring the availability of means for limiting radiation exposures, making arrangements for medical services, providing for clean-up after the accident, and training. The document also describes the FEMA program for assisting States, Tribal, and local governments in their planning if they request assistance. C.2.3.4 Federal Emergency-Response Training Training in emergency response is offered by several Federal agencies, including the FEMA, the DOT, the EPA, and the DOE. Information on the training courses that are available is given in the Digest of Federal Training in Hazardous Materials (FEMA-134, Washington, D.C., July 1987), which includes a summary of Federal training courses for emergency response to accidents involving radioactive materials. (The digest can be obtained from the FEMA Publications Office, 500 C Street S.W., Washington, DC 20472.) The FEMA operates the National Emergency Training Center in Emmitsburg, Maryland. Training courses are offered at this center by the Emergency Management Institute. They address such topics as the assessment of radiological accidents, planning for radiological emergency-response teams. Information on the Emergency Management Institute and a schedule of courses can be obtained by writing to the FEMA National Emergency Training Center, Emmitsburg, MD 20727. In addition, the FEMA sponsors a radiological-emergency-response course at the Nevada Test Site. This course consists of 8-1/2 days of instruction on such topics as accident assessment and procedures for response. This course is targeted for individuals in State governments who must respond to radiological emergencies, including those initiated by transportation scenarios. The DOT supports the Transportation Safety Institute in Oklahoma City, Oklahoma. In addition, the DOT has recently published and distributed the 1 987 Emergency Response Guidebook: Guidebook for Hazardous Material Incidents (DOT/P-5800.4, Washington, D.C., 1987). The guidebook contains an inventory of hazardous materials, including radioactive materials, and a series of 76 one-page guides listing potential hazards and recommended emergency actions. It is intended to be carried, for immediate use, in every emergency-service vehicle (fire, police, first aid, civil defense) in the United States. Copies can be obtained by writing to the U.S. Department of Transportation, Research, and Special Programs Administration, Attention: DHM-51, Washington, D.C. 20590. The DOE has created the Radiation Emergency Assistance Center/Training Site (REAC/TS) at Oak Ridge, Tennessee. This multipurpose facility, operated by the Oak Ridge Associated Universities, is designed to treat victims of radiological accidents and z es C-15 to train medical and health-physics personnel. It is designed to handle any type of radiation-exposure accident that might occur at Oak Ridge or elsewhere (see Subsection C.3.4.2). The DOE'S Transportation Management Division sponsors a series of workshops on radiation-related emergency response. These one-day introductory courses cover basic emergency-response issues related to hazardous materials transportation incidents, with emphasis on accidents. Designed for regulatory and enforcement personnel as well as first responders to transportation incidents, the workshops cover four major topics: hazardous materials in general radioactive materials, shipments of radioactive materials, and response to incidents involving radioactive materials. The DOE has also instituted a special training program for the transportation of TRU waste to the WIPP. This program is discussed in Subsection C.3.4. 0.2.3.5 Federal Information Services for Radloloaical Emergencies The DOE operates, in conjunction with the Defense Nuclear Agency, the Joint Nuclear Accident Coordinating Center (JNACC). The purpose of the JNACC, which is headquartered at the Kirtland Air Force Base in Albuquerque, New Mexico, is to exchange and maintain information related to radiological-assistance capabilities within Federal government agencies and the military. The JNACC also functions as a point of coordination for requesting military assistance in connection with radiological accidents. The DOE also has eight regional centers of emergency-response experts to provide information and assist in responding to accidents. The teams are located in Upton, New York; Oak Ridge, Tennessee; Aiken, South Carolina; Albuquerque, New Mexico; Argonne, Illinois; Idaho Falls, Idaho; Oakland, California; and Richland, Washington. Information is also available from the National Response Center in Washington, D.C. This center is maintained by the DOT through the Coast Guard and in cooperation with the EPA. It provides information and advice to all interested parties for meeting emergencies involving spills of hazardous substances, including radioactive materials. The Chemical Manufactures Association maintains CHEMTREC, a similar information resource, also located in Washington, D.C. Both the National Response Center and CHEMTREC can be accessed using a toll free 800 telephone number, 24 hours per day. C.2.3.6 Financial Responsibility for Transportation Accidents To provide a high level of financial protection for the public in the event of a nuclear incident. Congress enacted the Price-Anderson Act, 42 USC 2014 and 2210 (Act). The Act provides a system of financial protection for public liability for a nuclear incident or a precautionary evacuation arising out of or in connection with DOE contractor activity by providing Government indemnity to pay claims up to approximately $7.3 billion per incident. (Certain NRC-licensed activities are also covered by the Price-Anderson system through insurance and a pooling of utility funds, but those provisions are not applicable to the WIPP.) C-16 In the event that claims exceed the statutory dollar limit, the President is required to submit a compensation plan to the Congress providing for prompt and full compensation for all valid claims, and Congress has promised to "take whatever action is determined to be necessary (including approval of appropriate compensation plans and appropriation of funds) to provide full and prompt compensation to the public for all public liability claims resulting from a disaster of such magnitude" (42 USC 2210 [e]). Price-Anderson coverage applies to all DOE fixed facilities shipping waste to the WIPP, the WIPP itself, and transportation to or from these covered facilities. All transportation modes are covered, and the protection applies not only to the named party in the indemnity agreement, but to any person (except DOE and NRC) who may be liable for public liability. In addition to the Price-Anderson coverage, all motor vehicles carrying TRU waste to the WIPP are required by the Motor Carrier Act of 1980, 42 USC 10927, and implementing regulations, 49 CFR 387, to maintain financial responsibility of at least $5 million, which would be available to cover public liability from a non-nuclear incident and for environmental restoration. z C-17 C.3 EMERGENCY-RESPONSE PLAN FOR WASTE TRANSPORTATION TO THE WIPP This subsection specifically addresses emergency preparedness for accidents occurring during the transportation of TRU waste to the WIPP. It outlines the general responsibilities, illustrates the responses that might be expected by describing a hypothetical accident scenario, and then gives detailed procedures to be followed by the various cognizant organizations or persons. In transportation accidents involving shipments of TRU waste, the responsibilities will be as follows: 1) The carrier will be responsible for notifying designated authorities of the accident (see Subsection C.3.1). 2) State, Tribal, and local authorities will be the first responders at the scene of the accident. They will have command and control authority for emergency response, and they will be responsible for implementing measures necessary to protect life, property, and the environment. 3) The DOE, as owner and shipper, will be present at the scene to assess the damage, to verify the level of any release of radioactive material or that no release of radioactive material has occurred, and to help the State and local authorities promptly inform the public about the situation. In the unlikely event that a release of radioactive material has occurred, the DOE or its contractors will collect the TRU waste and any debris; decontaminate soil, vehicles, and persons as needed; reload the TRU waste into new shipping containers; and return the site of the accident to normal use. These responsibilities are illustrated in Figure C.3.1 and outlined in the sections that follow, which discuss the procedures to be followed by the carrier; State, Tribal, and local governments; and the DOE. Each of the responsible parties must make various notifications of the accident. The organizations to be called by each party are cited in the text that follows, and the notifications that are to be made are summed up in Figure C.3.2. C.3.1 EMERGENCY-RESPONSE PROCEDURES FOR THE CARRIER The trucking contractor (the carrier) for the WIPP has prepared an emergency-response plan, including an itemized list of the emergency equipment carried on the vehicle, and has submitted it to the DOE for approval. The trucking contractor has provided the tractors transporting the TRU waste with equipment to be used in the event of a C-18 • RELOAD WASTE INTO NEW SHIPPING CONTAINER • RETURN INCIDENT SITE TO NORMAL USE RE-ENTRY AND RECOVERY * Q UJ 1- LU * Q UJ Q 0) LU -UJ U. CO ^1 O < l-CO UJ OLu < < UJ 2 O Z -1 ^^ mo a. NTAMI VEHIC LE, AS MITIGATIO 30NSEQUE >.DCLU UJQJp^ >CL y QC UJ X O -CL CCO X h- Od O ZDQC UJ < UJ O UJ COCL > o Q CO Q. • • • CO QC !i^Q UjSco IRST R N TRIE ORITIE. OF F INDIA AUTH CO ... TION TATE OCAi O^-J "5:^ z Q Q z DCO CD CC Z) Oh- z < o Q< HIPPI HAZ S AR( SITE H COO^ m^^ T^ ^E^ COO com zco J szcn CC m h- ^mLU - 1- Q DC m m m Q. O DC O < 15 z Q_ < h- Q m o CO D m CO O \- m CO O CO m DC CO QC Q. < • • • UJ X I- z LL o >- ^ ° -I o -< 5 2 < O UJ tii LU li- H CQ O X Q >- LU ^ o > Q UJ X I- m < CD o o LU X ^Z >- o en- -; Q < < LU N _ S z h: DC < c O C3 > Sx^ LU 13 Q CO < o ,-» "J O 3 _l o < CO 5 o o i= 5 ^ o < d < oc 2 LU Llj UJ W h- CC LU < I- X H X I- (A HI Q (/) ZOJ oz o- O I- UJ z o Q. o < UJ O Q UJ Z I- 0-19 CO < LU \- CL < a X LU O Q T_I < 05 Q LU CO - LU cr. O LU U. I _J < LU Q CO LU LU n-z 1 DC LU X LU < H o o cc LU O Q k ■^ I < CC z UJ o z g z Q CC O O o DC LU h- z LU O CL CL LU X _J Z < u ^ 1— LU 5< h- HjN < d HO CO — < z LU < Q X < DC DC 1 Q. Q. < i oc Ui UJ X I- I- CD O z O >< o LL pe (N fonso z ii •E o E - O QJ :2 z £ O O Ui >^ r3 (0 CI. CO n = QJ ^ < Q T, OJ CO o :3 01 CO CO CO (0 CO •D c •a o o 5 li 1 Z CO E 2 o o 5 o o o enne - aho (OK) Hall (ID) Mohave (N ickapoo (OK) aguna (NM) > Z en CO CJ) > O CD CD cn o vajo ( AZ) tawa (OK) 0) ©5 •o k_ >-Q. ^ .. o H— « 03 o XJ CD <1>(0 ^ ^ CO CO H— J2jf o CO sz x: <- o fO -J CO t < O Q. o OC Q. (3 Z o < (0 UJ CQ E < D-9 COLORADO Waste Isolation Pilot Plant (WIPP) REF: RAND McNALLY, 1987. 25 25 50 SCALE IN MILES FIGURE D.2.3 PROPOSED TRU WASTE TRUCK TRANSPORTATION ROUTES IN NEW MEXICO D-10 gi L q§ <2 tm OI H Q m z h-O O LU O _IUJ LU H CO W LL-J Ouj g5 O CO &i^ j:d 3O tLCC 0^ UJ30 icn< Hl-D- lli 1- 2 DC LU (/>0 LJ QC HO go C/)uj zi o c o o c a> E w ■G 03 O d LU _l CO c o o w 0) O Q. CO i— O) o O o Q. 0) |2 T3 C O ID in O) c "55 CO o CO ^ CO ■D u. Q.T3 ^ >, 0) 3 O c o O) T3 CO CO 9- .9 0) _ > nr o ^ c 0) > 0) CO -Q IS ■g's Q. CO Q. 0) < 5 CO c CO CO c b CO c b c CD CO I CD CO z CD CO c CO SI o _c E O > I n O) X O) c CO CO c CO b c c (O x: i_ ^ S c CO C E 3 o > SI sz 9 0- C CO — X CO o JC B 2 2r CO c CO •- b S- (0 c 7^ ■^^—^ 5 (B 1 ^~ CO T3 c C 3 Q s c > 6 CO <= CO s>s ^ betw Coo co" Q. *- T3 E ^■■^ c C E O) CO >% CO n CD CO a> V N. $ CX> ■"^ T— ^9 0> CVJ CX> ^— t^ c C3 T— c s 0" ^1 ^ ^ 1 1 CO CO CO CD a c CO c c C CO b CO b CO b c c c c c <3 z p N. 1 N N. t T3 oc IB ■D C CO = CO ^^ CO CO .^ k. TO o CO ai CO c JO n c 3 o n to CO 3 O Ll_ CO c o o to CO CO c CO CO CO c *- CO 4-' CO CO .^ CO .^ 3 k- 3 k— -3 < -3 < in CO CO CO CO CO CO c c CO CO .^ ^ k_ k. ^ ^ T t D-17 0) o c o o c o u (0 Q •^ ^ CO fi s s. u Q o « 4-' to c b: a> «^ E 0) 0) N > 0) (0 0) Q. i: <-> C c 0) u CO s CO CD T3 1 0)" 3 O T3 to CD CO c o 'S "O c o u c CO o> c 1 ■D >. c c E a Q. 3 « o (0 C ffi 3 k» £i O) C c o o £ ^ n 0) CO ■D T3 C c 3 3 2 ^ (t% p 0) 3 O To 3 2 dj c o ■o o 3 c c o o d LU _l < C o o <0 0) •o o a (0 o d) O T3 0) 3 C O o c o c (0 CO "58 %'^ CD ^ II o o II X (0 E (0 a. H __;. CO « CO o o E tr 3 TO O O c o = o CO o h- "x CO ■x «** 0) ^ s B IE ^ ^ to ^ O (1) CO Q) ^z LU 0) c IS 0) ato < 5 2 85 >'=6 0) -c 5. Q. u Q. Q. CO IB W CO CO s Q. t?) ■^ to o a 0) o o TO 0) CO c o "5 Q O 0) c c I- 2? o o CO c _o (0 z "O (0 o 00 in CO CO j3 -^ 0) o CM CO o o o CM t T- 1 co S CO c 1 3 '§ '0' 0) C2. "i g if a s S (0 ■^ ^ ^ k. '^ CD CD 0) z_ z > 5 cc 2 t ID c c fi o o a. "co CO 'co CO o CM D-18 0) 3 C *-» C o o CM d LU _J CD 0) o c o o c o o w Q o w O Q. CO o a> O i a N ~ O w ■- 03 0) O CO 5 c o CJ) 0) is o •D c m r: o ir E o x: o CO c o CD C ^ ill ^ ^ J3 C E >- c (0 0) o c CO D o ■D C CD I — CO "5 5 « ^ c o ■E o .« O OiXi 20 8 c o O) o O o '^ CD •^^ C O o c o T3 C 0) Q. ■D ^ C C CD O •e Q. O B li. CO c o 0) 0) >,^ o ■D C CO CD E a. X 1^, o o k. a to Q. 03 < O cu o «> CO l.i c .« CO "D .. c CJ) O c o :> CD *i c c CO'O 8h c o >^ C CD O . c CD o o Q-CD iS CD E O) c •> CD CD CD^ ^ O o *- O CO *-- o CD 3 5 i= C 0) CD W i ^ > o ^ E o X c CD ■D O Horn Idaho n "cD c . n Idaho ill, Idaho E CD O 2 u. o o 2 « ^ CD •ft c 2-1 CD O 0) CO T3 a> Id CO O at East to G CO "D O O c CO 2 Q) <0 c c CD O cniti <- "^ I & ^ O C 0) id T3 CD ■D O) O $ CD ^ O ^ 3 -J .£ CO CD CD 0)"0 c 03 k- t? JZ o 03 C 3 CD CO O o f CD ■D o o o o in J3 CO o cxi o o 8 in o o q CM CO in 1^ o CO CM CM CM O) T— CM a , 1 1 1 1 1 0) ii C o c (O CO C^i o o o o o CO CM 8 00 CO CM 1 o o d o o d in o o d o o o ^' CM 2 c s ^^ ^_^ ^_^ o c c c 'B TJ 0) CO 1 o O o O o Ui 2 O 2 o" ^ p x: 03 x: 03 2 1 Q> CC CVJ 00 1 s s S s s s CO 0} "O — — 1 ^1 _^ _1 1 CO w c: o B •£ CD CO bic o C — O) > g.e ^ J8 " D CO •= O -C ^ CO > U. O T3 O |>^ C ^ o o o 00 o d 00 ^— • o o 1 00 o CO o ■^ 00 CD sz CO s CO o oo s o 00 CO N •D c U c > o 5 crt «2 0> CD 5 o « -c 2 o ^ »- o 0" CO _ ~ Q. .9 CA iS 3 ? "5 o 3 n c O n O) ■o «^ c O ^ 1 o -I CO o to 5 c O) c CO r: o CO o E c c CO 1=: c 4-^ 3 o" 00 1 c CO D c o 00 o o 00 o o CO 00 CO o 00 CO c o E c .. o CO o ?^ •> CO ?2 2 b) CO ■D c CO CO ■D CD o 2 T3 C CO CO o If ^ o ii CD CC cvi T- CM O) CO CO ,o > CO CO c ■^ o ^^ CO o CO CO o^a ■" "D CO 5 O O) ii CO o S 5 c CO T3 -a •i « CO ^ O O) CO c CO D O (0 CO c c ■500 CO o •;= 2 -D >- CO o i^ n o O) O) D) c c c g '^ £ 1 1 1 «? ^ «? CO c- O .ti CO E o V> CO D o ■a CO .0 TD C CO CO T5 CO N N C D O 5^2 X O) E >, ^ ^ ^^ CO" ii c CO 3 E CO ;|' c ^>' 'cO 3 2 9- ^^^ thern W Collins, Q. 3 e ^ CO x: LU X CO li- 00 a ■* <5 CJ> cvi in r«. CO CO CM 1 in Oi 00 in CO 0? CO CM OJ CM CO o T3 CO O o o in CM D-20 0) o c o o c o o CO o Q Q) C CO 0) re 3 co-Q 5^ >^ S 2 o t 5 o o CO .'^ c o X w o c/j 5 re o CO c g T3 C o o o D o CO 0) I o CO ^ a I ^ re c ^ CO ;= o 3 E CO C S .05 3 x: ^ CO c g ■o c o o g 3 o CO DC ■D Q) 3 C C o o d LU _J C o o CO CD ■D g ir a re o CD O 0) ? CD OQ o CD •fc ■g li. o CO 2 o 5 r re sz c CD (D c o o o •c o P > re c XJ T3 a O (D ft and the Alar University A near S orado miles f Colora re 1— o o O CO o ■D 2 Q s o ■a 2 o ay viaduct •enver, Coli Hill, 17 prings, co" c *^ Q. o o o o 2 o X ^^ o O SI O go CO o X CD o o O re a 5 8th St xit SOL c CD roadw r. in D §1 c o O o T3 re o o CO 3 .o T3 2 o o o" •5 < CO CD Q m Q 2 O o in O CO o a ■D 3 _C V- • c o c o CVJ CM CVJ rv" rsl N.* (j> o 0) 1 ■^ T" !^ ^ Rl ^ •»- CO in N.' y-^ CO U) t o <8 3 O J o re CO c *^ O ^ _ c o S: (0 o c (0 X o re o o o in o ■D 2 o o o in o re o o o in o ■o 2 o o O in CM o o •D T3 re re i_ 1^ o o o o o O in in OJ CM o ■o re i_ o o o in CM D-21 ?^S^SK O C o o c o o (0 0) Q •o D C c o o CM T3 d O UJ JZ _J a. CO J? ;< ^ I— Q o c o o w 0) o 0) o o 0) CO c 0) 0) Q ■o 0) _c c o o o B> c (0 (8 c o (0 (A a> T3 o c (Q z ^^ E o .•-I *- CD ^ I. so ^ ^ 8 ^ 0) 0) 0) c Q. 5-5 "^ g-T3 ^ c 0) .2> 3 J3 ^ o o o ■o m o o o in CM 0) tS 2 S" CO o CO 13 Cl> n (/) >» o (0 O o E CO T3 TJ C C/) o CO c E O s to to c £ ^ o o C/) 0) CO CD T CA O CO iri o CO o o o in CM 3 CO o -^ O -D ■_ C CO 5 CO CO ±i %9 ±i 5.E o o 1 o o 9, in CM CO O o T3 >< C CO o E " CO- ® £ f- .5 CO CO 3 c CO o 0. ^ c p C O o O ^ in O o o8 Cvl c S ado Fuel lorado ortlon of 1 olorado % •E o o z Q.O in CO r Co bio, CO T3 Q. c o CO ^ c "♦3 *i D .^ c CO Z Q. LU H LU cc o CO c:. D o E CO Q. > o T3 C CO CO O o o I CO o CO a o Is !i c o is c 5 Q. D CO o CO £ I c (0 CO ^ o o x c o c o C3) TO O ^ .9 ^ CO CO O) > CO (0 o o 'x -s 5 5 o O 0) O ^ 5 ? z z Z Z w ■^-^ in in in \P. CNl CM CM CM o o x 5 in CM D-22 0) o c o o B c o o (0 Q •D 0) C C o o C\( T3 d O LU ic -I Q. OQ CO f o 0) O c o o o a 0) o o (0 «^ 0) (0 c 0) Q ? en 0) If (0 Q. C P (C o 5 o .ti c ■D o 6 1 g'" Q.^ in 3 -h 00 en ^ ^ (0 (0 r Q. .9 CO II O TD 3 O S c ^ 8 1 8 CO c o B> c £ CO CO ^.-.^ ^ o o c x 0) o s ♦* s 1 0) z (0 a> in CVI ■o — w o H- C (0 X v> § N JZ I I o).g ■D O) Q) ^ If .^ >^^ CO i5 CO = 2 •> X (0 0) .ti o o 'x CD IE Z (0 £ (0 CO I o /•9 >< O CD co2 i I o z 0) CO c $" (0 c c3 § w a> .^ £ JC ^ o m — r Q. Z ^ O) o c c o CO o o co2 "5 CO x: o o o CO O o o CQ O > o O CD CO z o X c 2 E ^ o CO Z ^z - X ^ OT 2 1 O CC o »•— OJ o ■o "h c CO CO c CO o T3 c ■c > o Q. ^ O) CO ■o !c CO o 4^ s c ti n _co 5 o c CO 6 3 B CO .CO £ co CO ■Q > CO o CO CO LL o ^ X3 CO o o o cc o O E o in" ^. fV •^^ c o 4-» E *i c:) C CO CO Q. o CJO in CM in CO r>* IsJ g I CO i in CO In 1 CVJ CM CM § ^^ ^ O 'o' 'o' 'o' 'o' o ■a O o "x o 'x O X o x T3 S s s s s O 1 1 1 1 o O 2 o Z Z z. z z .. o ^^ c o in in in in in (0 SR CO CO 00 CO Q. W 6o" CO CM 1 CM ( CM «V ^ CM op CO CO 0) T3 O Z a 03 i_ o O « S ° x: 03 Qj ♦^ .»» sc c 1^ d TO ^ (A - Q) T3 05 O CC <0 0) CO x: _ E ^ ^6 5 O 0) c OJ (/) o CD OJ O 5 D O CO O) C CO TO (0 c 03 o E E o D CO o o 0) c c _ D < :C C ^ S) o §18 CM <3) > "5 03 03 Si o 1- ^ o t E ,_ O o "D ^ 2 Q> T3 CO •> x: CO o _ T3 1 = ■S c > E ' to c o 0) CO C O) 3 t- co o ■o 2 O) Q. o CO 03 O ^ 0) *i CO 4-* © CO D) C "S 03 H ^ o 4-^ O) d) c E 2 UJ ^ o o o 3 2 cd ♦i o CO O Q. 2r CO o to o o ^ o cvj CM -1 O) c k. ,— 1^ 0) o 0) x: c 03 0) 4-^ o> P D c ^^ o UJ O U. CM S" c CO CM 'o o 03 ^ X «•— (0 CO 0) z CO c o £ Q CO T5 E sz O 03 »_ *- *- 3 *-' c o o *- E o <="■£ 03 CO p in - T B — o k. Q> S CD .i= o T3 SiS O LU CO 13 r: 03 c o C o E 0) x: o E Q. O O d o o CM 00 CO in o CO o x: (0 x: (0 D c o 8 CO O E d O ^ >, >« >. 03 (5 X3 E 3 o c;) C co' 1 (8 03 ^ Q- a. o c CO d>£i o c o o (1) T3 s c o s •a c o 0) OS CO o sz * o 1 o ^ T3 03 4-' -3 c « O) CO o o CO Q a CO o o C "d . § -3 03 c ^ ^ CO O CO CO •ar> 1- C co" CO O D 03 *- o- CL o 5 00 o C in -3 o ^ — c Altam Coun From in Sa « c 1 •B 75 O o 2r CM CO o o <8 00 O ''t ia yS CO -1 ^^ <0 c o ^^ CO _ z 03 0> to" o o E 0) o o 03 > -1 § O 0) in in u — ^ c a> w s «» D-24 0) o c o o c o o Q i 0) 3 E w 3 o •2 (0 2 w ^ o ^ »- (0 c 2 '2 ? O OT 0) TO 3 O c I ■D a 0) 55 .9 w c o o CT3 ■•G 3 a> in E CO o) 3 0) 05 4-^ T~ O o C _ m al 0) CO 1- •So i 5J t3 c CO E CO x: to ■g ■^o Q. CO I o cc o o c "5 C 5 o C iS CO 3 ^1 «5 cj tr CO ■D CD 3 C **^ C o o CM d LU _l CO c o o CO 0) T3 O !c a CO i_ O) o o CD to o Q. Qi B 2 f o 0) CO c Q 8 g -i ^ c CD CO E CO 'c CO O) CD (D c CQ O) CQ C £ C CO (0 o CO o 2 CD CO 1 CO of "o 4-' c 2 in CO E o CO (/) CO o ■D 3 O ^ ^ CO CD Qi o Q. c c 'co t3 > 8 CD 2 C o o CO 3 o CO CD ^^ u- < 1 O 2 Z IB C ki O :te « O o •*- o> j: ._ CO w o n 2 'c? (B c C o o CO O "co" E CO CO z o o CO "cO" c Z o ' o^ 1- t- E o' in i ^g :2 0> y— CO CO (0 > C\j 1 o O (0 0) CO £ CO CO CO O) CD > CO CO u c CO in in in CO a> CO 0) CO ■D CO > CD 2 in O) CO CO CO c — c <^ s o^ CO '• o CO "D ^^ c o o *^ ~ CO t3 CO C O) ■D O CO X CO $ ig r^ o CO t^ ^ 1 y— 1^" CO in 1 ^s CO CO ^2 CO s , ^ o CO o T3 CD ■^ 2 2 in c:) 1 2 CO o D t T3 C CD CO C CO O c CD CD CD CO o o x CD CO c 3 O) CO o o 6 CO o o x CD CD O t D-25 0) o c o u c o Q (0 ^ p CO (0 (0 .^to 0) o = (0 0) <0 d snow and ons, steep 1 1 i 3 O ■D O a $ o o- o poor visibil osed for we a 3 c , accident a ttle over are Wind, rapi( accumulati T3 C 1 Icy, winds, may be cl( O c 5 Bad curve fog will se T3 3 C c o o CM d LU _J CO < c o o « T3 U !c a CO o a> O (0 c 3 O) o 05 O -" X to 5 ^ z o o x 0) 5 0) (0 c 3 O) CO CO z o g ■x CD 5 0) CO 0) k_ CO CO c 3 D) CO CO CD CO CO c 3 O) CO o o o o 'x x O 3 3 I (0 > > z ^^ CD 3 C 5 z < 3 cr — o >!=: X 0) CO -= CO > Is CO O" O 03 co 2^ k. 3 i= < •D O O $ CD ■o LU o o ^2 ,55 o 2 T3 C o o 'x v to ^ ^ z 0) o o CO o CA c o Q 0) T3 3 O c g n •a CO 0> £ (0 (0 ■D (0 > s in ;^ o o 1 I z o 00 S. S< CM a o 00 00 1 CO ^ CO o s? CM h" N o o o o o o o o "x 'x "x x 0) 0) o 0) s s s :s ^ ^ ^ ^ o Q) Q) o z z z z o o o o t t t t o o 1 I o s o o o o o o "x "x 'x s o 0) s :s s ^ ^ ^ CD Q) Q) z z z o o Q t t t D-26 a> o c o o C\J d LU _l CO < t -D g 5 o ^ E Q. 0) £ 0) (Q o S as (1> r= O *~ C LU C 05 CO c o o sz o 0) (0 0) 0) o 0) c5 0) T3 re 0) O CO CO .18 - c ^ 2 -D S w .N 3 < O O -D CO ^ ?S CO CO CO 0) £: B o CO ■D CD CO E CO LU C 3 O 0) u c 0) _3 c o o CO E o JO o o O) ^. o E CO B o c o 0) CO c o o ■o o CO \— -Q r5 O _| 0) o ■4-* *rr C CO "I O O) lo ."S ^^ *- o 8^ c ~ 8§ CO ^ c O CD CO On ,'K « CO _l ex. x: (0 'c i (0 O c O O (D 2 ^ o o n CO _j c '^. CD CD C c LU "cO c o o CO E o »^- CD B o o CO to o a o c 3 O O (0 D re CO 'c re CO CO o cx5 CM CO o Q. CO "c o re o c 3 O O CO 0) CD c < CO o c J c\i CO lO CO Cfl O Q. re c re 'c k. o re O a? c 3 O O O c re c ^. CD CD C t re i ? re O c 3 O o o c ■o re c CD OQ C re CO E o li- LL S E o c o lO C35 CT) *^ CO o Q. CD ^ in CO o Q. z UJ O UJ CC O cr o u. (0 HI I- D O OC Q < o oc -I < oc UJ z -I z < o UJ (0 o a O oc Q. D-32 Reporting the status of the TRUPACT-II containers and NuPac 72B casks to the Central Coordination Center Loading TRU waste into TRUPACT-II containers and NuPac 72B casks Meeting DOT and RCRA shipping paper requirements Dispatching loaded TRUPACT-II containers and NuPac 72B casks Notifying the Central Coordination Center of shipments Following on-site emergency response procedures for TRU waste loading accidents. The trucking contractor will be responsible for the actual physical movement of the TRUPACT-II containers and NuPac 72B casks between the waste sites and the WIPP. The contractor will provide a dedicated tractor fleet, dedicated drivers, and a dedicated manager for this contract. The responsibilities of the contractor are outlined in the summary of the management plan in Appendix M. The DOE will be responsible for the following transportation tasks: Interfacing on institutional issues with other Federal, State, and local agencies in conjunction with TRU waste facilities and local DOE field offices Coordinating with the waste facilities Planning TRU waste transportation Translating DOE policies into operating procedures Establishing and operating the Central Coordination Center Administering the contract of the trucking contractor Budgeting transportation operations Procuring transport packaging and trailers with placard holders Scheduling shipments in coordination with the traffic managers at the waste facilities Receiving shipments Maintaining communications equipment Complying with procedures and reporting requirements D-33 • Reporting routine activities and nonroutine incidents to appropriate autiiorities • Monitoring and evaluating the performance of the trucking contractor. D.2.3.2 Preoperational Checkout Before shipment of TRU waste, as part of an overall integrated operations demonstration, multiple dry runs from each waste facility to the WIPP will be conducted as a part of a series of preoperational checks designed to provide experience and hands-on training for the drivers of the trucking contractor and the operations personnel of the waste facility and the WIPP. A summary of the preoperational checkout plan is provided here to describe the types of testing and training procedures used by the WIPP, the waste facilities, affected States, and the trucking contractor. The checkout will provide a review of the completeness of the facility readiness review procedures, will determine the adequacy of facility readiness, and will allow the review process to track incomplete items to closure. The checkout is designated to: • Validate the facility's ability to load and ship a TRUPACT container • Provide experience in using the TRANSCOM tracking and communication system • Evaluate the responsibilities of States and Indian Tribes • Evaluate the procedures for waste receipt and emplacement at the WIPP. The intent of these dry runs is to incorporate as many realistic conditions and procedural checks as possible into a training exercise and to incorporate any changes into the existing procedures before actual shipment. At least two dry-run preoperational checkouts will be conducted at each facility before any actual shipments. If requested by appropriate authorities, additional dry-run preoperational checkouts will be scheduled to ensure readiness of all participants for actual shipments. It is expected that the products of the preoperational checkouts would include: • Final shipment procedures for waste facilities and the WIPP, including the WIPP Waste Information System • Final procedures for interactions with States and Indian Tribes regarding TRU waste shipments • Final procedures for TRU waste receipt, unloading, and emplacement • Driver training and familiarization with the preferred routes • Operational readiness reviews for each waste facility confirming readiness to ship TRU waste. D-34 A typical dry run will begin with the receipt of the empty TRUPACT-II container at the waste facility and end with receipt, unloading, and emplacement at the WIPP. The latter will be done at the discretion of the WIPP waste-handling operations manager. There is no mandatory requirement for the underground emplacement of drums for every checkout. During each dry run, various scenarios for en route events will be initiated by WIPP personnel or by the driver to test systems on the truck or at the WIPP. The locations of each event will be modified for each preoperational checkout to fit the participating waste facility. The dry runs will be tracked with the TRANSCOM system and monitored by WIPP personnel at the CCC; digital communication will be established with the driver on a periodic basis, following established TRANSCOM procedures. As a minimum, on the return trip, drivers will input simulated "shipment problems" via the TRANSCOM to test the CCC operator responsiveness. These may include, but are not limited to, mechanical problems, protestors, sabotage, vehicle accidents, severe weather conditions, or the need to deviate from the preferred route. The CCC operator, following approved procedures, will provide the appropriate direction. On at least one occasion, the operator will ignore a message from the driver to verify that the Trans- portation Control Center in Oak Ridge, Tennessee, is monitoring the shipment. Dry runs provide an opportunity to test various shipment scenarios. Data obtained regarding travel times to and from each facility will be used to establish a baseline for future shipments. All routes used during the dry runs will be those contained in the DOE-approved trucking management plan. On occasion, the driver will be instructed to deviate from these routes to test the alertness of the shipment monitoring agencies. Summaries of various dry-run test scenarios are provided below with the expected response. Those summaries marked with an asterisk were used on dry runs in January and June 1989. These dry runs used an "engineering model" of the TRUPACT-II on a prototype WIPP trailer. These initial dry runs were made to determine shipment time, and to give the driver experience in using the TRANSCOM keyboard, in interacting with the TRANSCOM operator, in using the mobile phone, in using the KAVOURAS weather forecast system, and in responding to a variety of simulated accident scenarios. ^1) Evaluator-induced scenario : National weather channel indicates severe storm approaching the shipper's area. KAVOURAS system indicates temperatures below zero and 1 5-mph winds. The operator contacts the facility traffic manager and Transportation Operations personnel to make a coordinated decision of appropriate action. The trucking contractor should be notified of delay if not alerted by driver. ^2) Evaluator-induced scenario : TRANSCOM. No communication capability with driver through The operator attempts to call the driver via the mobile phone. Instructs driver to call in every 2 hours or when crossing a State border. The operator provides the Transportation Control Center with location provided by driver for manual input to TRANSCOM. D-35 3) ^4) 5) ^6) ^7) ^8) Driver-induced scenario : Tractor placed out of service because of excessive play on the right front axle. Vehicle cannot be repaired locally and must be replaced. The gross vehicle weight at weigh station was 79,748 pounds. The driver will notify the Transportation Control Center, and secure approval for the Proposed Action. The operator should notify the trucking contractor of replacement requirement, as well as the receiver (WIPP Transportation Operations) and the shipper. Weight was specified, as it will require a special tractor not to exceed the 80,000-pound limit. Operator should be aware of weight limitations. Driver-induced scenario : Estimated 2-hour delay. Broken radiator hose. Driver can arrange repair. The operator will notify the trucking contractor and WIPP Transportation Operations. Driver-induced scenario : Protesters harassing shipment. Path blocked by protestor vehicles. Carrier tractor damaged by thrown objects. Demonstrators becoming more and more violent. The operator notifies WIPP Transportation Operations, the waste facility, local law enforcement agency, and trucking contractor. Tractor replacement may be required. The operator stays in contact with driver. Evaluator-induced scenario : Information provided by the State Highway Patrol: on the downhill slope of the pass, the tractor brakes failed; the driver attempted to keep control but the vehicle overturned. All three TRUPACT-II containers have broken loose and are scattered within 100 yards of the trailer. The drivers have been seriously injured. Not known whether there was any spread of contamination. No further information available at this time. The operator follows the notification plan given in Appendix C. Evaluator-induced scenario : Two vehicle accident. Collision between carrier vehicle and auto which entered interstate from on ramp, cutting off tractor-trailer. The auto was totalled. The tractor driver was injured seriously. TRUPACT-II containers are undamaged. The tractor is inoperable (right front fender and frame crushed). Damage to car~$1 2,000; to tractor-$7,000. Local authorities at the scene; the ambulance has departed. The operator notifies the WIPP Project Office, WIPP Transportation Operations, trucking contractor, and the facility traffic manager. The trucking contractor will arrange for replacement tractor and driver replacement. Driver-induced scenario : 100-mile check shows broken U-bolt in the third rearmost container, right rear corner. D-36 The operator notifies the WIPP Transportation Operations, which arranges for installation of a replacement by a qualified individual. Appropriate staff at the WIPP Project Office would be notified of the event. 9) An evaluator-induced scenario that is yet to be used is as follows: At some point while a dry run shipment is traversing a State, the State police and highway patrol will be notified that TRANSCOM contact with the shipment has been lost and their assistance is requested in locating the vehicle. The State will use its resources to locate the vehicle and pull it over. Once located, the driver will contact the Central Coordination Center and notify the operator of his location. The State police or highway patrol representative will also notify his headquarters that the vehicle has been located, and they, in turn, will notify the CCC operator. This will exercise both lines of communication. This may be implemented in each State the vehicle passes through. D.2.4 VEHICLE TRACKING SYSTEM The CCC at the WIPP will use the Transportation Tracking and Communication System (TRANSCOM) to track TRU waste shipments. This system is operated by the DOE's Oak Ridge Operations Office and is linked to the WIPP at the CCC via a dedicated telephone line. TRANSCOM will use a land-based LORAN-C positioning system to obtain longitude/latitude information. This information is calculated by a LORAN-C receiver and transporter antenna attached to the trailer. Signals will be transmitted via satellite to a commercial ground station and then to the TRANSCOM Control Center (TOO). The satellite communications system allows digital communication between the driver and the CCC at the WIPP. The CCC is able to communicate directly with the en route driver by mobile telephone. The TCC will provide access to the tracking system to those Indian Tribes, States, and facilities that need to monitor TRU waste shipments. The location of the tracked vehicle will be monitored by the CCC so as to detect any deviation from the preferred route. Frequency of detection is limited by the frequency of vehicle location transmissions to the TCC. For TRU waste shipments, the frequency will be approximately every 15 minutes. In New Mexico, as elsewhere, the officials of the State and the Indian Tribes will also have access to limited functions of TRANSCOM. The appropriate software training will be provided to enable them to receive data regarding TRU waste shipments passing through their jurisdictions. Integrated with the TRU waste shipment system will be a set of activities that function to deter, protect, detect, and respond to unauthorized possession, use, or sabotage of TRU waste shipments. These activities will include: 1) Close, continued surveillance of the en route shipment by means of the TRANSCOM vehicle tracking and two-way communications system. 2) Efforts to minimize intermediate stops for each shipment. D-37 3) Constant surveillance of the vehicle and cargo during transit. One of the drivers in the two-person truck crew will remain with the unit at all times, including refueling, food, and relief stops. A vehicle will be considered to be under surveillance when one driver is in the vehicle, awake, and not in the sleeper berth, or is within 100 feet of the vehicle and has the vehicle within an unobstructed field of view. 4) Use of a tamper-proof fifth wheel locking device. 5) The use of an escort vehicle would be a decision made by the appropriate State agency, with due consideration for DOT regulations. The DOE does not plan to use any escorts because with real-time tracking of shipments, accident situations would be identified and communications with the vehicle would take place almost immediately. D-38 D.3 TRANSPORTATION RISKS D.3.1 INTRODUCTION This section presents an analysis of the risl E E 0) .^ coco O ^^ II CO 15 ^ iZ V) 00 O 03 t; 1- n CO S - "S S Q C S CM O) TO ^ g CO CO Q 00 -D ^ 1— S.fO^ fc CO o £, O) C\J (/) O 03 ■D "O C 03 X O CD C O O (0r^(D<£)CJ^CvJ in ininM-t'j-cococj'tco OOOOOOOOOO X X X X X X X N- C3) O CJ) 1- N- CJ 1- N. CO ■»-' cvj CO r^ iri CD c\j ^ c\i ^' CD cj) CO o in T- 1- C^ CD CJ5 OOOOOOOOOO xxxxxxxxxx oo'^-.-r^.cocjcooc^i-t- ocDcqcDin-T-T-coino T-^ CM CO n! iri cd" c\i CO c\i t' 00 (o in in m "^^ oco m ■<*■ OOOOOOOOO X X X X X X OJ ■tt 00 CD OJ o CT) O CO T- O CM C3i ovi in CO CO '^ XXX o c:^ 1- 00 in o CO cvi •^ o ^"b'b o^^ ° ° ° X X X XXX o CO in c:^ T- ■^ CM CO \Si O T- ■r- CM CD CM 1- CM lU o to o 03 c ^ 2: c ^ LU C — i5 TO Q. g w CO tr CO (U CO k_ CD > CC sz 03 c c 03 > CO CO 5^ =-0 03 -Q ;-. _l 03 CO CD '■!-• _l c" Z O ^ o to ^ 03 03 UJ -I ill 03 ^i^- 2 E o o 0) 2 O CD ,<1> ^ CJ)I— 03 !2 03 < °= "S ^ CD 03 CD 6 O 2 03 2 CD c c »= 03 o n O £> ^ "I- Tt in CO ro O O O O O n O T— ■»— T- ■>— T- ■>— X X X X X X frS in in c:o oo S r^ 'I- CM ^ o in ^ t^ T- a> (X) •^ -^ -^ rr> n o O O o o O 1— 1— T— T— T— ■»— X X X X X X CO O N. CD N- •^ CO O) t^ in CO "l- * CO CO O O O O O X X X X X O t O O ■.- CO en in in CM in CM in co" -r^ (O "S- CO CO CM O O O O O X X X X X CM CD 't O -r- 00 CO in in CO ^ CM CJ) CO T^ O CM «*■ O O O Ooo o ■>— 1— T- ■»— XXX X in 00 in 00 '^r in CM O en CO Tf t- CM "I- O O O CO ^ 03 ?-> ^^ o 2 ^^tD C3) c 2 o o 0) o c 'en Labc bora Lab C o 03 — ' c c c -= o •D LJJ CD o CO -y CD ^ ■^ 5 ^ "D Td 0) 03 O 2 C 03 I CD O c o CD 2 o o CD o X3 CO CD cr ■D O Ridge N )nne Nat Alamos _l < E n 03 c V "-J p CD 03 _l CD o oc "O X O < Jj H D-43 TABLE D.3.2 Projected number of CH TRU and RH TRU waste shipments from generator and storage facilities to the WIPP Facility Number of shipments 100% Truck Maximum rail Contact-Handled"''^ Idaho National Engineering Laboratory Rocky Flats Plant Hanford Reservation Savannah River Site Los Alamos National Laboratory Oak Ridge National Laboratory Nevada Test Site Argonne National Laboratory-East Lawrence Livermore National Laboratory Mound Laboratory TOTAL Remote-Handled'^ 4046 7608 3103 2640 2065 228 80 14 969 150 20903 2023 3804 1552 1320 2065*^ 114 SO*' 7 485 75 11525 Idaho National Engineering Laboratory Hanford Reservation Los Alamos National Laboratory Oak Ridge National Laboratory Argonne National Laboratory-East TOTAL 487 2470 101 4605 300 7963 244 1235 101 '^ 2303 150 4033 ® Shipments based on 3 TRUPACT-lls per truck shipment and 6 TRUPACT-lls per railcar shipment. ^ Truck shipments calculated from a drum volume of 0.2 m^/drum x 14 drums/TRUPACT-lls x 3 TRUPACT-lls/Truck. Rail shipments from a drum volume of 0.2 m^/drum x 14 drums/TRUPACT-lls x 6 TRUPACT-lls /Railcar. ° Los Alamos National Laboratory and Nevada Test Site do not have access to rail, thus truck shipments are included in the maximum rail case. ^ Truck shipments calculated from a NuPac 72B volume of 0.89 m^/NuPac 72B x 1 NuPac 72B/Truck. Rail shipments calculated from a NuPac 728 volume of 0.89 m^/NuPac 728 x 2 NuPac 72B/Railcar. D-44 per shipment. Using the waste volumes presented in the 1987 Integrated Data Base, and the information on waste characteristics provided by the facilities, the radioactivity characteristics of average truck or rail shipments of TRU waste from each of the sites were determined and are shown in Table D.3.3 for CH TRU waste and Table D.3.4 for RH TRU waste. Site-specific values of the Transport Index (Tl) for a typical shipment of CH and RH TRU waste were developed by the WIPP and generator/storage site personnel. The Tl represents the radiation dose rate at 1 meter (3.28 ft) from the surface of the shipping container (TRUPACT-II with a load of 14 drums of waste or an RH cask) and depends on waste density, distribution of radionuclides, quantity of radionuclides per shipment, mix of waste types, self-shielding provided by the waste, and shielding provided by the TRUPACT-II container or RH cask. The Tl is very sensitive to small quantities of gamma-emitting fission products such as Cobalt-60 and Cesium-137. Tl values for typical shipments from each facility are shown in Table D.3.5. The radiation dose rate represented by the Tl was used to calculate radiation exposures of occupational populations (i.e., crew, shipment inspectors, waste handlers) and nonoccupational populations (people living or traveling along shipment routes, and people in the vicinity of the shipment while it is stopped). These Tl values are very conservative (see Appendix B) in that they were based on two key assumptions: 1) the maximum drum surface dose rates as measured by the facilities and 2) a drum source term and energy of 1 MeV. A more typical source term energy would be 06 to 0.1 MeVg for CH TRU waste. In the RADTRAN model, the people living along shipment routes were classified into urban, suburban, and rural fractions with respective population densities of 3,861, 719, and 6 persons per square kilometer as specified by the NRC (1977). These population densities are quite typical of urban, suburban, and rural environments. For example, statistics from the Denver Regional Council of Governments show that along Interstate 25 through Denver only a small area around downtown Denver has a population density exceeding the urban figure used in RADTRAN (3,997 persons per square kilometer for Denver versus the 3,861 assumed by RADTRAN). Other segments through Denver have much lower population densities than the RADTRAN urban value. Fifteen miles south of downtown, population densities along 1-25 approach the rural value of six persons per square kilometer. For truck shipments, the HIGHWAY model (Joy et al., 1982) was used to estimate trip lengths from various facilities to the WIPP and the corresponding population density fractions along these routes. The routes selected generally follow interstate highways as specified by the DOT for shipments of route-controlled quantities of radioactive materials. For rail shipments, the INTERLINE model (Peterson, 1984) was used to estimate trip lengths and population density fractions. The selected routes follow Class A/Class B main lines. These distances and population density fractions are summar- ized in Table D.3.6. Other major input parameters to RADTRAN are summarized in Table D.3.7. D.3.2.2 Results of the Analysis The radiation exposures that would be received from the normal transportation of CH and RH TRU waste by truck and rail are shown in Tables D.3.8 and D.3.9. These exposures are summarized for both occupational and nonoccupational populations. The radiological exposures are presented on a per-shipment basis for each facility and are given in doses (person-rem) received by the exposed population for each shipment. These per-shipment exposures were used to calculate the total incident-free transportation exposures for the Proposed Action and the two alternatives (see Table D-45 (0 Q. O « OS cr I- X o c o E Q. > (0 o TJ CO o « (0 CO d m _i CO < J3 c 3 o z 3 LU z z < I m _i z < « 3 C o (0 IT o ^ ^ ^ 'o ^ ^ 'o o o o o 2 8 8 8 8 s S 8 CM CO ca E 3 8552588 CJ Oi CO f^ ^ o o «- ,_ '^ >* •- Q_ 0, ▼- ^ »■ ▼" *" ^" ■^- *" *" X X X X X X X X X X X X X 8 8 8 8 8 !^ S in a s f^ 8 8 id ^" ^ ' CO 00 d ■* ro Ki '^ X X X X X X X X X X X X X Si 3 in S 8 K Si 8 8 8 s 8 ■* CO ^r^ 1= 10 ^ »- (O '- OOOOOooo""'-'"'-' xxxxx^xxxxxxx 88888g<$sfS888 8 CDCDOdcDt^jO^mOOOO 000000 000°°°° xxxxxx><><>^ ^ CO o d <\j u\ o *>■".' T o CNJ «- «- r>j v oj q^ cs o b o o o ° o ° ° o ° ° ° '"^^XXXXXxXXX in o cvi X s 8 S S S 8 ^ S S CO cvi T- in r^ iri i-^ in X ai X X X X X X 8 S S 5 S 8 d •r- in 1- •*' d X X 8 8 d d -o '^ o X 88888S8^8|8888 =0 °o °o °o '0 °o °o °o "b °a '0 °o °o X X X X X X X X X X X X X 8 8 8 8 JR % 5- J8 8 5 8 8 o> m CO »- CO ^ 8 CM E 3 CM CM CO CM E 3 a. z 00 CO CM E 3 'c o Q. C» CO CM E _3 'c o Q. O 1- CM E 3 'c o ^-» _3 Q. »- CM CM CM E 3 'E o a. CM E 3 'o o E < 5 CM E _3 3 O CM in CM E 3 CD O 00 X X o X X X CM '% in CM '^ CO o '^ in i <0 8 9 c <0 % d. m _i a. § -J z .^ _l CO £ _l . . •5 £« > Q if ^ « £ « e9 C Q. « c a _l c ■5 c <0 CO 8 ■0 1 ■3 Z CO •5 8 ii C c 2 8 i 0. TJ o> < in C c S (0 ■5. a <0 £ in -J ^ 2 \^ ^ (2 . a 3 « Q. u. a. T3 CL h- ^ « Si «> J3 ■53 10 f^- c:> 1 « E x: 'c» g . -ts c ■M, S 111 ■5 Z *^ "5 m c c « « CJ) (0 >• CB E a ■3 z C '£ JC a fj > z ^ « *c 1— c 3 c x> -1 CM < •5 ^^ CD uJ c .0 > _l '1 z < to Z 1- S T3 - >. <0 a: c E ^ E D-46 TABLE D.3.4 Average radioactivity in a shipment of RH TRU waste' Waste facility Radionuclide ANLE HANF INEL Cobalt-60 Strontium-90 Ruthenium-106 Antimony-1 25 Cesium-137 Cerium-144 Europium-155 Thorium-232 Uranium-233 Uranium-234 Uranium-235 Uranium-238 Neptunium-237 Plutonium-238 PlLJtonium-239 Plutonium-240 Plutonium-241 Plutonium-242 Americium-241 Curium-244 Californium-252 TOTAL 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.83 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 1.21 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 2.52 X 10"^ 9.27 X 10"2 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 2.97 X 10° 6.76 X 10° 1.89 X 10"^ 0.00 X 10° 9.46 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 5.41 X 10""* 8.11 X 10-^ 2.43 X 10"® 5.41 X 10'^ 0.00 X 10° 9.73 X 10"^ 1.38 X 10° 4.05 X 10 8.11 X 10° 8.65 X 10"^ 5.95 X 10"'' 0.00 X 10° 0.00 X 10° 0.00 X 10° 4.08 X 10° 0.00 X 10° 0.00 X 10° 5.81 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.68 X 10'^ 2.46 X 10"^ 0.00 X 10° 1.63 X 10"^ 8.80 X 10^ 3.58 X 10^ 0.00 X 10° 0.00 X 10° 3.27 X 10"^ 0.00 X 10° 0.00 X 10° LANL 0.00 X 10° 7.99 X 10° 6.31 X 10° 1.95 X lO""" 6.18 X 10° 6.22 X 10^ 3.13 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 9.48 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 8.29 X 10"^ 2.73 X 10"^ 1.26 X 10^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 0.00 X 10° ORNL 0.00 X 10° 1.12 X 10° 0.00 X 10° 0.00 X 10° 4.42 X 10"2 0.00 X 10° 0.00 X 10° 0.00 X 10° 4.56 X 10*^ 0.00 X 10° 1.87 X 10"® 1.96 X 10"® 0.00 X 10° 1.18 X 10"^ 3.67 X 10"^ 0.00 X 10° 0.00 X 10° 0.00 X 10° 1.88 X 10*2 1.69 X 10'^ 2.91 X 10"^ 9.18x10° 2.98x10^ 1.34 x 10^ 9.68x10^ 1.68x10° ® Radioactivity in curies per shipment for the volumes of waste assumed for the SEIS analyses (i.e., volumes scaled up to correspond to the design capacity of the WIPP--see last column, Table B.2.4). The volume per shipment is 0.89 m^ (one shipping cask per shipment). Key: ANLE, Argonne National Laboratory-East; HANF, Hanford Reservation; INEL, Idaho National Engineering Laboratory; U\NL, Los Alamos National Laboratory; ORNL, Oak Ridge National Laboratory. D-47 TABLE D.3.5 Transport index values® Facility Idaho National Engineering Laboratory Rocky Flats Plant Hanford Reservation Savannah River Site Los Alamos National Laboratory Oak Ridge National Laboratory Nevada Test Site Argonne National Laboratory-East Lawrence Livermore National Laboratory j Mound Laboratory CH TRU waste RH TRU waste itory 1 .0 5.0 1.5 b 0.7 16.0 2.7 b 4.1 8.9 11.0 3.2 1.2 b 7.5 2.5 Dratory 0.4 b 0.4 b ® mrem/hr at 1 meter from transporter surface. ^ Blanks = RH TRU waste not stored at facility. D-48 TABLE D.3.6 Average distances to the WIPP and percent of travel in various population zones Average distance Population zone Miles R U Truck Idaho National Engineering Laboratory Rocky Flats Plant Hanford Reservation Savannah River Site Los Alamos National Laboratory Oak Ridge National Laboratory Nevada Test Site Argonne National Laboratory-East Lawrence Livermore National Laboratory Mound Laboratory Rail Idaho National Engineering Laboratory Rocky Flats Plant Hanford Reservation Savannah River Site Oak Ridge National Laboratory Argonne National Laboratory-East Lawrence Livermore National laboratory Mound Laboratory 1521 85.0 13.8 1.2 874 82.3 15.7 2.0 1913 85.7 13.4 0.9 1585 74.3 25.1 0.6 343 90.1 9.9 0.0 1350 78.6 20.7 0.7 1286 86.8 11.2 2.0 1387 78.1 21.8 0.1 1458 86.2 10.1 3.7 1472 75.4 24.1 0.5 1761 89.5 9.8 0.7 1098 86.7 11.6 1.7 2296 87.8 11.5 0.7 1915 76.0 22.4 1.6 1630 79.8 18.9 1.3 1469 81.6 17.0 1.4 1873 85.0 14.3 0.8 1677 76.8 21.3 1.9 ® Mean population densities are utilized and correspond to: R = Rural (6 persons/km^) 8 = Suburban (719 persons/km^) U = Urban (3861 persons/km^). Source: Madsen et al., 1983. D-49 TABLE D.3.7 RADTRAN general input data^ Parameter CH TRU waste Truck Rail RH TRU waste Truck Rail Package type Package waste volume, m^ Packages/shipment Transport Index (Tl), mrem/hr Package length dimension, m Number of crewmen Distance from source to crew, m Speed, km/hr Urban population zone Suburban population zone Rural population zone Stop time per kilometer, hr/km No. of people exposed while stopped No. of people per vehicle Population density, people/km^ Urban population zone Suburban population zone Rural population zone Avg. rad./trailer-load of pkgs., Ci Accident release fractions TRUPACT-II 2.8 2.8 3 6 Cask 1.0 1.0 1 2 (Site-specific, see Table D.3.5) 7.32 7.32 3.61 3.61 2 5 2 5 4 152 5 152 24 24 24 24 40 40 40 40 88 64 88 64 .011 .0036 .011 .0036 50 100 50 100 2 3 2 3 3861 3861 3861 3861 719 719 719 719 6 6 6 6 (Site-specific, see Tables D.3.3 and D.3.4) (See Tables D.3.17 through D.3.22) I ® Source: Madsen et al., 1983. D-50 TABLE D.3.8 Radiological exposures per CH TRU shipment (person-rem)^''''^ Truck Rail Facility Occupational Nonoccupational Occupational*^ Nonoccupational Idaho National Engineering Laboratory 5.0 x 10 Rocky Flats Plant 4.0 x 10' Hanford Resen/ation 3.9 x 10 Savannah River Site 1.4 x 10" Los Alamos National Laboratory 2.8 x 10" Oak Ridge National Laboratory 1.3 x 10" Nevada Test Site 5.0 x 1 Argonne National Laboratory-East 1.3x10 Lawrence Livermore National Laboratory 1 .7 x 10 Mound Laboratory 1.9x10 -2 2.0 x 10 -2 -4 1.0 x 10" 2.9 X 10 2.7 X 10'^ -2 2.3 X 10 7.0 X 10' 8.0 X 10' 2.0 X 10" -2 -2 -1 -2 -2 2.0 X 10 -2 1.4 X 10" 9.0 X 10 9.0 X 10" -3 2.6 X 10 8.4 X 10" e 2.1 X 10" e 1.8 X 10' 1.2 X 10" -4 3.0 X 10" 2.0 X 10 4.0 X 10" 1.2 X 10 e -2 -1 2.0 X 10"'' e >-1 1.9 X 10 1.6 X 10 -2 1.1 X 10 -4 1.4 X 10 -2 Exposures per waste shipment are expressed in equivalent whole body dose and are tabulated in units of person-rem. Values for rail are expressed per railcar shipment. Exposures per waste shipment are presented as a function of the Transport Index (Tl) which is defined as the dose rate in mrem/hr at 1 meter from the waste package. Calculations are based on three TRUPACT-lls per truck and six per railcar. Rail occupational exposures resulting from normal transportation include the impact of DOT inspection activities (01 X Total Stop Time (hr) X Tl). No railheads present. D-51 TABLE D.3.9 Radiological exposures per RH TRU shipment (person-rem)^'''''^ ] Shipment origin facility Truck Rail Occupational Nonoccupational Occupational^^ Nonoccupational Idaho National Engineering Laboratory Hanford Reservation j Los Alamos National Laboratory I Oak Ridge National Laboratory Argonne National Laboratory-East 1.0 X 10 -1 1.7 X 10' 2.8 X 10 -2 6.3 X 10 5.0 X 10 -2 8.0 X 10 3.3 X 10 1.2 X 10' 4.4 X 10' -1 4.0 X 10 -2 1.3 X 10"^ 3.5 X 10"^ e 7.7 X 10"* 5.5 X 10"* 1.3 X 10' 2.9 X 10' e 7.4 X 10' 5.0 X 10" I ^ Exposures per waste shipment are expressed in equivalent whole body dose and are tabulated in unKs of person-rem. ^ Values for rail are expressed per railcar shipment. j ^ Exposures per waste shipment are presented as a function of the Transport Index (Tl) which is defined as the dose rate in mrem/hr at 1 meter from the waste package. Calculations are based on three TRUPACT-lls per truck and six per railcar. j "^ Rail occupational exposures resulting from normal transportation include the impact of DOT inspection activities (.01 X Total Stop Time (hr) X Tl). ® No railheads present. D-52 I s s t> < « a r o a c ■D 0) 3 V> o a X 0) O O CO d UJ _l CO < Q. O Q. O Q. O a o Q. 2 Q. lA « c c « E Q. x: CO o 8 c "o "o "o •^ 'b "b °o o o o O b '^ X X X X X X X X X X X o 1- (O CM CO (O CO CO CO CO o ^ 2 (O t^ (O »- "- CM '- >- r- Tf o °o b ^ "b b °o C>J b b b "b o X X X X X X X X X X X (J O) o o T- CO ^ o CO CO CM in o ir> *" ^ *" in CM ^ »- in CO CO o 'o 'b "b '^ 'b "b °o °o ^=0 °o '^ o c X X X X X X X X X X X o »- (O y- CO CO CO - h- *■ '" t '" CM CO '- ^ % % •^b •^ "b 'b °o °o 'b °o '% X X X X X X X X X X X o O o o CM r>«. CO o o CO CO CO ^ C\J (0 *" CO in CO ^ '" »- CM '- o 'b "b "b % *b "b °o °o °o b '^ ? X X X X X X X X X X X o m CO in CJ) CO '" in CM CO »- in r* in o o "o 'b 'b % "b "b °o °o °o °a •^ o c X X X X X X X X X X X o CO CO •* 1^ m r- ^ CO CD CM 'i- 2 t^ (O to '- '- M- '- "- r^ 'T '>b •^ '^ "h 'b "b °o °o 'b °o •^ u X X X X X X X X X X X 6 CO h- »- CO CM r^ CO CO in CO c» »- cvi '- CO in CM CO y- y- CM CJ> X XX »- CO >- X XX O O CM CM CO ^ .c o o c D) c m « it « -2 X CO X CO X CO X CO iri X CO X o CO X o t o V) tD a: % ^ I o o c I o £« ■« o CO ^ o ^ o CD ^ _l (0 _l « CO .« c (0 (J) o c ^ ♦- o 4> CD 4^ > 2 CD 2 CO r O « CJ) c CD < CC > tf> .^ CO o CO -I u X o CM X CO (A « LU o co' X CO (0 c CO .2 o CO _I 4) _ CO CD ■D (0 > O) CD c o ■s 2 « c c o O) (0 2 X X t CO '^ X X 00 ^ CM 1-^ o o I i 0! CO o ■s o CD _l XI c 3 o c ■«) u V o ■D T I 3 a o c^ 75 c o "S a 3 3 a o a. . XI in Q. fl) Q. .•c 5 > ■5 <1) a £ c O t) JQ fl) c a « E c Q £ (0 t> CD u a E E 3 C ■D CD 3 O C O r C ^ 1 on CO (0 O c CD « CO 1- ■5 CO t V _c c 3 V> E a X rt fl) c •*- ^ fl) O) T3 c 3 ■~ +j ♦-• 3 CO <0 fl) CD t- 4) a v> (- ^ fl) (0 (0 3 fl) <0 fl) 7) a. <0 X (0 3 o fl) a __ JI X (0 0) fl) c .0 to 4) . (0 u. __ a ♦- CD X 3 tr fc 'co >. CM nr li CO -Q n fl) CO fl) TJ 3 O CD E CO o C 3 fl) T3 c c T ts •D < R O ■5; 'S ^ c P u. c c < « i: £ »*- I u £ 0. r 0) 2 o .a "^ 1 8 i « I S S Q i| E « •5 o E x: -o Q- « t» (0 0) E >- ■s c o ■•§ a 3 O c o Z o o E a 8 I •5^ c c o O 73 c o a 3 I 4) P C 3 O I '^ IS- T3_ cs .•5 E o E I O Q. O 35 ■^ I I I I I I I I t 1^ 0000000000 xxxxxxxxxx iq 10 o> tf) »-. ^ cj *_ cvi CO I I I I I I I < ■ * 0000000000 xxxxxxxxxx l/JCNJT-<0'-»-IOCplOtO I I t I I I I I I I 0000000000 xxxxxxxxxx oiotof'-inooqp'-; irir>^ »-; 00 cvj X CO (Nj T- fvj »- ro 1 I 1 I 1 00000 X X X X X T- CO in in CO •^ to -r^ ci ■^ X CO ro ro 10 ro ro o 'o o o 'o X X X X X in o in CO CO. cj CO >«■■ 1-^ »-' X CO I I I I I I I >■ * 0000000000 xxxxxxxxxx r«.cor».. incN!cocJ)r>«: t»- ^ ^ O ^ O *" • ' C3 I o 000000000 xxxxxxxxxx 00O00CJ)<0O'-;CM'it; coinc\i'i-c\j^'-o)co''r- c 05 X c o O ■D « c/) < 5 :^ tr X CO 21 O yj 2: CO ^8 Q. Q. o T3 iJ TJ C (0 X 4> "6 E tr 'b'^^'0'0 X X X X X in CO CM cj T- co »-^ cvi CO «- 00000 X X X X X 1- CVJ T- K CJ> T-^ 0> CO ■r-' C\i •- »- o «- o 00000 X X X X X ^ ^ ^ in r>; c\i CO '-^ ^ t~» •D , o _j =! uj Z X J o < X CO 0) p> «* in ■* in ^ V- ^ ^ ''~ X X X X X X CO ^ * CM CO y- CM 'T CJ (O ■o°o o°o '0 'o X X X X X X o> en CO in CVJ * i>- ^ CO 1- »- % ^ C>J C\J 1 I ■ 00000 1 X X X X X X X s CO in CM CO cj CO r>^ '-^ ''>''? oooooooooo xxxxxxxxxx ^. t^. C'4 ■>- 1^, 00 a> o » CM ^ CO •r- CJ ^ •- in ■- tT 0) T- •- «d T- cJ CO cvi °o 'o°o o°o 'o o o X X X X X X X in CO in CO CO ^ cJ ^ O) 00 oooooooooo xxxxxxxx t«>.-" 1-^ Tt iri ,-'<£) r^ X X to y- 5" o X cJ ■- f\J C>J 1 (3 I O t o o o o o c> o_ O O «- O O • 1 1 oooooooooo xxxxxxxxxx coT-ooh-in<\jr>.^»-'cood'»-iri CM OJ CJ (M K> fO ^* oooooooooo xxxxxxxxxx pN.ino)cooo*o)0) CMC\JT-^cocNii-^i-'i-:
i o «- o O I o o o o o X X X X X o o cvj in CM cvi cj c\j cJ ; .-■ (d TT :'^ .■>•<■< v/:-'. encompasses a range of very severe accident conditions that are applied sequentially to determine cumulative effects; it includes impact (free drop), puncture, thermal, and water-immersion tests. The 1977 NRC study (NRC, 1977) conservatively estimated that approximately 9 percent of all truck accidents and 20 percent of rail accidents involving Type B containers or casks could result in radioactive material releases. More recently, however, Fischer et al. (1987) determined that only 0.6 percent of truck and rail accidents could cause a radiation hazard to the public. To estimate how much radioactive material could be released to the environment for the very small number of accidents that exceed the containment design capabilities of the Type B containers or casks, a release fraction analysis was performed. Release Fraction Definition . The release fraction analysis determined how much radioactive material could be released to the environment in a respirable, airborne form after a very severe accident that affects the containment capabilities of the shipping containers or casks. The calculation focused on respirable particle sizes with a mean aerodynamic diameter of less than 10 microns because inhalation is the primary exposure pathway for TRU elements. Particles that are larger will be expelled from the body and consequently are not as significant in estimating health effects. This calculational approach is consistent with existing NRC risk assessments (WASH-1400 NUREG-0170, NUREG/CR-4829). Method of Calculating Release Fractions . In order to calculate release fractions for very severe accidents, it is necessary to: • Characterize the radioactive material being transported • Identify and quantify the response of the shipping containers or casks (loss of containment) to accident conditions • Identify and quantify the release mechanisms resulting in the escape of radioactive material from the containers or casks to the environment. This analysis used representative values for parameters where published data and test results are applicable and reasonable, and conservative estimates where uncertainties exist. "Conservative" is used in this discussion to mean using such parameter values that the consequences of potential accidents will be overestimated. Characterization of the TRU Waste . The radionuclide compositions, quantities, and volumes used in the analysis are based on the waste inventory data and projections presented in Appendix B. As noted in Subsection 2.3.1, the DOE has established criteria and procedures which govern the physical, radiological, and chemical composition of the waste. Physical restrictions require that the waste not be in a free- liquid form and that particulate waste materials be limited to specific levels in accordance with DOE (1989). Transuranic radionuclides are generally present as oxides with concentrations exceeding 100 nanocuries per gram. z es I s i I D-69 Response of Shipping Containers and Casks . If a shipping container or cask is involved in an accident, the extent of damage will depend on the design of the container and the severity of the accident. Accident severity is categorized in terms of mechanical (e.g., impact) and thermal loads. Many methods can be used to classify accidents in terms of mechanical and thermal parameters. The relevant mechanical parameters may include impact speed, impact force, impact location and orientation, impact surface hardness, and impact puncture characteristics. The thermal parameters may include flame temperature, fire duration, fire source size and orientation with respect to the containers, and heat transfer properties (e.g., flame emissivity and convection coefficients). The analysis conducted for the SEIS used the accident severity model developed by the NRC (1977) as discussed in the preceding subsection. This model conservatively predicts the frequency of accidents whose severity exceeds Type B package test requirements (accident severity category three through eight). Because NRC regulations do not require Type B containers to be tested to failure, and because there are no historical data on the response of containers to very severe accidents, certain assumptions were required to estimate the extent of damage sustained by the TRUPACT-II container and the RH cask from accidents in severity categories three through eight. Guidance was obtained from the analysis and test data presented in NRC (1977), Fischer et al. (1987), and Jefferson (1978). The data indicate that a catastrophic failure (e.g., gaping hole, container severed in half) of a Type B container or cask would not be expected for accidents more severe than those in severity category two. Because of margins in the materials of construction (e.g., minimum versus actual rupture stress) and structural design (e.g., absorption of energy by plastic deformation), more likely failures would include the formation of cracks in the side of the container or cask, the failure of the closure seals, or the failure of any valves or penetrations. To define the response of Type B containers or casks to transportation accidents, the following conservative assumptions were made: • For shipments of several Type B containers on one transport vehicle, it was assumed that all containers would sustain the same damage. No credit was taken for the mitigating effects of one container shielding the others from impact forces or thermal loadings. • Two package response states were defined for the shipping container or cask: 1) No leak path and no release of radioactive material 2) A leak path is present, allowing the release of all respirable airborne radioactive material present inside the containers. The second state was postulated even though catastrophic failures are very unlikely. This state is consistent with NRC's position (Fischer et al., 1987) and does not take credit for any processes that will tend to reduce radioactive material releases (e.g., D-70 particle settlement, vapor plate-out on interior surfaces, filtration effects along leak path) from the containers. The response states are influenced by both the mechanical and thermal conditions of the accident. The response to the impact conditions will be largely independent of the thermal conditions, with impact effects immediate and thermal effects delayed. Consequently, the analysts elected to use two components for the response state (one for the impact event and one for the thermal event) for each accident severity category. Both components have two accident response states as defined above. Once the potential response states for the shipping containers or casks have been defined, it is necessary to assign the appropriate response state components to each accident severity category. As previously noted, there are few data that can be used to determine failure thresholds for transport containers involved in accidents with conditions more severe than NRC certification test requirements. NRC (1977) Model II release fractions (Table 5-8 of reference) were used as a primary guide. From impact test data, the NRC (1 977) projected Type B shipping containers for plutonium to have a failure threshold at accident severity category six. With current development programs, more recent container designs (1 985) were projected to have an increased failure threshold, corresponding to accident severity category seven. The NRC (1977) also projected Type B casks to have a failure threshold at accident severity category three, with more significant releases occurring at accident severity category five. These projections included effects from both impact and thermal events. For response to an impact event, a failure threshold corresponding to severity category five was assigned; it corresponds to the more significant release state projected by the NRC (1 977) for Type B casks. For response to a thermal event, a failure threshold corresponding to severity category three (an accident with conditions slightly exceeding the NRC's test requirements) was conservatively assigned. Release Mechanisms . Any release of radioactive material due to a transportation accident would normally progress in two stages: release inside the shipping containers or casks, followed by release to the environment. Releases from the container to the environment were addressed in the preceding discussion of accident response states. The discussion that follows evaluates how much radioactive material would be released into the cavities of the shipping containers or casks. There are multiple release mechanisms and pathways that may lead to the release of respirable radioactive material into container cavities. Impact release mechanisms include waste container (e.g., a 55-gallon drum or standard waste box) failure, fragmentation of solid waste, particulate suspension, and aerodynamic entrainment of particles. Thermal release mechanisms include heat-induced failures of the waste containers; aerosolization of particles by combustion, gas generation, or the heating of contaminated surfaces; and potential volatilization of radionuclides. Impact and thermal release mechanisms were evaluated by using applicable test data and analyses available in the published literature, as supplemented by conservative assumptions where only limited data exist. It was assumed that all failed waste containers, without regard to waste form or type, release an average amount of material for each accident severity category. z CO i D-71 I I In assessing releases from impact events for each severity category, the following j procedure was used: I • Identification of the fraction of failed waste containers inside the shipping container or cask • Determination of the fraction of radioactive material released from the failed waste containers • Calculation of the fraction of radioactive material released from the failed waste containers that is aerosolized in a respirable form by the mechanical stress of impact • Calculation of the fraction of radioactive material released from the failed waste containers that becomes aerodynamically entrained in a respirable form after the loss of containment by the shipping containers and any subsequent depressurization (e.g., TRUPACT-II design pressure of 50 psig). Studies by Huerta (1983) and Shirley (1983) were used to determine the fraction of failed waste containers. The fractions of radioactive material released from the failed waste containers were consen/atively estimated using reports by Huerta (1983) and the NRC (1977) for guidance. The fraction of radioactive material converted to a respirable aerosol from impact stresses was calculated by using a resuspension factor approach. This is an accepted analytical method for predicting airborne concentrations of material above contaminated surfaces. The mechanical action of vigorous sweeping was used to represent the respirable airborne contamination fraction, using data taken from an NRC report (NRC, 1980), for the resuspension factor. It was judged that this approach would be at least representative, if not conservative, in estimating the release of respirable contaminants by impact stresses. The aerodynamic entrainment of respirable particulates was determined by using data from wind tunnel tests for uranium dioxide power (Mishima and Schwendiman, 1973a). This release mechanism will occur only to the extent that the shipping container is pressurized by the release of gases from the waste containers. The analysis conservatively assumed that maximum pressurization of the container cavity will always occur for every shipment. Based upon the nature of potential container damage previously described, and the void volume space within the container cavity, a depressurization duration of approximately 30 minutes at an average velocity of about 2.5 mph was calculated. For these conditions, the average entrainment value given by Mishima and Schwendiman (1973a) for four surfaces (asphalt, sand, vegetation, and stainless steel) was conservatively assigned. The algorithm used to calculate the release fraction of respirable radioactive material from impact stresses is summarized in Table D.3.17. Values for specific algorithm parameters are presented in Table D.3.18. D-72 TABLE D.3.17 Estimate of potential accident release fractions for CH and RH TRU waste shipments due to impact events Impact release fraction (IRF) = (FFC x FMRC) (FMAI + FMEI) (FMRPI) Where: FFC FMRC FMAI FMEI FMRPI = Fraction of failed waste containers = Fraction of material released from failed containers ! into package cavity Fraction of material aerosolized from impact Fraction of material entrained to environment during impact event Fraction of material released from package cavity during impact event I TRUPACT-II® RH Cask a,b Severity category FMRC FMAI 1 Ox 10° Ox 10° 2 Ox 10° Ox 10° 3 1 X 10"^ 8 X 10-5 4 3 X 10"^ 8 X 10-5 5 5 X 10"^ 8 X 10-5 6 7 X 10"^ 8 X 10-5 7 1 X 10° 8 X 10'^ 8 1 X 10° 8 X 10-5 FMEI FMRPI FFC IRF FFC 0.0 X 10° Ox 10° 0.0 X 10" 0.0 X 10° 0.0 X 10° 1.5 X lO"'* 1.5 X lO'"* 1.5 X 10"^ 1.5 X 10"* Ox 10" Ox 10° Ox 10° 1 X 10° 1 X 10° 1 X 10° 1 X 10° r1 3x 10 5 X 10"'' X 10° 7 x lO"'' 7x 10 r1 8 X 10" 1 X 10^ 1 X 10° 2 X 10"^ 1 X 10^ 1 X 10° 1 X 10° 2x 10" 1 X 10^ Respirable release fractions. Release fractions are the same for truck and rail transportation modes. IRF X 10° X 10° Ox 10° X 10° X 10° X 10° Ox 10° X 10° 3 X 10"'' Ox 10" Ox 10° Ox 10° 1 X 10"^ 1 X 10"^ 2x 10-^ 2 X 10"^ 1 X 10° 2 X 10"^ z a & i s D-73 TABLE D.3.18 Impact release algorithm parameters for CH and RH TRU waste shipments Parameters FFC FMRC FMAI FMEI FMRPI Value .2728 InF -2.814 Table D.3.17 Table D.3.17 ! 1.50x10-^ Basis/reference I Accident severity 1-4: j 0.0 I j Accident severity 5-8: ! 1.0 Huerta (1983); Shirley (1983). Where F is NRC (1977) accident severity breach force (Newtons) Huerta (1983) and NRC (1977) used as guidance NRC (1980) resuspension factor of 2.00 X 10"^ m'^ used (mechanical stress of vigorous sweeping) Mishima and Schwendiman (1 973a) average entrainment value for 4 surfaces used with airflow of 2.5 mph for 30 minutes Type B package design and NRC (1977) used as guidance D-74 Fischer et al. (1987) estimated that 1.7 percent of truck accidents and 6.8 percent of rail accidents will involve fires. For fire events, the following method was used for each accident severity category: • Identification of the fraction of radioactive material subject to thermal release mechanisms • Calculation of the fraction of radioactive material released by combustion in a respirable form • Calculation of the fraction of radioactive material released in a respirable form by the release of gases and the heating of contaminated surfaces • Determination of the fraction of radioactive material released in a respirable form from any volatilization of radionuclides. In the absence of detailed knowledge about the responses of shipping containers and waste containers to fires more severe than those specified in regulatory test requirements for Type B packagings, it was conservatively assumed that all radioactive material was available for release for all accidents exceeding severity category two, as limited by the specific release mechanisms. For combustion related releases, it was assumed that combustible materials could be ignited in all accident severity categories exceeding category two. To maximize the amount of combustible waste burned for a given amount of oxygen, incomplete combustion, producing carbon monoxide (CO), was assumed. The amount of oxygen present to support combustion was calculated by assuming an 85 percent void volume for a loaded shipping container and observing that there would be no external sources of air or oxygen (no major breach of container). From a review of the inorganic compound tables in the Handbook of Chemistn/ and Physics , it was concluded that any decomposition of metal hydroxides (e.g., Ca(0H)2, AI(0H)3) present in cemented sludges would not act as an internal source of additional oxygen. Finally, the results of experiments conducted by Mishima and Schwendiman (1973b) were used to assess the fraction of radioactive material released in a respirable form from the burning of combustible material. For accident severity categories four through eight, the fire event may last longer than 1 .5 hours. For these more severe conditions, it was assumed that more radioactive material could be converted to an aerosol form because of the release of gases from the waste at elevated temperatures. Potential gas generation was assumed to be comparable for all five accident severity categories and was calculated by assuming a graphite/steam reaction as the off-gassing source. For an upper bound gas generation estimate, it was further assumed that all waste containers within the shipping container were loaded with solidified process waste (water/steam source) and that there was adequate graphite (e.g., molds) present to react with all of the steam. With these assumptions, gas generation was calculated to be in excess of 600 TRUPACT-II void volumes and 700 RH cask void volumes, at atmospheric pressure. I I 1 D-75 =^^ The fraction of respirable radioactive material present in the gases released from the waste containers and subsequently to the environment was calculated by using a resuspension factor approach. A resuspension factor value corresponding to a vigorous and continued surface stress of people walking on a surface contaminated with Plutonium dioxide (at a rate of 36 steps per minute) was used in the analysis. Vaporization was reviewed as another thermal release mechanism. As previously noted, TRU radionuclides are generally present in an oxide form. They are highly stable at elevated temperatures. Alexander et al. (1986) report that volatile releases of transuranic radionuclides are not of any significance until temperatures of 3140°F are reached. The volitization of uranium oxide (e.g., UOg) becomes measurable at approximately 2960° F. Flame temperatures for the open burning of hydrocarbon fuels (e.g., JP-4, gasoline, diesel) range from MOCF to 2400 °F, with a median temperature of approximately 1 800 ° F. Consequently, a volatile release of TRU or uranium oxide material is not credible for a transportation accident. This is consistent with the release analysis presented by Fischer et al. (1987), in which the releases of TRU material are quantified in terms of particulates only. In conjunction with waste characterization data, it can be concluded that potential accidents involving CH TRU waste shipments cannot result in radioactive material releases in a vapor form. However, RH TRU waste contains activation/fission products that may volatilize at elevated temperatures. These radionuclides are identified as being present in RH TRU waste. Testing conducted by Lorenz (1980) indicates that cesium, antimony, and ruthenium may volatilize at elevated temperatures. Assuming that volatilization mechanisms for RH TRU waste would be similar to the referenced test conditions at 1290°F, it was concluded that the releases of cesium, antimony, and ruthenium vapors would be comparable to the values estimated for respirable particulate releases. The algorithm for estimating the respirable release fraction of radioactive material from thermal accident events is illustrated in Table D.3.19. Values for specific algorithm parameters are summarized in Table D.3.20. Total Respirable Release Fractions . The calculated impact release fractions (Table D.3.17) and thermal release fractions (Table D.3.19) were added to determine the total respirable release fractions due to very severe transportation accidents and are summarized in Table D.3.21 and D.3.22. A maximum release fraction of 0.0002 was estimated for accidents involving both CH and RH TRU waste shipments. This is consistent with or bounding of previous transportation risk studies such as the NRC modal study (Fischer et al., 1987), which estimated particulate releases of 0.000002 and vapor (Cg)releases of 0.0002 due to spent fuel shipments, and the WIPP FEIS (DOE, 1980), which incorporated a release fraction of 0.00018 for CH TRU waste shipments. D.3.3.1.3 Dispersal Conditions . The dispersion of airborne radioactive material during an accident is controlled by meteorological conditions at the time of the accident. The airborne radioactive material moves downwind from the scene of the accident and its dispersal and transport are affected by the degree of atmospheric turbulence. For this analysis, the materials were assumed to move downwind and disperse. As the radioactive cloud disperses, the people in its path will be exposed to external radiation, internal radiation from inhalation, or internal radiation from ingestion. For inhalation and D-76 TABLE D.3.19 Estimate of potential accident release fractions for CH and RH TRU waste shipments due to thermal events Thermal release fraction (TRF) Where: FAT PMC FMAC FMAT FMRPT FAT [(FMC X FMAC) + FMAT] FMRPT Fraction of accidents involving a thermal event Fraction of material consumed by combustion Fraction of material aerosolized by combustion Fraction of material aerosolized by thermal event Fraction of material released from package cavity during thermal event Truck® Rail® Severity Category FMAC FMAC FMAT FMRPT FAT TRF FAT TRF TRUPACT-II 1 Ox 10° Ox 10° Ox 10° Ox 10° 1.7 X 10-2 Ox 10° 6.8 X 10-2 Ox 10° 2 Ox 10° Ox 10° Ox 10° Ox 10° 1.7 X 10-2 Ox 10° 6.8 X 10-2 Ox 10° 3 9x 10-^ 5x 10-4 2x 10-^ 1 X 10° 1.7 X 10-2 8 X 10-^ 6.8 X 10-2 2x 10"^ 4 9x 10-^ 5x 10-4 1 X 10-5 1 X 10° 1.7 X 10-2 2 X 10'^ 6.8 X 10-2 7 X 10-'^ 5 9x 10-4 5x 10-4 1 X 10-5 1 X 10° 1.7 X 10-2 2 X 10"^ 6.8 X 10-2 7 X 10-^ 6 9x 10-4 5x 10-4 1 X 10-5 1 X 10° 1.7 X 10-2 2 X 10-^ 6.8 X 10-2 7 X 10-^ 7 9x 10-4 5x 10-4 1 X 10-5 1 X 10° 1.7 X 10-2 2 X lO-"^ 6.8 X 10-2 7 X 10-'^ 8 9x 10-4 5x 10^ 1 X 10-5 1 X 10° 1.7 X 10-2 2 X 10-^ 6.8 X 10-2 7 X 10-^ RH Cask 1 Ox 10° Ox 10° Ox 10° Ox 10° 1.7 X 10-2 Ox 10° 6.8 X 10-2 Ox 10° 2 Ox 10° Ox 10° Ox 10° Ox 10° 1.7 X 10-2 Ox 10° 6.8 X 10-2 Ox 10° 3 7x 10-4 5x 10-4 2x 10-^ 1 X 10° 1.7 X 10-2 6 X 10-® 6.8 X 10-2 2x 10-® 4 7x 10-4 5x 10-4 9x 10-5 1 X 10° 1.7 X 10-2 2 X lO-'^ 6.8 X 10-2 7 X 10-'^ 5 7x 10-4 5x 10-4 9x 10-5 1 X 10° 1.7 X 10-2 2 X 10-^ 6.8 X 10-2 7 X 10-'^ 6 7x 10-4 5x 10-4 9x 10-5 1 X 10° 1.7 X 10-2 2 X 10"^ 6.8 X 10-2 7 X 10-^ 7 7x 10-4 5x 10-4 9x 10-5 1 X 10° 1.7 X 10-2 2 X 10-^ 6.8 X 10-2 7 X 10-^ 8 7x 10-4 5x 10-4 9x 10-5 1 X 10° 1.7 X 10-2 2 X 10-^ 6.8 X 10-2 7 X 10-^ I I s Respirable release fractions. D-77 TABLE D.3.20 Thermal release algorithm parameters for CH and RH TRU waste shipments Parameter FAT ! Value } 1.7 X 10-2 (Truck) 1 6.8 X 10-2 (Rail) Basis/reference Fischer et al. (1987) FMC FMAC ! FMAT ! FMRPT I Accident severity 1-2: I Ox 10° j Accident severity 3-4: I 9x 10-^ (TRUPACT-II] I 7x10"^ (RH Cask) { Accident severity 1-2: j 0x10° I Accident severity 3-8: I 5x10"^ } Accident severity 1-2: I 0x10° j Accident severity 3: ! 2x10-^ Accident severity 4-8: 1x10-^ (TRUPACT-II) 9x 10"^ (RH Cask) [ Accident severity 1 -2: j Ox 10° j Accident severity 3-8: ! 1 X 10° Type B package design Limited internal oxygen source: 3.95 lb O2 (TRUPACT-II) 0.73 lb O2 (RH Cask) Type B package design Mishima and Schwendiman (1973b) Type B package design Only combustion assumed to occur, with attendant off-gas (combustion) products Off-gasing assuming steam/graphite reaction and resuspension factor of 5.00 x 10"^ m'"* corresponding to a surface stress from walking (NRC, 1980) Type B package design NRC (1977) used guidance as D-78 TABLE D.3.21 CH TRU waste transportation release fractions Total respirable release fraction (TRRF) Impact release fraction (IRF) + Thermal release fraction (TRF) Accident Impact severity release category fraction^ Truck 1 Ox 10° 2 0x10° 3 Ox 10° 4 Ox 10° 5 8x 10-^ 6 2x 10"^ 7 2x 10-^ 8 2x 10-^ Rail 1 Ox 10° 2 Ox 10° 3 Ox 10° 4 Ox 10° 5 8x 10-^ 6 2x 10-^ 7 2x 10-^ 8 2x 10-^ ^ From Table D.3.17. ^ From Table D.3.19. Thermal release fraction'' Ox 10° Ox 10° 8x 10-^ 2x 10-^ 2x 10*^ 2x 10-^ 2x 10-"^ 2x 10-^ Ox 10° Ox 10° 2x 10"^ 7x 10-^ 7x 10-^ 7x 10"^ 7x 10-^ 7x 10-^ Total respirable release fraction Ox 10^ Ox 10^ 10- 10- 10' 10- 10" 10" Ox 10° Ox 10° 2x 10-® 7x 8x 2x 2x 2x 10" 10 10- 10" 10" -5 I I 1 D-79 TABLE D.3.22 RH TRU waste transportation release fractions Total respirable release fraction (TRRF) Accident severity category Impact release fraction® Impact release fraction (IRF) + Thermal release fraction (TRF) Thermal release fraction*^ Total respirable release fraction Truck 1 2 3 4 5 6 7 8 Rail 1 2 3 4 5 6 7 8 Ox 10" Ox 10° Ox 10° Ox 1 x 1 X 2x 10" 2 X 10' 10^ 10-^ 10-- Ox 10^ Ox 10^ Ox 10^ Ox 10^ 1 X 10 1 X 10- 2x 10' 2x 10 -4 -4 Ox 10" Ox 10° 6x 10"^ 2x lO-'' 2x 10'^ 2x lO"'^ 2x 10-''' 2x lO*"^ Ox 10° Ox 10° 2x 10-^ 7x 10-^ 7x lO-'' 7x 10"' 7x 10-^ 7x 10'^ Ox 10° Ox 10° 6x lO-'^ 2x 10 -7 1 X 10- 1 X 10- 2x 10" 2x 10 -4 Ox 10" Ox 10° 2x 10-® 7x 10-^ 1 X 10-^ 1 X 10-^ 2x 10-^ 2x 10-^ ® From Table D.3.17. ^ From Table D.3.19. D-80 mm ingestion, the degree of exposure depends on the amount of material retained in the lungs or other organs of the exposed persons. Airborne transport and diffusion can disperse radioactive materials over large areas. The degree of dispersion is influenced by many factors, such as season (which influences atmospheric turbulence), time of day, degree of cloud cover, land surface features and characteristics, and other meteorological parameters. Dispersed material can expose people in many ways, as shown in Figure D.3.3. The principal effect of gamma-emitting materials is a direct external or internal dose. Material that emits alpha or beta radiation if it is converted to an aerosol and inhaled by people produces the largest consequence. Figure D.3.3 illustrates that radioactive materials can also be incorporated in the food chain. Radiation doses received by the population through the food chain pathway are usually more significant if a continuous release exists. One of the pathways of note is resuspension. This occurs when deposited particulate material becomes airborne through the action of pedestrians, vehicles, plowing, the wind, etc. The resuspended material then becomes available for inhalation and can deliver an additional dose that accumulates with time. D.3.3. 1.4 Pathwavs and Exposed Populations . RADTRAN or similar analytical tools can be used to evaluate the radiological impacts of transporting radioactive materials under accident conditions. As input to RADTRAN, the exposure pathways must be identified and the size of exposed populations must be estimated. Transportation accidents may be divided into those accidents in which the shipping containers maintain their integrity and there is no release of radioactive materials, and those accidents in which the integrity of the shipping containers is compromised. The exposure pathways and the exposed population subgroups are discussed below. In an accident that does not compromise the containment of the shipping containers, the exposure pathway is limited to direct exposure by penetrating radiation from the intact package. The dose delivered to any member of an exposed population is evaluated in the same manner as the exposure from normal (incident-free) transporta- tion, with adjustments made for the duration of exposure and the distance between the shipment and the exposed individuals. The exposed populations include the truck or rail crew, the occupants of the other vehicle(s) involved in the accident, bystanders and pedestrians, the occupants of nearby buildings, and the members of emergency response crews. In an accident that results in a failure of the shipping containers and possible release of radioactive material, exposures may result from both nondispersible and dispersible materials. The exposure pathway from accidents involving shipping containers with nondispersible materials is direct exposure resulting from the loss of shielding of the contents of the containers. Certain radioactive materials are not dispersible because of their chemical or physical form, such as irradiated steel hardware; these materials may nevertheless result in exposure by penetrating radiation. The doses received by exposed individuals are evaluated in the same manner as other direct exposures, with adjustments made I I s I D-81 r^ /—N d z u. Ill oc CO UJ CO < UJ -I UJ oc UJ Q li O 3 Z O Q < OC CO v-y v_y UJ u. "■o H 0) I X a. UJ £ (7) (A o Q. D-82 for increased dose rates resulting from shielding loss as well as exposure time and distance adjustments. The exposed populations are the same as identified above. Four exposure pathways may result from accidents that cause a release of dispersible radioactive materials: • Cloudshine : The exposure from cloudshine is the direct external dose from the passing cloud of dispersed material. Dispersion depends on the meteorological conditions at the accident scene, as well as the fraction of failed shipping containers and the fraction of released material that becomes airborne. • Groundshine : The exposure from groundshine is the direct external dose from material that has deposited on the ground after being dispersed from the accident site. The degree of deposition depends on the material being deposited (i.e., the rate at which the dispersed material settles out) and the amount of dispersed material available to settle out (i.e., how much material from the original release has dispersed far enough to deposit on the area of interest). • Inhalation : The exposure from inhalation is the internal exposure that results from breathing aerosolized material. Exposure from inhalation depends on the fraction of failed shipping containers, the fraction of material that becomes airborne, the aerosol fraction of respirable size, the radiation dose delivered per curie of radioactivity inhaled, the dilution factor for radioactive material in the surrounding air, and the breathing rate of the exposed individual. • Resuspension : The exposure from resuspension is the internal exposure that results from the inhalation of material that was dispersed, deposited at a distance from the accident scene and then resuspended as an aerosol and inhaled. Exposure from resuspension requires combining the mechanisms of dispersion, deposition and inhalation described above, as well as estimating the fraction of deposited material that is resuspended. (Resuspension may result from changing weather conditions, such as changes in wind speed or direction, or from disturbing deposited material by other means, such as traffic through a deposition area.) Note that exposure by ingestion is not included in evaluating the radiological impacts of accidents because it is assumed that emergency response and governmental authorities would intervene to impound foodstuffs, provide an alternative water supply, and clean up contaminated land. The population subgroups that are exposed by an accident that results in dispersion of radioactive material include the individuals who are directly exposed at the scene of the accident and the individuals who are present in the areas over which dispersion occurs. I I 1 D-83 D.3.3.2 Results of the Accident Analysis The radiological exposures associated with truck or rail accidents involving CH TRU waste are expressed as the exposure per shipment and as a cumulative exposure over the shipping campaign for the alternative being considered. The exposure is the sum of the products of the probability of a given severity accident times the consequences of such an accident for each of the severity categories. The radiological exposures from an accident involving CH TRU waste are expressed in equivalent whole body dose and are tabulated in units of person-rem, and assume three TRUPACT-II containers per truck shipment and six TRUPACT-II containers per rail shipment. Table D.3.23 presents the exposure per shipment for each facility that ships CH TRU waste and the total per shipment exposure for all facilities for truck and rail modes. Table D.3.24 presents the cumulative exposure for all facilities that ship CH TRU waste to the WIPP. This table shows the estimated radiological exposures for transportation accidents in the Proposed Action, which consists of the Test Phase (10 percent of CH TRU waste shipped and all shipments by truck) and the Disposal Phase, in which truck or rail could be used. No radiological exposures from transportation accidents were calculated for the No Action Alternative because no shipments to the WIPP would be made. For the Alternative Action, the radiological exposures from truck accidents are the sum of the exposures from the Test Phase and Disposal Phase (Table D.3.24). These exposures would be incurred in a continuous 20-year period after an approximate 5-year Test Phase during which no waste would be shipped to the WIPP but during which approximately seven truck shipments of CH TRU waste would be made from the Rocky Flats Plant to the Idaho National Engineering Laboratory to support bin tests. The accident contribution for these shipments was calculated by subtracting the per- shipment radiological exposure from accidents (Table D.3.23) for a shipment from the Idaho National Engineering Laboratory to the WIPP from that for a shipment from the Rocky Flats Plant to the WIPP. This difference, which represents the Idaho-to-Rocky Flats transportation segment, was multiplied by the number of shipments to arrive at the transportation exposures from the bin tests. Thus, an accident contribution of approximately 5.90 x 10"^ person-rem is expected from the bin test shipments. The radiological exposures from rail accidents for the Proposed Action and the Alternative Action are shown in Table D.3.25. The radiological exposures from an accident involving a truck or a railcar carrying RH TRU waste are expressed in equivalent whole body dose and are tabulated in units of person-rem, assuming one RH TRU cask per truck shipment and two RH casks per rail shipment. Table D.3.26 presents the per shipment exposure for each facility that ships RH TRU waste by truck or rail and the total exposures for all facilities. Table D.3.27 presents the cumulative exposure for all facilities that ship RH TRU waste to the WIPP. These lifetime radiological exposures from transportation accidents involving RH TRU waste are shown in Table D.3.27 for a 20-year shipping period. No RH TRU waste shipments would occur during the Test Phase of the Proposed Action or the Alternative Action, and therefore no accident exposures result. The radiological exposures of RH TRU shipments are identical for the Proposed Action and the Alternative Action. D-84 TABLE D.3.23 Per shipment accident radiological exposures of CH TRU waste shipments (person-rem)^''^'^ Nonoccupational accident contribution Facility Truck Rail Idaho National Engineering Laboratory 7.9 x lO"^ Rocky Flats Plant 2.0 x 1 0"^ Hanford Reservation 9.9 x 1 0^^ Savannah River Site 4.2 x 1 0"^ Los Alamos National Laboratory 1.3x1 0"'^ Oak Ridge National Laboratory 4.4 x lO'^ Nevada Test Site 8.9 x 1 0^^ Argonne National Laboratory-East 4.9 x 1 0^^ Lawrence Livermore National Laboratory 1 .9 x 1 0^^ Mound Laboratory 2.8 x 10"^ 5.7 X 10-^ 1.9 x 10-^ 8.9 X 10-^ 1 4.0 X 10-2 d 1 1 4.??> c 10-3 , d 3.5 X 10-4 2.94 > :10-4 1 5.4 X 10-^ ! I 2 ^ Population group exposures per waste shipment are expressed in equivalent whole [ body dose and are tabulated in units of person-rem. ^ Values for rail are expressed per railcar shipment. ^ Population group exposures per waste shipment are presented as a function of the [ Transport Index (Tl), which is defined as the dose rate in mrem/hr at 1 m from the | waste package. No railheads present. D-85 TABLE D.3.24 Lifetime radiological exposures for accidents during transportation of CH TRU waste (person-rem): Proposed Action and Alternative Action®"^ Proposed Action Disposal Phase (20-yr) Alternative Action Disposal Phase (20-yr) Facility Test Phase"^ Truck Max. rail Truck Max. rail Idaho National Engineering Laboratory Rocky Flats Plant Hanford Reservation Savannah River Site Los Alamos National Laboratory^ Oak Ridge National Laboratory Nevada Test Site"^ Argonne National Laboratory-East Lawrence Livermore National Laboratory Mound Laboratory Total 3.2 xlO-"" 2.9x10° 1.0x10° 3.2x10° 1.2x10° 1.5 xlO-"" 1.4x10° 6.5x10-^ 1.5x10° 7.2x10-'' 3.1 X lO-"" 2.8 X 10° 1.2 X 10° 3.1 x 10° 1.4 x 10° 1.1x10^ 1.0x10^ 4.8x10^ 1.1 xlO^ 5.3x10^ 2.7x10-"' 2.4x10° 2.4x10° 2.7x10° 2.7x10° 1.0x10-"' 9.0x10-'' 4.3x10"^ 1.0x10° 4.8 x 10-^ 7.1 X 10-^ 6.4 X 10-^ 6.4 X 10-^ 7.1 x 10"^ 7.1 x 10-^ 6.9x10-^ 6.2x10-3 2.2x10-3 6.9 x 10-^ 2.4 x 10-^ 1.8x10-2 1.6x10-'' 1.3x10-'' 1.8x10"'' 1.4x10"^ 4.2 X 10-^ 3.8 X 10-3 3 6 x 10-^ 4.2 x 10*3 4.0 x 10-^ 1.2x10^ 1.1x10^ 5.4x10^ 1.2 xlO^ 6.0x10^ 3 Population group exposures are calculated by multiplying the exposure/shipment idenfrfied in Table D^3^23 by the total number of shipments to the WIPP by truck or rail, as determined from the projection in Table D.3.2. ^ Test Phase assumes 10% of shipment completed by truck. ^ Nonoccupational population. d waste shipments from this facility are limited to truck mode, thus rail exposures are the same as truck exposures. D-86 TABLE D.3.25 Summary of lifetime radiological exposure changes between Proposed Action and Alternative Action: CH TRU accident nonoccupational risk (person-rem) Facility Proposed Action Alternative Action Truck Rail Truck Rail Idaho National Engineering Laboratory 3.2 x 10° 1.3 x 10° 3.2 x 10° 1.2 x 10° Rocky Flats Plant Hanford Reservation Savannah River Site Los Alamos National Laboratory Oak Ridge National Laboratory Nevada Test Site Argonne National Laboratory-East 1.5x10° 8.0x10""' 1.5x10° 7.2x10 •1 3.1 X 10° 1.5 X 10° 3.1 X 10° 1.4 x 10° 1.1 xlO^ 5.9x10^ 1.1x10^ 5.3x10^ 2.7x10° 2.7x10° 2.7x10° 2.7x10° 1.0x10° 5.3 xlO-"" 1.0x10° 4.8 xlO-"" 7.1 x 10-^ 7.1 X lO-'* 7.1 X 10"^ 7.1 x 10"^ 6.9 X 1 0"^ 2.9 X 1 0"^ 6.9 x 1 0'^ 2.4 x 1 0"^ Lawrence Livermore National Laboratory 1 .8 x 1 0'"" 1.5x10'"' 1.8x1 0'"' 1.4x1 0""' Mound Laboratory 4.2 x 10'^ 4.6 x lO'"* 4.2 x 10"^ 4.0 x 10"^ Total 1.2x10^ 6.6x10^ 1.2x10^ 6.0x10^ I I \ D-87 TABLE D.3.26 Per shipment accident radiological exposures of RH TRU shipments {person-rem)^'°'^ Nonoccupational accident contribution Facility Truck Rail } Idaho National Engineering Laboratory 1.6 x10-^ 1.3 xlO-^ j Hanford Reservation 4.34 X 10'^ 4.44 X 10'^ j Los Alamos National Laboratory 3.09 X 10-® d 1 Oak Ridge National Laboratory 4.84 X 10-^ 5.21 X 10-^ Argonne National Laboratory-East 6.4 X 10-^ 5.2 X 10-^ ' 3 Exposures to the population per waste shipment are expressed in equivalent whole ' body dose and are tabulated in units of person-rem. ^ Values for rail are expressed per railcar shipment. I ^ Exposures to the population per waste shipment are presented as a function of the ' Transport Index fTI) which is defined as the dose rate in mrem/hr at 1 meter from the waste package. Calculations are based on three TRUPACT-II waste packages per truck and six per railcar shipment. ^ No railheads present. D-88 TABLE D.3.27 Lifetime radiological exposures for accidents during transportation of RH TRU waste (person- rem): Proposed Action and Alternative Action®'^ Facility Idaho National Engineering Laboratory Hanford Reservation Los Alamos National Laboratory^ Oak Ridge National Laboratory Argonne National Laboratory-East Total 100% Truck 7.8 X 10"^ 1.1 X lO-"" 3.1 X 10-^ 2.2 X 10-2 1.9 X 10"^ 9.1 X lO'"" Maximum rail 3.2 X lO-"" 5.4 X 10-2 3.1 X 10-^ 1.2 X 10-2 7.8 X 10"* 3.9 X lO-"" ® Population group exposures are calculated by multiplying the exposure/shipment | identified in Table D.3.26 by the total number of shipments to WIPP by truck or rail, | as determined from the projection in Table D.3.22. Rail occupational exposures j resulting from normal transportation include the impact of inspection activities. ^ Nonoccupational populations. i ^ Waste shipments from the facility are limited to truck mode. Rail exposures are thus j the same as the truck exposures. i I I 1 D-89 D.3.4 RADIOLOGICAL CONSEQUENCES OF BO UNDING CASE TRANSPOR- TATION ACCIDENT D.3.4.1 Assumptions: Boundi ng Case Accident As discussed in Section 5.0, "bounding case" transportation accident scenarios were developed for this SEIS. These scenarios were used to calculate the impact of very severe accidents in higher population areas along the WIPP-preferred transportation routes. Postulated accidents involved both OH and RH truck and rail shipments using TRUPACT-II containers or RH casks. Based on comments received on the draft SEIS, a revised bounding case accident was calculated based on higher curie content OH waste primarily from Los Alamos National Laboratory, the Savannah River Site, and the Idaho National Engineering Laboratory. In the draft SEIS. calculations assuming aver- age OH waste from the Rocky Flats Plant waste were used because these shipments comprise the majority of the total OH waste shipments. Less likelihood of the current bounding case accidents is expected because the number of shipments of maximally loaded containers (WAG or TRUPACT Payload Compliance Plan limits) are smaller than the number of shipments with average waste loadings. Waste compositions from Los Alamos National Laboratory, Savannah River Site, and the Idaho National Engineering I Laboratory were analyzed for OH TRU shipments, and from Hanford and the Idaho ! National Engineering Laboratory for RH TRU shipments. These waste composrt^^^^^^^^^ I were scaled up to the maximum total curie content of radionuclides allowed by either ! the WIPP Waste Acceptance Criteria or the TRUPACT Payload Compliance Plan. i During each accident, all TRUPACT-II containers or RH casks were assumed to be equally breached and subsequently engulfed in fire for two hours (it is estimated that at least 17 000 gallons of fuel would be required to provide sufficient fuel to sustain a two-hour fire). External air/oxygen sources were assumed L° .^® ^'Il'lf^ ^'TUlfl cTmbustion is limited) because a major breach of the Type B TRUPACT-II containe s or RH casks is not credible. Radioactive contamination and hazardous chemicals were assumed to be evenly distributed throughout the waste volume and 0.02 percent of the hazardous and radioactive particulate materials were postulated to be released in a respirable form (less than 10 micron particle size). Each accident was assumed to occur during a period having very stable atmospheric meteorological conditions, so as to limit dispersion or breakup of the plume and maximize radiation doses and hazardous chemical concentrations. The accident risk analysis method discussed in Subsection D.3.3 relies on the probabilistic approach in RADTRAN to determine cumulative risks of a series of increasingly less probable but more severe accident scenarios. To determine the accident consequences of the "bounding case" accident scenarios, a probabihty of 100 percent was specified. The specific conditions assumed for these bounding case accidents are summarized in Table D.3.28. The probability of breaching all Type B containers or casks during truck or rail accidents and engulfing them in a two-hour fire (requiring the fuel f q^'^f "t oHwo fully loaded fuel transports) in an urban area during adverse meteorological conditions D-90 TABLE D.3.28 Bounding case accident scenario assumptions The waste shipment is assumed to be three fully-loaded TRUPACT-lls or 1 RH cask on a combination tractor-trailer truck or six fully-loaded TRUPACT-lls or two RH casks on a railcar. The origin facilities of the waste shipments are those with the greatest likelihood of having a trailer load of waste with a curie content set at the maximum thermal or fissile gram limits specified by the WIPP Waste Acceptance Criteria or WIPP Payload Compliance Plan. All waste Is packaged in Type A drums. A major breach of any of the Type B TRUPACT-II containers or RH casks that compose a TRU shipment is not credible, limiting external air/oxygen sources. Loss of packaging containment will result in .0002 fraction of the radioactive waste material in the TRUPACT-II containers or RH casks being released to the environment in a respirable form. These respirable materials are airborne particulates and aerosols, which are all less than 1 microns aerodynamic diameter in size. Radioactive contamination is evenly distributed throughout the waste volume. The highest accident severity category, category eight, is assumed, with a fire duration of two hours. All TRUPACT-II containers or RH casks on the trailer or railcar are equally breached. The accident occurs in the urban or suburban portion of a nonspecific large (greater than one million population) metropolitan area with a mean population density of 3,861 persons (urban) or 719 persons (suburban) per square kilometer in the subarea immediately surrounding the accident site. An aerosol cloud of respirable radionuclides is dispersed downwind. to s i 2 D-91 is verv small. The probability would be a small fraction of the fraction, 0.05 x 1.5 x 10-5 /^^ 3 truck shipment or a small fraction of 0.05 x 1.0 x lO'^ for a rail shipment (Tables D 3 15 and D.3.16). Additional conservatism in the analysis included the use of a range of population densities higher than currently exist along most WIPP transportation corridors, including Atlanta, Georgia; Denver, Colorado; and Albuquerque, New Mexico. These conditions were input to the RADTRAN computer code to determine radiological consequences of these bounding cases. These radiological consequences measure the potential to cause immediate and delayed health effects in the affected population, including early fatalities, early morbidities, latent cancer fatalities, and genetic effects from the inhalation, resuspension, groundshine, and cloudshine of the aerosol cloud of the released radionuclides. As a check on estimated consequences, each bounding case scenario was also analyzed with the AIRDOS model. A comparison or RADTRAN and AIRDOS parameters for CH and RH bounding cases is shown in Tables D.3.29 and D.3.30. D. 3.4.2 Results: Bounding Case Accident The RADTRAN and AIRDOS codes were used to predict the consequences of the bounding case accident scenarios. As previously discussed, health impacts may result from external exposure (e.g., cloudshine, groundshine) and internal exposure (e.g inhalation, resuspension, and ingestion) to the dispersed radioactive material. Since it was assumed that the accidents occurred in an urban or suburban area, ingestion impacts associated with contamination of agricultural products were not applicable. The analysis assumed that stable to extremely stable atmospheric conditions predom- inated This assumption conservatively predicted high airborne radioactive contaminant concentrations and limited the dispersion of the contaminants to outlying areas. In an urban area, surface irregularities and thermal anomalies will tend to preclude the probability of a prevailing stable atmospheric condition. The revised results of the bounding case accident analyses are presented in Tables D 3 31 through D.3.34 for CH and RH truck and rail scenarios. Contributions to the total committed effective dose equivalent (CEDE) for the exposed population from van- ous pathways (initial inhalation, inhalation from resuspension processes groundshine. cloudshine) are shown as calculated by both RADTRAN and AIRDOS^ The dose expected for the maximally exposed individual as directly calculated by AIRDOS is also shown for each scenario. Population doses were converted to estimate's of heaUh effects (latent cancer fatalities) using a conversion factor of 1 person-rem - 2.8 x 10 LCFs. For all the scenarios analyzed, neither RADTRAN nor AIRDOS estimated any early fatal- ities or morbidities. The estimated population doses were dominated by inhala^on contributions (initial or from resuspension processes). Two values for the resuspended inhalation dose contribution were calculated using RADTRAN. These values were calculated using resuspension particle half-lives of 365 and 60 days and are designated D-92 TABLE D.3.29 CH bounding case accident inputs Input factor RADTRAN III AIRDOS Curies per TRUPACT-II Same for eacfi model Release fraction Release height Weather Wind speed Population density Directly calculated Pathway doses Calculation of "Maximum Individual' Directly Maximum allowed per thermal or fissile grams limits set by WAC or Payload Compliance Plan: LANL 1080 PE-Ci* SRS 1100PE-Ci INEL 1200 PE-Ci (7170 total Q) (3750 total C») (6540 total Q) Ground release (3.5 meters) .0002 released of all Ci as airborne, respirable fraction for both models Ground release Same, Stability Class F for both models 1 meter per second Same for both models (Urban: 3861 people per square kilometer Suburban: 719 people per square kilometer) 2 meters per second Inhalation Resuspension Groundshine Cloudshine Ingestion No Inhalation Groundshine Cloudshine Yes * PE-Ci is Plutonium equivalent curies calculated using weighting factors in Appendix F. Si 0^ i D-93 o o flj _ £ <) (.) ^ UJ UJ © CL Q. u. TJ o 4) 1 i m to to s- 5 CO CO CO O Q gc < E -K < I E in CO. o •D C o 5 a> Q. c to o fl> 13 0) E £ c <0 ^ ■D <0 c TJ 3 D o O O O >- a c c V T3 ■D C 3 O JO X cc o TO TO d lU _l m < JO o ■o o E o _2 Si ft a. in o o c o ■o o E o Si S (5 3 o- <0 a _« Q. o V to OO TO c t9 J3 ■D c ■D O E E to 0) Q. c X CC o o to to o 8 T5 C 3 o O o to 55 «" E to T3 C o o <0 (O -o o E o E to 0) .2 S. 2 c to XI £ V o o fl> o « a to 'to >. 5- <0 (0 & 5 to to T3 3 o CC 5 5 o CL (5 to 0) 3 ■g '> o E c _«) to .> 3 CT « E 3 o LU D-94 ^ c: E •D « O u i (d O o a o LU O) n c LU ■D () c 3 ^rf o c ^ « TJ I u O s ,. ^ to o CO 3 Q 1- lU -1 CD < .S -;^ =D E c » '" '— ' 3 (/) (O CO (0 00 2 ^ »- -^ T— ■w— i d i C) 1 d 1 d ■^ •J- (O ^_ O 0) C\j CO CO CO 00 in ^ -1 DC CJ) M^ ■r^ CJ) ^ ^ u> sl in N. CVI ^ *- 5 O) CO 03 5 00 0) s - _c « vt >jf l/V ir> •o >o sr ro c x: 'o 'o 'o 'o 'o 'o 'o 'o TJ X X X X X X X X 3 o s ? s § ?? 8 S s o <\j cvi N-' Tt cvi cvi CO CO V c o o 1 1 1 1 o o o o o o o o o o ■D *" ^ ^* *~ ^~ ^~ ^~ *~ c X X X X X X X X 3 o ■5 o ^ (D CVJ o r>. o in 'T 00 T»- "i- in CO in O c\i cvi r>-' I-' CO "-' co' CO ^o -% '% '^ ^o ^o ^o ^o . X X X X X X X X « (0 CVI CVI in N. ,_ ^ s. ^ CO in CO in CO in CM C7> _c cvi CO aS CO cvi CO CO CO -Q_ ^o '% ^o ^O cL ▼~ T— (0 X X X X 3 (0 t; 1 O 1 CO o 1 CO t DC cvi i (d cvi i cvi ' (D U1 -* in in O o O O cL *~ ▼— 3 X X X X 9 V 1 o , h- O CO q 1 CO ^ 1 tL '- 'T i ■^ ■r^ i O c o c c C8 « ^ ^ rsi c c c c C c * . (0 ^^, 1 «> o <0 ■o v> (8 c o 5 in c k- « o CL T3 (t> C w <1> j: O) .t; c 0) V) v> 3 LU CO Q o LU CO O 3 CO E "co > V) O T3 C 3 JZ CO o 0) o CO ^ o c i o in o 73 O O to ■D E SI C .2 'in c o 3 E in V (0 O T3 Q. E C o ., E in 3 V* V CO 0) in o tf> s: CO CO in v> ^ o Ol o 3 (0 c o a. II o 2 s o c in o 2 (A 3 (0 « c v> c §L 3 « s: o C3> E OJ c o in a w 3 c o Z < CC < X) CO O Q DC ir < H O DC < S D-95 > E u (U c i o ^ ■D <1> si o 3 «> « ^^ C ^_^^ a) E ffi o ^ c o <]> «> (A <)> 0) o Q. c UJ Q LU t5 c O 3 O c T ■o O o o « CvJ CO <0 fO tr Q UJ —1 CO < t2 5 D) C O o cc 0) CC Q. O Q. V CO T3 O 1^ X to CM C\J si X vO * Kl o o O O X X X X C UJ Q UJ o Q o a 3 ■o ■> ■o c o T3 E 3 E X o CO CS CM u II o E « 2 ■o c o c v Q. <0 o T- TJ > c o O fe ^ o C o c V ex o T> CO o Q CC < D-96 I E 4> Si 10 II « 3 10 « c 11) E T3 « O o c (0 o (0 U> CO <1> a o m O) n c UJ TJ () c 3 o c n o ■o I u DC o OS CO JC CO o CO 3 O K Ol _l m < <0 |2 ^ c ■D 3 _o O O fl) cc cc Q. O a. <7) T5 O CvJ d in CO s d '^ o X (£> '^ < I < cc C\J d OJ in ^ ^o ^ X X X 81 CO CVJ 8 0) '° ^o '^ X X X 81 od o 8 ro •- -* r\i O o o o ■»~ ▼~ ^~ ^- X X X X ^ in CM in s r- CO y- T- CD 00 % '^ X X o 8 r~. CJ> '^ c c c C (0 «3 «3 «J li ^ ^ ^ < I LU 111 2 U) f}' en O fe O a Q cc < ^ DC < I ■D « (0 (S 5 a 3 T3 <1> C CO jO) o if in CD § CO cu ., ^ o -C :& D) Q < DC 2 o E < o •D C o 'w c (D a. o «> o T3 c o ' T3 C o T3 I I s I o ■D U) OJ E V (U "cO ^ ■6 3 c i^rf CO (0 H c ( 1 < DC I) o ■o CO O Q < D-97 > ^.^ -n fc r 2 O 11. o S _l rr ^ u. rr O -1 i ►2 $ c ^^ £: « o o r (0 o 4> i» (/) o U a c 111 o ■5 c HI O 3 o T T3 cr a Q 9 CC < gc < CC < in CO CO c o c « c < CC r T3 0) x: C (0 3 lU Q . LU (0 () 3 ■D (0 > o T3 c c (/) 1 o ■o (0 E 3 c 0) c o < ■D 0) (0 O Q. O c o (5 t: o a « c (0 o d m _j m < J3 0) W (tJ CL o Q. w 0) s c 0) ■g o o < c o 03 CO (0 c 03 O Z o O v> I- c II (0 o ♦-' c a> ■D o o < u « W CO CO ro Z« 03 O o i2 k- c Q) Q X) F b a z C o N O 03 Li. ^ -- (SI o b b ^ O T o o o ^ •- ■^ C\J CO in o 00 CJ Tj- T- o) 00 ■»— 1— T— ■<^ O) lO t^ T- C\J T- COx> in h~. og o ^ «? o o o 0^9 o o o o ^ 9 o o o ^ cj •- T- r^ o O) in T- ■.- O CO Tj- t^ T- O ■.- ^ CJ 1- lO CO ■•- CO 1- - CO ■^ b b o XXX T- CO CO T-^ ^' cvi CO b X o o cvi in o c OJ 03 2 03 ^■§€ QC CO 3 co" O V "■!" O O O XXX in CO CO T^ CO f o b o XXX 04 CD T- T^ in ^ CO b X CO O O TT (O c 03 £1 03 2 03 = ■§■9 DC CO 3 CM b X in o 1-: CO (7) 0-7 9 o o o ■■- CO f b b o XXX ■>- CO CO 1-^ 'd^ T-" X O T-' o CO c 03 n — ^ c 03 03 LU z Q. u. cc \M =. VV CC CO 3 u. z < I 9 o X 04 o o CO CO 04 T- ^ CO o o b o X 04 CO CM CO in b b b XXX 00 CO r>«. CD in r^ o X o o o oo' 5 04 c 03 -CO = \\j ^ <.\j QC CO 3 CO cr CO CO in 00 T 9 o o XXX XXX XX 'i^pcD cocoi- r^cn T-^T^CO' oot-'co r-^cd-D b o X X '^ 0C5 1- CO TJ O OM C 05 n 03 2 03 »- -Q n 3 D -fc: DC CO 3 O O in o OvI CM CO 'f b b o XXX in 1- r*. CO 00 c\i CO 't <9 b b o XXX CO in r^ iri CO co" in b X Oi o o co" CO 04 c 03 n 03 2 03 = ■§■9 DC CO 3 z DC o 2^ I U i I D-107 »- cvj cy CJ CO IT) o T ^V ^ CM CO ^ boo boo O O O O O O O 0) CO T- 1— T— ▼- T- T- •r- T- ■>— ■t— 58 CO o Q) XXX XXX XXX XXX X 3 -.- CO CO rs. o oj O) CO 00 O O CM rs. s C cj T-= c\i CO in cvj CM T^ in ^" CD T^ in 0) p o o CM Tj. in 9 T fr T ^ CO CO T- ^ Q. LL T-* in in CO c\j in Cvi CD T^ CO CM CO -* TO C/) o5 O c CO Q. o (0 (/) ^= '•<-.» JI 5 CO CO CO f «9 Cjl f CM Q. £ ^ U. o b o o b *i >- o O Q O _J X X X X X CO c CO c:> 00 in CM CJ) <0 o o iri o o in o O -r^ O O CO CJ> SI o "S CO ? mber ( pment CO CM ^ CM o E 00 5 !E ^ '^ CO' CO Z CO t— ■o c (0 CO •a 3 Csl CO f 9 Tf <9 ^ CM pp CM C^ f <3 e c b b o o o o boo boo o <1> o 1 1- T- T- T- 1- T- 1- T- ■»- ■^ E N O O XXX XXX XXX XXX X 0) c O d CO 3 ^ in u) Oi 00 f^ CM oo in 51- rs. t CO ^ o 0) •D O (0 c\i T^ cvj CO lo

"? r> - in ▼^ (1) o S O O CO O O Tt o o ■»-= O O CO T^ J= E 'to Q. 0) E Q. CO c CL •t-f o ^ o w -Si CO 1 it: O w c 0) C *-< 0) 0) >^ B 03 C c o o < T3 CD W O Q. O c o CO t; o a w c (0 d LU —I CQ 0) w CD o c Q) ■D O X) < 0) CO CT3 x: CL C/J o S-l CT3 CD z « CD CD o (0 u. C Q) CD X3 F b Q. D sz Z V) o w CD O c ■D O O < u « Cfl CD a. *^ CO c o Id CD Z^ CD O CO »_ c ft 0) c o N O CD U. o o o o o o o o o o o o o o ^ - C\J 00 b b b 1- CvJ CO b b b »^ CM CO bob o o C\J CO ■^^ b b o XXX CD C3^ O c\i c\j cvi n b X o o oi c\j CO o o o XXX 'd- -.- 1- T^ T^ CJ) »- CO 'f b b o XXX ■>- CO CO 1-' ^ cvi CO b X O O cj 10 O XXX O O 00 CO ^' in eg b X N. o o c\i Ct) CO" o o o XXX in CO CD »- CO 'I" boo XXX W CO 1- ■r^ ui ri % X CO o o t* CO c c S CD n n t C k- CD 2 CD 5E 3 3 "i: D 3 CC CO D CC CO c CD X3 XXX CD rf O CJ CO CNJ <0 b X o o CD CD CO o o o XXX Tt O CO T-: T^ CO »- CO Tf b b o XXX T- CO CO T^ ^' T-' CO b X O T-' O CO c CD CD LU Q. LL CC VW ** VV CC CO Z) < I XXX CO CD CO T-^ 'a-' CO CM b X X CO '«t 1- COx> X in o -r^ 00 00 >- ■.- CO o b b fvj CO in O 'o 'o XXX 00 CD r>- CO in n: o X o o o CO CVI T> X> 73 CO in 00 T- CO b b XXX XX CO CO T- r^ c:) CO T-^ co' T-^ cotj b o X X 't CD T- CO "D T3 T3 T3 o C\J c c CD CD n jD *- c k- CD 2 CD 2E 3 3 i: 3 D CC CO D CC CO c CD XXX CM C:5 O T-^ Cvj CM b X cvj o o CO o cy CO 'T b b b XXX in 1- r^ CD ocj cj CO ^ CO bob XXX CO in r^ iri CO CO b X c:^ O O CD CO CVJ c CD CO CC CO z 5 z CC o c CD i3 D-109 ■D ■D D O c o o d HI _i < 0) (/) CT3 O ■g o xf < 0) Cfl CO CT3 CO O Q. Crt c o CD CD E ^ o z o i2 , c n E E .9- Z CO CO CD o c CD ■g o o < CO CD CO CD CO c g 'tD o CO c CD O O CO »_ c (1) CD, f I CD C o N »- C\J CO bob CO CO t b b o T- ^ m 'o 'c3 'a CJ C\l « b b b 'b XXX XXX XXX XXX X T- CO CO c\i T^ c\i ■.- 1^ "^ 00 T-^ T^ to T- CJ CD 1-^ CO ^ CO -^ ccj c\i c\i If) CM T lO b b b 1- «T in b b o OJ C\l '»• b b o CO CO "I- b b b o o XXX XXX XXX XXX X r^ CD r^ T^ in iri CD CO CO 1^ T^ T-: CO •>- CJ) CO ■r^ If) O C\J o CO cJ cvi ^ o o CM ^o X Oi in in b X CO O O CD 00 CD CO c CD IT) b X o O CD" CO b X O O CVJ (O CO b X O O '^ o o b o o en 00 CD OJ CO 1- b b b CO 'T CD 'o 'o 'n y- oi n b b o CJ CO "l- b b b XXX XXX XXX XXX 't in in CM T-: cvi CJ) 00 N CM CO T-: CVJ 00 in CO T^ CO xd- r- t •^ CD T-: CO in - DC CO 3 DC CO 3 DC CO 3 CO LU _J z < T3 C o CJ b X CO 00 CO CD o o X CO CO X 00 CJ b CJ5 o CM CD ■4—' o CD CO CD JZ CL ■*-' CO Q) CD sz c CD E a !c (0 CO 4—' C c CD o E N Jt CD c o O CD o 3 k— O. , CO >, E o CM > in _J z CO CD CO o CM CO CO c ■o E 0' CO CO o 'cD E CO to E CO CO CO CO o ■t; CD CO CD x: CD CO CO x: CL X 3 O ■o E D E ■^ E CL CL CD ^-f CO CO CO h_ CO k— h- o X3 CL x; CL E v— t~ CO D n o 1- b z 1- LL 03 XI o ■D -i ^li ■bo -^ e 8 8 S S 8 a a s s ^ ■%> % Si a « $ »- r^ 3 '-' 1_ i. E ^ OJ o S Si ■A O s ^ •2 O si gl 3 < 2 §1 o?< o w c <= E 2 a « 9 u) CQ O » S -J < E In I 5^^ ? Q) O o m u CD 5 L. „ 2 D (/) , Is 5 E if 15^ H ^^ ^ i 3 CO 2 D 3 3D < C3 D i= SIC u D-115 m i I z ^ u. ll < s s > 8 8 8 8 S » *-3 5 o o ^ <0 (C F h- Q 3 CL r ai c (0 _i < m < H O 3 c • i o Is go S< « 1 # se « 8 8 $ $ s s "1 5 S 2 O « o i<5 O £ a. < 2 CD O X) 5(5 o ^ if < in ^ 5V5 o ? — » 5 ^ 25 I? IS O CD ^ 13 D D OJ 3D 13 I- ? ! ! ! D-116 II c 2.2 < £ 8 X O ■» o 5R - = „ « 2 c o a < <> Zk 3 '-' (- ^ t "i> "Q^ •^ "J, "i. ■^ § s ^ .2 4^1 ■D & ur. (1) o a c o o Q. O 'ur U) it) a D D O d- _ *- CO 3 2 O J3 U « C 2 2 < ca E r J < c »{ D D M o 95 E El _- 1^ ~ « 2 oc 2 O) o _ •J s c? E i (0 o OC .tr E cr 5s s 1° ■9 05 2 2 21;: CO u i i$ D-117 m ^ s o 1^ • > < s o o X o ^s -.i o lA ^ ej |« m < h- t8?S nnual Tru ehlcle-MII travel (Vf ADT'365 <> o^i 8 8 $ $ B> o o 2 ■^ ■% "% "^ X X X X ass? "t •^ E .0 if O) o S2-CJ So Kc I ^ O UJ ^ E 25 V) o ^i >. o iS O o otr !2 -c 3 Z UJ 2 •> o " i 8 o 2 -e 2 D -e:5 3 ^ UJ O tn UJ .3° . U CO So Ji M 2 a 3 c o O S .t? c e UJ ^ z 55 UJ S D-118 i^ 11 c 2.Z < £ 5| ^ f c £ 5 9 <> -5'^ I I I I I S 8 o X 3 ^ S 8 "^ I ■fe ^ s H ae ^ # i 5 ^ f; s odd o o § § o o d Si- E 2 O) O O (0 <>5 » 8 CO D) o c O o '5 « 2| f c c o a -Q S § 3 2 o Q- ■° -9 D 3 1 5 Z o ^1 O t: 3 q E Q. O Zl o 3 o E O 3 I H o . ■o .« 2 E in-' ^ 3 2 0) E w) J3 0> o Dec a. u Z U) Q Rl j3 0> - o W Q a! Q IP D-119 ; :^^H '/.i ■>':^?; ^ U- 11 - % ^ * It \\ eo CO 'o O X X 1^ 1^ 1^ 3 S - s s a 5 "T 2 ~ E ^ « z So > c 2 5. o E i «, E o E '^ E ■«' >. c 2 2 " 5 c z o 2 E 25 i1 1 2 (0 m 2 TJ -I 25 2 " "5 m (vi ^ r) 2 CD "° >. > i^ m u — o ® ^ ir o S a 5 S 2 S •0 —UJ -So: cy ^ Q- (/) £ U- c 3 o o § § § D-120 < S Is s > 1 ? S 8 8 ° ^ 8 — « o £ c - O g s * <-> » Q — £ < 3 ^5 t T O -■ o 2 ^ •" o D x: •- - • c S § "J m E m 2 Z3 Z z o D t 8 I UJ 0. ?" 2 -I H- 2 UJ S Si ^^555 3 i ? 5>: S o > lo g in D-121 li -I 3 S < £ li S. -I < S E H 5 => o c a If in o !« » 8 »- u> * 'o o T- ^ 8 t CM in (^ r^ <6 <0 Si Rl ^ !^ ^ ^ Steffi ^ § s § Pi S 3 R S^ ^ 3 r^ ri I-' N ri cj cvj cc ecu a. a. a. (L a. 5 !2 i I I i s 8 S :2 E 55 ? E " > I S^ D 2 3 I a. 111 1 t) .E 3 o D 88 a 5 £ s § S § 5 E -Cu 2 S «2 Q. <0 EO Cj; 2 (0 3 D-122 c 2.Z < S Is e > ■bI c 5 Z 8 g C; Cli CO r-' S 2 S 5 S <\j ^ ^' jQ s s § s s s s "> ir> in 10 IS s; $ $ •BJO ^1 JS 3 .— k2 > H- a O ■ < »- »- 10 ^ a % O I. 5 ~ <^ a. cc (n (0 D (E oc a: a: a: oc oc S S o> r^ 5 o 2 a I i a 2^ IT St o 12 o> = c a 2 > ^ E 3 CD O — <0 ..: W ° uj £ I"" 55 a 52 §1 •9 «? T D-123 Ig 3 S = c d i |.z < s I? ■^ ^ ■^ 1 1. ^ ^ % ^ ^ "^ » o •h f O «> ^ ^ sP 1 # ■s^ ■t •% ^1" ^ .2 If o> o ^ 9 z w « -i o S =i " _1 Iff _i TJ E _ O 3 2 « S w < -> is to < •- If • o > O it Si - f -I ?5 ° a c « O (0 o 2 « E g" E a E° ^■2 o c c= t •• 0(00 voo =o 2Si ioo 2oi £ o 03 I o^^ so- i O O -J :^ u- D-124 s > c 2.Z < £ S > ^ ^ ^ 2 8 S ^ ^ ^ CM "1 ? h s s s cvi t c » ♦- < <> •sL J => 3 '-' H ~ ■^ •t % rt S 3 Si c^ o o o £ ^1 i V « CD o S "^ S o 5 T. t — 0) c Q 0)"2 52 ca f; ^ ri ?;2 8.S u. c 3 ,?;2 -• c £ o 1 c » :2 o> r>. o> o E o ■* o^ > T3 CO ♦- £ t! o CO §^ — ' M 0) ca si. 3 4> -C c 3 < CO o CO -D ? r ui Ol If ,1 s S 8 O) c CO « oc o o lO « O cfl CO D D I CO m CD B _!. -J ^ i ^ S > 2 in (O in io Si m ^ ^ S Si s S 3 3 D D 3 D ex. VIC I. D-125 m^ it 3 .5 11 £ 2.Z < S II -I 1 '^ %, to I I c c c c I I -55 I I I I I I c c ^ ■:< c c $ '^ -55 S 3 "b 3 8 S 5 d o c4 M g ^ X X t .2 I !« Q It a a So) t- o fAs 8 8 3 CO do: S 12 5 < S S| '< !ii CTS «o : 'o 1:1 ■fi t 5 r 8 $ P 2. Z -I "^1 - I C £ c e « < S 8 3 '-' t- ~i- ^ 5R a? d d t "b •b t s m o fa E o< OS 51 cr Q. O m < 0} ■ — E i o o ■£ "» -2 J ^ a » O c .£ o _i lU Q. . ° al" ^li 40 -ir O C 3 £ < o o eI 1° e -1 >. S c o 5 ^ oO<3 Iee S c f 3 O c E •> O S si II < = ^1 -1^ ^ s o 'o X X « If o c 5 ^ 3 = ^ lO « * • • < >• :^ P K|-j •^ •^ ^ X X X s s i8 "i. ^ § • s £ .9 It Gb u • 5 5 5 o O ?-■ 55| § cr ;: CD :s 5 cr ffi5 -J 3 -£- O 3 -" 3 . o ^ ^ 5 5. C « ^ « -S E Q- Q 0- « Q2 "S z . ^ — 3 jp .2 o n^ F - i-o a o> ^5 fr S u. > w5 I I O O 8 i^ ? D-128 II a z < c -I u-. c e c < ■X n % a 55 o 'o X s 1^ 3 ^ 13 # * r^ ^ S 3 ^ ^ ^ % % i (9 E % ■o i- o o 2 3 ■t o i t» o • £ O E -J j: «> Q o> « 11 5 o a * 01 c J3 c « • o TJ H a 8^ 5t 2 E Cu !l 3 3 o n D-129 -I ^ u. < £ -I 5^ I! •Si «b «b £ 5 CM T- * ^ s s o o ^ § P- 5 S iQ ° 8 8 8 8 5 o do d o s g g g % "c. % % X K X lo tH ? o> »- •- 3 CC CC t t W cJ Off « X o ^ X X » 3 I, 8 H .9 if " ■'J 3 Q. in CD E w w 3 >Ui §2 U a O < = 2 S ■5 = m:5£ « o 2 a s o 2 < or (0 w « S 2 ** c § ^1 2 ^-^ 5o 5 .£ O c c ^ 1^ a « I g: g I 2 5 3 8 8 8^ D-130 .'*•. 11 c 2. Z < S £ > % t> c r^ a o o o •— 1^ X X X s CO s "1^ 1^ -JJ S lo i^ s 8 s .??^ c o <> % s s •fe S S ?5 S ■^ * ^ * t^ ta 5 8 s s Si- 8 S £ 9 if oo u 2< Q. £ 2m a c — a O 6 8| -L ffi E o> 1 5 CD " ■ a 8 a C B 5^ 3 5 Is ^ a « 00 < -1. §1 s - % II -• J S c a to ^ 15 M < G S " s i 2 J55 o o .5 a -J ^ » R §>? o < o O 2 a a 8 8 w £ 15 D-131 TABLE D.4.12 Concluded Notes: Average annual truck shipments of both CH TRU TRUPACTs and RH TRU NuPac 72B casks during 20-yr Disposal Phase of the Proposed Action, going both to and from WiPP. Alternate route to preferred route. Based on the assumption on that truck equals the overall motor vehicle accident and fatality rate. 2-yr total; accident, injury, and fatality rates are 1-yr averages. 3-yr total; however, the resultant accident, injury, or fatality rate is an average 1-yr rate. New freeway segment; 3-yr of accident history not available. 1.917-yr period; based on the assumption that truck equals the overall accident rate. 2.75-yr period; accident, injury and fatality rates are an average 1-yr period. Truck value includes only combination tractor-trailer trucks. Estimated truck volume, based on typical values for given land use area. Land Use Key: R = Rural, S = Suburban or Small Urban; U = Urban N/A = Not Available D-132 TABLE D.4.13 Traffic statistics: Recent year Statewide and systemwide annual weighted averages Jurisdiction/ Route^ Accident rate/ Injury rate/ Fatality rate/ statistics source miles truck vehicle-mile truck vehicle-mile truck vehicle-mile New Mexico 888.1 7.95 X 10''' 2.97 X 10"^ 1,11 X 10*^ Colorado 312.2 1.24 X 10'^ N/A N/A Wyoming 368.3 1.26 X 10"'^ 6.10 X 10'^ 3,71 X 10"^ Utah 167.9 9.16 X 10"'' N/A N/A Idaho 368.1 7.05 X 10"'' 5,03 X 10"^ 4,26 X 10"^ Oregon 213 2.86 X 10"'' 2.08 X 10'^ 0,00 X 1 0° Washington 50.2 1,09 X 10"'^ 6.99 X 10'^ 0.00 X 10° Arizona 359.3 7.28 X 10"'' 3.73 X 10"^ 1.77 X 10'S California 625.7 6.68 X 10"'' 3.71 X 10"^ 2.53 X 10"^ Nevada 142.6 2.36 X 10"^ 9.03 X 10"^ 1.56 X 10"^ Texas 824 6.94 X 10"'' N/A 2.29 X 10"^ Oklahoma 539.1 8.02 X 10"^ 3.26 X 10'^ 2.31 X 10"^ Missouri 285.9 2.49 X 10"^ N/A N/A Illinois 428.3 N/A N/A N/A Indiana 157 N/A N/A N/A Ohio 53.2 2.16 X 10'^ 9.33 X 10"'' 1.10 X 10"^ Arkansas 285.6 6.11 X 10"'' 2,01 X 10'^ 3.16 X 10"^ Tennessee 363.8 7.46 X 10"'' 3,88 X 10"^ 1.38 X 10'^ Louisiana 188.5 1.78 X 10"^ 4,34 X 10'^ 1.26 X 10"^ Mississippi 156.1 8.03 X 10"^ 2,68 X 10"^ 3.29 X 10"' Alabama 218.1 4.81 X 10"'' 1.86 X 10"'^ 1.38 X 10"^ Georgia 212.2 1.02 X 10" "• 4,52 X 10'^ 2.03 X 10"^ South Carolina 27.4 2.03 X 10"*^ 6,02 X 10"'' 0.00 X 10° Weighted avg. 1.37 X 10"'^ 3,75 X 10"'' 1.98 X 10"^ Systemwide 6649.3 Miles 5059.3 Miles 5983.3 Miles NUREG-0170 (1977) 1.70 X 10"*^ Chem-Nuclear (1989) 1.16 X 10"^ Cashwell et al. (1986) Rural 1,33 X lO"'^ 1.09 X 10"^ Suburban 6,32 X 10"'' 2.69 X 10"2 Urban 6,16 X 10"^ 1.54 X 10"^ Only route miles for which traffic data was collected is listed. Excludes States and route segments of States where insufficient truck accident, truck injury, and truck fatality data Us ■niifi ■•iS fflfl. r, I was available, N/A = not available. D-133 TABLE D.4.14 Summary of nonradiological and nonctiemical impacts: Traffic accidents, injuries, and fatalities A. WIPP shipment-miles summary statistics: CH and RH c ombined Proposed Action Mode: 100% Truck Shipment-miles Test Phase and Disposal Phase 75,658,244 Mode: Maximum rail Test Phase (all truck) 5,139,642 Disposal Phase (rail, 8 sites) 34,506,160 Disposal Phase (truck, 2 sites) 1,529,058 Alternative Action Mode: 100% Truck Shipment-miles Test Phase and Disposal Phase 75,658,244 Mode: Maximum rail Disposal Phase (rail, 8 sites) 44,600,508 Disposal Phase (truck, 2 sites) 1 ,691 ,536 B. Comparison of WIPP lifetime risks bv traffic statistics source B.I Proposed Action - Mode: 100% Truck Statistics Source Cashwell et al. (1986) (SEIS Tables D.4.6, D.4.10) Accidents Rate/Mile Total Injuries Rate/Mile Total 92.3 Fatalities Rate/Mile Total 7.18 NUREG 0170 (1977) 1.70 X 10"*" 129. Chem-Nuclear (1989) 1.16 X 10'® 88.0 WIPP route highway system (1987-1988) 1.37 X 10"® 104. 3.75 X 10''' 28.0 1.98x10'® 1.50 B.2 Propc Accid Rate/Mile Dsed Action - Mode: Maximum rail Statistics Source ents Total Injuries Rate/Mile Total Fatalities Rate/Mile Total Cashwell et al. (1986) (SEIS Tables D.4.7, D.4.11) NUREG 0170 (1977) Test Phase (truck) 1.70 x 10'' Disposal Phase (rail, 8 sites) 1.50x10 Disposal Phase (truck, 2 sites) 1.70x10 TOTAL 28.4 2.54 -6 -6 8.74 51.8 2.60 63.1 D-134 TABLE D.4.14 Continued B. Comparison of WIPP lifetime risks by traffic statistics source B.2 Proposed Action - Mode: Maximum rail, continued Accidents Injuries Fatalities Rate/Mile Total Rate/Mile Total Rate/Mile Total Statistics Source WIPP route highway system (1987-1988)/ Fed. R.R. Admin. (1987)* Test Phase (truck) 1.37x10"® 7.04 3.75x10"'' 1.93 1.98x10"® 0.102 Disposal Phase (rail, 8 sites) Disposal Phase (truck, 2 sites) TOTAL 4.55x10"® 157.00 1.05x10'® 36.2 1.14x10"^ 3.93 1.37x10'® 2.09 3.75x10"^ 166. 0.57 1.98x10"® 0.030 38.7 Statistics Source Rate/Mile Cashwell et al. (1986) (SEIS Tables D.4.8, D.4.10) NUREG 0170 (1977) 1.70 x 10"® Chem-Nuciear (1989) 1.16 x 10"® WIPP route highway system (1987-1988) 1.37 x 10"® B.3 Alternative Action - Mode: 100% Truck Accidents Injuries Total Rate/Mile Total 92.0 129. 88.0 104. 3.75 X 10" 28.0 B.4 Alternative Action - Mode: Maximum rail Statistics Source Cashwell et al. (1986) (SEIS Tables D.4.9, D.4.11) NUREG 0170 (1977) Disposal Phase (rail, 8 sites) 1.50x10'® Disposal Phase (truck. Accidents Injuries Rate/Mile Total Rate/Mile Total 23.1 66.9 2 sites) TOTAL 1.70 x 10"® 2.88 69.8 4.06 Fatalities Rate/Mile Total 7.20 1.98x10"® 1.50 Fatalities Rate/Mile Total 2.16 •am,' I If T I D-135 TABLE D.4.14 Concluded B. Comparison of WIPP lifetime risks bv traffic stat istics source B.4 Alternative Action - Mode: Maximum rail, continued Statistics Source Accidents Rate/Mile Total Injuries Rate/Mile Total Fatalities Rate/Mile Total WIPP route highway system (1987-1988) Fed. R.R. Admin. (1987)^ Disposal Phase (rail, 8 sites) Disposal Phase (truck, 2 sites) TOTAL 4.55x10"^ 203. 1.05x10- 1.37x10'^ 2.32 3.75x10 -7 205, 46.8 1.14 xlO'^ 5.08 0.634 1.98 X 10'^ 0.0335 47.4 5.11 « See Tables 1 (p. 5) and 8 (p. 16) of reference. "Accident/Incident Bulletin No. 156, Calendar Year 1987," U.S. DOT, Federal Railroad Administration Office of Safety, July, 1988. D-136 D.4.3 RESULTS D.4.3.1 Results from Per-Shipment Risk Approach The results in Table D.4.5 show very small per shipment nonradiological and nonchemical risks for all facilities. The volumes of particulates and sulfur dioxide emitted by a single truck or rail shipment in an urban area are so small that one million or more similar pollutant generating shipments would be needed simultaneously to achieve the minimum required pollutant volume of particulates and sulfur dioxide to cause one latent cancer fatality (LCF). The probability of causing one injury from a truck accident from a single shipment ranges from 1.7 x 10"^ to 4.4 x 10"^. The probability of causing one fatality from a truck accident ranges from 4.3 x 10'^ to 3.6 x10-^. By summarizing estimated fatalities and injuries in Tables D.4.6 and D.4.10 for the Proposed Action, approximately 7 fatalities and 92 injuries were calculated for combined CH and RH shipments using 100 percent trucks. Approximately 3 fatalities and 28 injuries were calculated for combined CH and RH shipments in the Proposed Action for the maximum rail case. (See Tables D.4.7, D.4.9, and D.4.11.) Similar results for the Alternative Action were calculated from Tables D.4.8 and D.4.10. Approximately 7 fatalities and 92 injuries were estimated for combined CH and RH shipments for the 100 percent truck case. Approximately 2 fatalities and 23 injuries were estimated for combined CH and RH shipments for the maximum rail case. (See Tables D.4.9 and D.4.11.) D.4.3.2 Results from Lifetime Risk Approach Table D.4.12 summarizes traffic statistics along the WIPP preferred routes. For each segment, a description is provided of endpoints, length, average daily truck volume, population density, annual truck vehicle-miles, estimated annual TRU shipments, TRU shipments as a percentage of total miles, and annual average accident injury and fatality statistics. The route-specific truck injury and fatality rates are very low; they are usually lower than the corresponding rates from Cashwell et al. (1986), as shown in Table D.4.13. There are no segments with a recent history of relatively high injury or fatality rates which could indicate a high-hazard segment. Estimated TRU shipment volumes as a percentage of total truck volumes are extremely small for most route segments. The highest TRU shipment volume percentage is 4 percent to 5 percent for US 285 in New Mexico between 1-25, Eldorado and US 70, Roswell. Because future truck volumes will likely increase, percentages calculated are conservative upper bounds. Average State and systemwide truck accident, injury, and fatality rates compare favorably with the corresponding rates from other quoted sources (see Table D.4.13). The calculated WIPP Route Highway System Weighted Average accident rate is 1 .37 x ^-6 10'°. This is less than the rate (1.70 x 10'^) quoted by the NRC (1977) and slightly ;"3 itiflM I? D-137 -111 higher than the rate (1.16 x 10'^) experienced by Chem-Nuclear Systems, Inc. for Type B nationwide shipments. The WIPP Highway System Weighted Average injury and fatality rates are also less than the corresponding rates quoted by Cashwell et al. (1986). Consequently, statistical analyses indicate that the preferred WIPP highway routes are safer than the U.S. highway system as a whole. The SEIS analysis of nonradiological and nonchemical risks based on Cashwell et al. data is conservative. Table D.4.14 and Figure D.4.1 compare lifetime risks for 1) Proposed Action-100 percent truck, 2) Proposed Action-maximum rail, 3) Alternative Action--100 percent truck, and 4) Alternative Action-maximum rail using the two methods discussed above to estimate nonradiological and nonchemical consequences. Figure D.4.1 shows a range of forecasted estimates based on various statistics and indicates no clear difference between 100 percent truck and maximum rail modes. D.4.3.3 Comparison of Transuranic Waste Transport Accident. I niurv. and Fatality Projections In the draft SEIS, impacts were assessed for waste transport by truck (34,144 shipments) and by maximum rail (18,467 shipments) for the proposed 25-year combined Test Phase and Disposal Phase at the WIPP. Based on revisions to the overall number of projected shipments required to transport waste to the WIPP, the final SEIS estimates a total number of truck shipments (28,866 shipments) and maximum rail shipments (15,558 shipments). For the truck shipment of TRU waste, the total estimated consequences for the projected 25-year Test and Disposal Phases in the draft SEIS was 8.3 fatalities and 106 injuries for the Proposed Action, as opposed to the revised final supplement which calculated 7 fatalities and 92 injuries, respectively. The total estimated consequences for the maximum rail shipment mode for the Proposed Action in the draft supplement were 3 fatalities and 34 injuries. For this final supplement, the numbers have been revised to a projection of approximately 3 fatalities and 28 injuries. It is important to restate that the total number of injuries and fatalities projected for truck transport in the draft SEIS were calculated based on Cashwell et al. data (1986). However, only in those projections, the projected injury rate per truck vehicle-mile ranged from 6.16 x lO""^ for urban areas to 1.33 x 10'^ for rural areas. This is in contrast to the actual values that were obtained from 23 States during the preparation of this final SEIS, which indicate an overall weighted average systemwide of 3.75 x 10' . which is significantly lower than the number that was projected in the EIS (see Table D.4.1 3). Similar analyses of 100 percent truck mode fatality rates show that the Cas^hwell et al. (1986) numbers used in preparation of the SEIS ranged from 1.54 x 10' for urban areas to 1 .9 x 1 0''' for rural areas, as opposed to an overall preferred route highway system weighted average as presented based on State data of 1.98 x 10 fatalities per truck vehicle-mile of travel. D-138 Table D.4.13 also compares the accident rates used in the draft SEIS (1.70 x 10'^ accidents per truck vehicle-mile) to the State data (overall average of 1.37 x 10'^ accidents per truck vehicle-mile) supplied for the final supplement. Probabilistic risks calculated using the higher rate (1.70 x 10"^) from the NRC (NRC, 1977) are thus conservative given expected lower numbers of accidents based on actual route-specific data. Table D.4.15 summarizes data on radioactive material shipments. The data was compiled from actual shipping records supplied by private sector radioactive waste transporters and the Department of Energy/Albuquerque Operations. As shown, the industry and the DOE have compiled an excellent safety record for shipping radioactive materials. The use of certified TRUPACT shipping containers and casks for TRU shipments and the extensive system of oversight and management developed for these shipments ensure that transportation risks for the Proposed Action or Alternative Action will be comparable, if not less, than those in similar shipping campaigns, as shown in Table D.4.15. I'*:: I? D-139 I! HI I TABLE D.4.15 Comparison of radioactive material shipments Source Total Number of Accidents/ mileage shipments incidents Injuries Fatalities SEIS Truck 74 million^ Rail 30 million 28,866 1 5,558 Chem-Nuclear*^ Truck 26 million NR Spectra Research/SNL^ Truck NR Rail NR 2,000,00 NR° NR 92 25 7 3 828 25 NR NR NR NR DOE/Albuquerque Truck 30.8 ^ The total estimated mileage was not presented in the SEIS, the total estimated mileage represents a 25-year shipping campaign. b NR = Not reported. ^ Reporting period of 1987-1988. ^ Reporting period of 1971-1988. ® The number of shipments were not broken down in truck and rail. ^ Fatalities, but not attributable to project. D-140 t::^i;y. PROPOSED ACTION Accidents MODE 100% Truck m^, 88. 1 29 Max. Rail mmm> 63. 1 1( 36 50 100 150 200 Injuries MODE 100% Truck Max. Rail 28.0 92.3 25.2 38.7 — I — 20 40 60 Fatalities — I I 80 92 MODE 100% Truck Max. Rail 1.50 7.20 2.54 4.06 -I 1 r-^ p — I 1 1— 01 2345678 ALTERNATIVE ACTION Accidents MODE 100% Truck VM 88.0 129 ^^999S^$^^9S^S^^ >^ Max. Rail >>y>o<>>y^c< /VV- 69.8 ■ 1 1 205 MODE 100% Truck Max. Rail 1.50 7.20 2.16 5.11 01 2345678 50 100 150 200 MODE Injuries 100% Truck W////////////////M 28.0 92.3 Max. Rail 8^^ 23 .1 47.4 1 1 20 40 60 80 92 Fatalities FIGURE D.4.1 LIFETIME NONRADIOLOGICAL AND NONCHEMICAL TRANSPORTATION RISKS: RANGES OF PROJECTIONS FOR OH AND RH SHIPMENTS D-141 0] 4 c'ic^F^ REFERENCES FOR APPENDIX D AEC (Atomic Energy Commission), December, 1972. Environmental Survey of Transportation of Radioactive Materials, to and from Nuclear Power P lants, WASH-1238, Directorate of Regulatory Standards, Washington, D.C. Alexander et al. (C. A Alexander, J. S. Ogden, and L Chan), 1986. "Actinide Release from Irradiated Fuel at High Temperature," IAEA-SM-281/10, Source Term Evaluations for Accident Conditions , from the Proceedings of a Symposium in Columbus, Ohio, for the International Atomic Energy Agency, Vienna. Cashwell J W et al., 1986. Transportation Impacts of t he Commercial Radioactive Waste Management Program . SAND 85-2715, TTC-0663, Sandia National Laboratories, Albuquerque, New Mexico. Chem-Nuclear Systems, Inc., 1989. Letter, TR-89-T055, dated May 25, 1989, from Leonard Toner to John Arthur, DOE-AL for WIPP SEIS, "Chem-Nuclear Systems Accident Analysis for Radioactive Material Shipments: 1987 and 1988." DOE (US Department of Energy). 1989. TRU Waste Acceptance Criteria for the Waste Isolation Pilot Plant . WIPP-DOE-069, Rev. 3, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1986. Transportation Assessment and Guidance Report . DOE/JIO-002, Rev. 1 , Albuquerque, New Mexico. DOE (U S Department of Energy), 1980. Final Environmental Impact Statement Waste Isolation Pilot Plant . DOE-EIS-0026, Vols. 1 and 2, Carlsbad, New Mexico. FRA (Federal Railroad Administration), 1988. Accident/Incident Bulletin No. 156: Calendar Year 1987 . Office of Safety, U.S. Department of Transportation, Washington, D.C. Huerta M et al., 1983. Analvsis. Scal e Modeling, and Full-Scale Test of Low-Level Nuclear Waste Drum Response to Accide nt Environments, SAND80-2517, Sandia National Laboratories, Albuquerque, New Mexico. Jov D S et al 1982 Hiohwav. a Transportation R outing Model: Program Description " and User's Manual . ORNLTTM 84-19, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Lorenz et al. (R. A. Lorenz. J. L Collins, A. P. Malinauskas, O. L Ki^kland and R. L Towns) April 1980. Fission Product Release Fro m Hiahlv Irradiated LWR Fuel, NUREG/CR-0772, ORNUNUREG^M-287/R2, Oak Ridge National Laboratory, Oak Ridge, Tennessee. D-142 Madsen et al. (M. M. Madsen, E. L Wilmot, and J. M. Taylor), 1983. RADTRAN II User j Guide . SAND82-2681, TTC-0399, Sandia National Laboratories, Albuquerque, New Mexico. Mishima, J., and L C. Schwendiman, 1973a. Some Experimental Measurements of | Airborne Uranium (Representing Plutonium) in Transportation Accidents . BNWL- 1732, Pacific Northwest Laboratories. Mishima, J., and L C. Schwendiman, 1973b. Fractional Airborne Release of Uranium j (Representing Plutonium) During the Burning of Contaminated Wastes . BNWL- 1730, Pacific Northwest Laboratories. Fischer et al. (L E. Fischer, C. K. Choy, M. A. Gerhard, C. Y. Kimura, R. W. Martin, \ R. W. Mensing, M. E. Mount, and M. C. Witte), 1987. Shipping Container [ Response to Severe Highway and Railway Accident Conditions . NUREG/CR- 4829, Vols. 1 and 2, Lawrence Livermore National Laboratory, UVermore, j California. NRC (Nuclear Regulatory Commission), 1980. Accident Generated Particulate Materials and Their Characteristics - A Review of Background Information . NUREG/CR- 2651, PNL-4154, Pacific Northwest Laboratory. NRC (Nuclear Regulatory Commission), 1977. Final Environmental Statement on the Transportation of Radioactive Material bv Air and Other Modes . NUREG-0170, Vol. 1 , Office of Standards Development, Washington, D.C. ORNL (Oak Ridge National Laboratory), 1988. Integrated Data Base for 1988: Spent Fuel and Radioactive Waste Inventories. Projections, and Characteristics . DOE/RW-0006, Rev. 4, for U.S. DOE. ORNL (Oak Ridge National Laboratory), 1987. Integrated Data Base for 1 987: Spent Fuel and Radioactive Waste Inventories. Proiections. and Characteristics . DOE/RW-0006, Rev. 3, for U.S. Department of Energy. Peterson, B. E., 1984. INTERLINE: A Railroad Routing Model . ORNLyTM-8944, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Rand McNally, 1988. Rand McNallv Handy Railroad Atlas of the United States . Rand McNally & Company, 1988. Rand McNally, 1987. Rand McNallv Road Atlas . Rand McNally & Company, 1987. Rao et al. (R.K. Rao, E. L Wilmot, and R. E. Luna), 1982. Nonradiological Impacts of Transporting Radioactive Material . SAND81-1703, TTC-0236, Sandia National Laboratories, Albuquerque, New Mexico. Shirley, C. G., 1983. "A Simple Risk Analysis Tool for Estimating Release Fractions From Shipments of 55-Gallon Drums Involved in Transportation Accidents," ii!? I? D-143 Seventh International Symposium on Packagi ng and Transportation of I Radioactive Materials (PATRAM 83) . New Orleans, Louisiana, pp. 160-167. ' Taylor J M and S. L Daniel, 1982. RADTRAN II: R evised Computer Code to Analyze ' ' Transportation of Radioactive Material . SAND80-1943, TTC-0239, Sandia National Laboratories, Albuquerque, New Mexico. Taylor J M and S. L Daniel, 1977. RADTRAN: A Computer Code to An alyze ' Transportation of Radioactive Material . SAND76-0243, Sandia National Laboratories, Albuquerque, New Mexico. Wilmot EL M. M. Madsen, J. W. Cashwell, and D. S. Joy. 1983. A Preliminary Analysis of the Cost and Risk of Transpo rting Nuclear Waste to Potential , Candidate Commercial Repository Sites . SAND83-0867, TTC-0434, Sandia ' National Laboratories, Albuquerque, New Mexico. Wolff, T. A., 1984. The Transportation of Nuclear Materials . SAND84-0062, TTC-0471, Sandia National Laboratory, Albuquerque, New Mexico. ' Wooden, D. G., 1986. Railroad Transportation of Spent Fuel , SAND86-7083, TTC-0574, i Sandia National Laboratories, Albuquerque, New Mexico. D-144 APPENDIX E HYDRAULIC AND GEOTECHNICAL MEASUREMENTS AT THE WIPP HORIZON J13 Si 'Ac I. 'iiitC % •"'3 7] I E-i/Ji m •.>>,y- TABLE OF CONTENTS Section paqg E.1 INTRODUCTION E-1 E.2 BRINE INFLOW MEASUREMENTS (Deal and Case, 1987) E-3 E.3 BRINE INFLOW MODEL (Nowak et al, 1988) E-47 E.4 WIPP HORIZON GAS FLOW MEASUREMENT RESULTS SUMMARY THROUGH 1986 (Stormont et al., 1987) E-1 07 E.5 1984 GAS FLOW MEASUREMENT TEST RESULTS (Peterson et al., 1985) E-1 33 E.6 N1420 DRIFT GAS FLOW DATA ANALYSIS AND EVALUATION (Peterson et al., 1987) E-1 71 E.7 WASTE-HANDLING SHAFT PULSE TESTING DATA SUMMARY AND CONCLUSIONS (Saulnier and Avis, 1988) E-1 99 E.8 DELINEATION OF THE DISTURBED ROCK ZONE (DRZ) E-285 (Borns and Stormont, 1989) E.9 SEAL DESIGN AND EVALUATION (Stormont, 1988) E-297 E.10 REFERENCES FOR APPENDIX E E-363 He f I? E-ili/iv m -^y.y. :<^, E.1 INTRODUCTION Appendix E contains excerpts from published documents that primarily support conclusions regarding the hydraulic and geotechnical characteristics of the Salado Formation. This appendix is not intended to provide a complete understanding of the various studies, but is intended to provide enough data and interpretation to provide the reader with an adequate level of information to independently assess the conclusions presented in the text. In this final SEIS, the introductions to all sections (E.1 through E.7) are published, as well as a modified Section E.3; a new Sections E.8, Delineation of the Disturbed Rock Zone (DRZ); and a new Section E.9, Seal Design and Evaluation. The reader is referred to the draft SEIS for the complete sections E.1, E.2, and E.4 through E.7, which remain unchanged. lii: as a] ■Xii "Si I? 7 i E-1/2 t-,»f E.2 BRINE INFLOW MEASUREMENTS This subsection of Appendix E describes preliminary sampling and evaluations of brine occurrences at the WIPP facility horizon. Included is a discussion and description of sampling methodology, the manner in which the data were used, calculations made, and a location-by-location description of sampling results. This subsection was excerpted from Appendix D of Deal and Case, 1987, Brine Sampling and Evaluation Prooram. Phase I Report . This subsection is included to provide evidence of brine inflow rates defined in the text. ^ E-3/4 m E.3 BRINE INFLOW MODEL This subsection of Appendix E presents and describes the WIPP Darcian Brine Flow Model that has been used to analyze brine inflow rates to observed boreholes and moisture release experiments and is provided here to support brine inflow rates defined in the text. Included in this section are the assumptions inherent in the model. This subsection has been excerpted from Chapters 2 through 6 of Nowak et a!., 1988, Brine Inflo w to WIPP Disposal Rooms: Data. Modeling, and Assessment . Sections specifically related to nonisothermal flow have been deleted. The nonisothermal aspect of the model was used to simulate inflow due to heat generated by high-level waste. Since high-level waste will not be disposed of at the WIPP, these sections are no longer pertinent. Some reference with respect to nonisothermal conditions is left in portions of the text to provide more generic aspects of the model development. "HI* ""4 E-47/48 m •Ml ■'>/>' 2. WIPP BRINE FLOW MODEL A brine transport model for both isothermal and non-isothermal conditions in bedded salt was developed with data from the W PP Targe scale in situ experiments [25.28.29]. This model is for transient OarcJ flow n "s?oraaf-'n^r- ^IW ''''''''' '^ ''' "^^ ^"^ brine a CO n'fo ^th " storage of brine that supports transient flow, and thermal effects are accounted for by including the thermal expansioi of the brine and the host Any model for transient flow of fluid in a porous medium requires the stipulaiion of a mechan sm of **storaqe/' that is lorai rhLllc 171^-1 ^ass per unit volume of the medium. In'a r ;o 'm d urn ' e oni?^' JIuJ '''^fnT 'J^'" i? compression, or the iScal density change of e Jn ?;t Jn «A'k°""'? 5 P?Tf '"'^^"'"' ^^°'*^9« "" be accommodated by wl D??a?It nn n?^ h '^'^'^'' ?"? I'^'^ compression of the solid, as ?n !;.c? n c! ?" °I ^f^e porous skeleton Is the principal mechanism of nterest in soi and rock mechanics, and is the cornerstone of consoidation" theory. Rock salt, of course, exhibits plastic as well a e astic properties It is. however, plausible that the im^edia e elastic response of the salt and brine and the subsequent relaxation of the ^HnV:?''"'' ^^w^ °^ ^° ^^' excavation are the predominant mechan mso brine storage and transport, at least over short time scales. For a linearly elastic skeleton. Biot [36] generalized the w??h° Jrf!nM/^'°'^ J"^.*^^".^"^ Cleary [37] later recast it in terms with straightforward physical interpretations. An extension of this model he'nu°idMH'.n?-;^'K''^r'^ '''''''^ ^^^°^^'"9 for thermal expan of the fluid and solid, has been presented recently [38.39]. The essence of the model Is embodied in a diffusion equation for the pore pressure that, in certain special cases, reduces to: ap ae — - cy'p - b'— , at at (1) where p is the fluid pore pressure, c is the fluid diffusivity. b' is a source coefficient, and e is the temperature. The fluid diffus vity c depends upon the permeability, fluid viscosity, and the elastic properties HlnlnSc'^^'"* I'J^ 11"'^ "' '^PP'"^^^ ^y- ^^' 5°"»*« coefficient, b'! depends upon the therma expansivities of the solid and fluid (see Appendix A . For isothermal conditions, the right hand side of (1) vanishes. aSd the r tprw?L?y^"''°!: '"^'Z^'': l'^^'^'^ ^^°^ ^^ recovered! Various pecia cases wdely considered in hydrologic modeling are embedded in this formulation [25]. For non-isothermal problems in which conduction heat transfer dominates (i.e., small Peclet number), as is certainly true in salt, the source term in (1). which represents the generation of pore Jnfnnln rH Jt^^J^l ^xpansi on . must be evaluated from the simultaneous solution of the heat equation: ;5 "IX [9 i E-49 de dt - kV e - , (2) where k Is the thermal diffusivity. Extended discussion of this system of equations as well as various solutions to representative initial value problems can be found in [38,39]. The explicit relationships between properties of salt and brine and the coefficients appearing in equations (1) and (2) are given in APPENDIX A if this report. The host rocic salt permeability, k, and other properties of the host rock and brine appear in the fluid diffusivity, c. In the data analyses that are discussed here, permeability values, k, were chosen to match or bracket the brine inflow data. Other values for the host rock and brine properties in the diffusivity, c, were taken from known properties of salt and saturated brines. The permeability values thus obtained were used to calculate brine inflow to WIPP disposal rooms with this model. Si all! III! Ml 2.1. Isothermal Flow Consider now an idealized model for the introduction of a mined drift into a deeply buried region. The rock is assumed to be homogeneous and isotropic, and the undisturbed stress state is taken to be lithostatic, i.e., isotropic, compressive, and equal in magnitude to the overburden load. The initial pore pressure in the neighborhood of the tunnel is assumed to be constant: P(r,0) (3) The pressure Po is expected to be between hydrostatic (about 6 MPa) and lithostatic (about 15 MPa) [25); this has been corroborated by field measurements from which pore pressures of 8.3 MPa and 10.3 MPa were estimated [31]. Superimposed on the hydrostatic pressure 1$ a portion of the Increased mean stress Induced by the presence of the tunnel. The fluid pressure then relaxes by Darcy flow toward the tunnel, and the load 1$ transferred to the solid skeleton. The pressure field corresponding to this sequence is governed by (1) with the right-hand side zero and with the Initial condition (3) and boundary conditions: p(a,t)-0 , lim p(r,t) - p^ r -» • (4) (5) E-50 where a Is the excavation or borehole radi thl fK. n ;! "cavanon or Dorehole radius. Equation (4) simply states at tL nh r r ' ^''' ^° ^^°" ^° ^^' ^drained'' face/whkh Is^main n ai aimospnerlc pressure. ed Jh! Jn^"S^°" I? ^^! '"^^^^ ^° <5) ^5 ^ell »^nown (e.g., [40]); the flux at dIfferenUauJJ- ''^'•^*^' ^°^^°*'' Inmedlately from olrcy's lavi by q(a.tj kp^4 pa ir exp(-u^t^) du 05{") + Y'{u) u (6) where k Is the permeability, p is the fluid viscosity, t* - ct/a2 is the normalized time and Jo(x) and Yo(x) are zero-order B^ssel fSnctio of the first and second kind, respectively. Note that the sign of the flux is negative because it is in the (-r) direction. It is convenient a so to introduce the asymptotic expansion for early time- lim q(a,t*) • - — ? t*- pa 1 A 2 4 — t-^/2 + -^^ tl/2 . ^ , ^* + 7 • — t/ + - t* + . . . . (7) and that for late time: t* lim q(a.tj kp. //a ln(4tj - 27 [ln(4tj - 27] ■X + • • • (8) ^6? 7fil tllll ^^^"lr s constant. Values calculated with equations l!;iw J-^ !^T '" '^^Sure 1. Note that the flux falls off rapidly at early time, and changes only slowly for t* > 10 "iii ilia ?l IPJ id E-51 "-i^^rwr; •an' ' •m 1 III' I •HI w" I rl "i T — I » T I i r y f i I n| 1 1 — I — I 1 i» i| 1 1 — I » lilt LATE TIME APPROX. ■ .. ■■■■ I ' ^ I I I I i li I — I I I t II J 10 •1 10' 10' 10' 10 DIMENSIONLESS TIME, t/(aVc) Figure 1. Flux to a circular tunnel or borehole. E-52 2.3. Assumptions Inherent In the Hodel It 1$ our judgement that uncertainties associated with the assumptions in the model introduce uncertainty in brine inflow predictions for waste disposal rooms of no more than about an order-of-magnitude. Both the Darcy model itself and some of the assumptions invoked in order to represent the practical problems of interest are Idealizations of very complex systems. It can be anticipated that some of these idealizations are conservative, in the sense that they tend to lead to overpredictions of brine flow at the ' WIPP, and some are 'liberal," in the sense that they probably lead to underpredictions. The directions of uncertainties that may arise from some of the other model assumptions are difficult to assess at this time. Assumptions that are likely to lead to overpredictions of brine inflow (conservative) include the following: There exists a network of interconnected porosity extending outward without bound. This assumption implies a limitless reservoir of brine. The far-field brine pressure is lithostatic. Aside from the stress perturbation due to the presence of the excavations, it is difficult to imagine a mechanism by which the pressure could rise above lithostatic. Brine flow is radially symmetric (two dimensional). The effect of the third dimension 1$ to weaken the flow by geometric spreading of the disturbance. The backpressure from the room contents is negligible. Any backpressure due to interaction of the salt with solid, fluid, or gas in the storage room will mitigate the flow to the room. Inelastic dilatation of the salt 1$ neglected (see also below). Dilatation of the salt near the excavations due to Inelastic mechanisms, such as opening grain boundaries, tends to decrease the pore pressures that drive flow. Assumptions that are likely to lead to underpredictions of brine Inflow ("liberal') include the following: The storage of available brine in the host rock Is due entirely to elastic compression of the brine and salt. Additional (inelastic) storage mechanisms would decrease the brine diffusivity and, therefore. Increase the decay time for the flux. Thus, Integrated fluxes over long time would be larger. The magnitude of the Initial (maximum) flux, however. Is unaffected by the storage. ; !i« iiil •m id d ■""3 E-53 Inelastic dilatation of the salt Is neglected (see also above). Dilatation of the salt near the excavations due to Inelastic mechanisms such as opening grain boundaries tends to Increase the permeability in that region. However, calculations that account for extreme Increases In permeability near the wall (given in Section 4.3.4 of this report) show relatively small increases in the cumulative brine flux, because the flow over long periods of time is controlled by the far-field properties. The directions of uncertainties about possible effects of inelastic, volumetric deformations and of heterogeneities in the host rock salt are now difficult to assess. Such effects have not been the focus of the laboratory testing program for host rocic salt. Also, the effects of heterogeneity are difficult to anticipate. Some further worit to reduce these uncertainties will be described below. However, these effects on predicted brine inflow values are not expected to exceed an order of magnitude. •Hi . IMP ' ■m E-54 ■.>f>y 3. WIPP BRINE FLOW CHARACTERISTICS DATA BASE Data pertinent to WIPP brine inflow predictions are available from several sources. Brine accumulations were measured by periodic bailing in boreholes located over a wide area of the WIPP facility. These measurements were part of the WIPP Brine Sampling and Evaluation Program [30]. Brine inflow rates were also calculated from moisture release data obtained from isothermal and heated boreholes In the Moisture Release Experiment for Rooms Al and 6 in the WIPP [29]. Host rock permeability values are available from WIPP in situ brine and gas flow measurements that support the WIPP Plugging and Sealing Program [31]. The data from these sources are described in the following sections. 3.1. WIPP Brine Samplino Data Deal and Case [30] monitored 54 drillholes throughout the WIPP, most of them for about 500 days. They show graphical results for the time histories of the total flux for 20 holes. The flow rates to two of the holes, BX02 and 0H37, fell essentially to zero after 600 days. The flow rates to the remaining 18 holes at the end of the reporting period are considered here (Table 1). Hole A1X02 exhibited a nearly monotonic decay in flow rate for nearly 400 days, but then experienced a steady increase in flow rate. The value entered in Table 1 for A1X02 hole corresponds to the value at the end of the period of declining rate. The recorded flow rates represent the integrated flux over the borehole surface areas, and are recorded in Table 1 in units of liters per day, i.e., a volume flow rate. n '•'lilt {9 T, E-55 Sill S: (Ill III •Ml •ni mi k:: u, Hole Flow Rate Area Radius number l/day in2 m IG202 0.014 5.20 0.0572 IG201 0.025 5.91 0.0572 NG252 0.250 0.26 0.0190 AlXOl 0.026 4.84 0.0508 A1X02 0.010 5.74 0.0508 A2X01 0.025 4.87 0.0508 A2X02 0.015 5.13 0.0508 A3X01 0.023 4.91 0.0508 A3X02 0.001 4.93 0.0508 BXOl 0.055 4.87 0.0508 DH36 0.250 4.38 0.0444 0H38 0.055 4.04 0.0444 0H40 0.005 4.34 0.0444 DH42 0.030 4.35 0.0444 DH42A 0.095 3.44 0.0444 DH35 0.002 4.42 0.0444 LIXOO 0.028 3.72 0.0380 0H215 0.004 1.22 0.0508 Table 1. Observed flow rates for WIPP boreholes [30] E-56 3.3. WIPP Host Rock Permeabilities from Independent In Situ Flow Measurement^ Permeability values In the range of 10-21 to 10-20 m2 (j to 10 nanodarcy) or lower have been derived for intact WIPP host rock from independent in situ measurements of brine flow during fluid transport experiments [31,42,43]. Independent measurements of the both gas and brine permeability of the salt at the WIPP facility horizon have been made using constant-pressure and pressure-decay methods in 6.5 cm radius boreholes [31,32,42J. These tests showed that permeabilities near the drift wall were mostly of the order of 10-20 to 10-18 ^2 (10 to 1000 nanodarcy) or higher in some cases. ^ few meters into the wall , permeabilities were of the order of 10-22 to lO-^O^mZ (0.1 to 10.0 nanodarcy). Measurements in the WIPP waste-handling shaft at levels above the proposed disposal horizon confirm the range of 10-21 to 10-20 ^2 (i to 10 nanodarcy) for undisturbed host rock salt [43]. The permeability range implied by comparisons between model calculations and brine inflow measurements will be compared with these results. 3.4. Data Reduction 3.4.1. Radial Darcy Flow Model for Isothermal Data Reduction An idealized model was Introduced previously [25] to investigate the order-of-magnitude agreement of observed fluxes with the proposed Darcy flow mechanism. This model was described above. In particular, it was assumed that mined faces and boreholes introduce zero-pressure surfaces into a region of porous salt in which the brine is initially at hydrostatic pressure. (It is easy to argue that the initial pressure may be as large as lithostatic, but this changes the initial conditions only by a factor of about two. The uncertainty in the permeability is expected to be much greater.) In this case, the Darcy flux, q, to a circular borehole scales in the following fashion [25]: ""f :Cii> rtd* (9 z q a , pa (15) where k is the permeability, Po is the Initial pressure, p is the brine viscosity, and a is the borehole radius. This factor is multiplied by a E-57 lime-dependent function of order unity that represents the decay of the flux as the pressure disturbance propagates away from the hole. The characteristic time over which this decay takes place, to, is given by: to ■ c" » (16) where c Is the fluid diffuslvity. For elastic rock, the fluid diffusivity scales like: C - — (H) where K is an elastic modulus for the porous skeleton. It can be argued from the model that, for WIPP salt, the appropriate modulus and viscosity yield a diffusivity of the order of c - 1.1 X lOl^k n2/s , (18) 531; 1 •«l;:| ("' Ml Ml,. II,. where the permeability is given in units of m2. Previous calculations [29,39] suggest that the brine diffusivity is of the order of 10-7 m2/s. For a borehole of radius 0.05 m, then equation (16) gives a characteristic time of the order of 2.5 x 10^ s, or about levin hours! Therefore, after 500 days, the drillholes in the WIPP can be IxpeaeS to be in the asymptotic limit of -late" time, n this case, the fl5x can be approximated bj the first term in the series given by equation (8): iQl • — T. (19) pa ln(4ct/a') - Zi where 7 - 0.57722 is Euler's constant. 3.4.2. p^rmeabnitips from Br ^np Sampling Data Deal and Case [30] report the dimensions of the holes from which they collected and measured brine, so that it is simple to calculate the vertical wall area of each. These values are recorded n Table 2. The averlgroarcy flux (or "Darcy velocity") for each hole is easily calcu ated bHd the integrated volume flux by the total bore e area. Th s step is not taken here, because the comparison can be •" pleading. If the flow does occur by a Darcy mechanism, then the Darcy velocity is expected to sca^e inverselj with the borehole radius. Thus, the appropriate measure f?r a hole to'ho e comparison in this context is the product of the Darcy E-58 flux and the borehole radius. The values of this product apoear In the fifth column of Table 2, labeled "qa". The values for the product of the Darcy flux tirr.es borehole radius which are proportional to the total flow rates per unit length of borehole center around 3 x lO'lZ mZ/s. The oiaximum value is for hole NG252, at 2 1* X 10-10 njZ/s. This hole samples an anomaly in the WIPP host rock; consequences of this anomaly will b€ discussed below. The apparent permeability was calculated for each borehole using values for 'qa", the Darcy flux times the borehole radius, and equation (19). In particular, it was assumed that the initial pressure is po - 6 X 10^ Pa, corresponding approximately to hydrostatic pressure for a depth of 600 m. The brine viscosity is taken to be 1.6 x 10*3 Pa-s. The time was assumed to be t • 4.32 x 10^ s (500 days) for ew^ry hole The diffusivity was assumed to be given by (18). Finally, for each drillhole, values for the flux times the radius, qa, are known (Table 2). Thus, the only unknown parameter Is the apparent permeability, kaoo The explicit relationships between the properties of salt and brine and the coefficients appearing in the above relationships are given in APPENDIX A of this report. Also given in that appendix are the typical properties for WIPP salt that were used. The nonlinear relationship for "kapn", represented by equations (18) and (19) Is then easily solved numerically. The results of this exercise are shown in the last column of Table 2. The values shown may be read directly as nanodarcies (10-21 mZ - 1 nd). Figure 2 shows a histogram of the logarithm of the apparent permeabilities given In Table 1. The mean of the log is -20.45 (k • 3.5 x 10-21 m2, or about 3.5 nanodarcy), and the standard deviation of the logarithia of kjpp is 0.81. Also shown is the lognormal distribution corresponding to these values. These limited data and the highly idealized model suggest a lognormal distribution for the apparent permeability This is a common observation in other rocks. The highest value of apparent permeability shown in Figure 2, *:* ? !° : ;:• ^^ l^J^fly to be anomalously high. That datum represents the brine inflow rate to borehole NG252, i borehole that is known to Intersect a horizontal fracture associated with Marker Bed 139 [301 Thus the ideal smooth borehole model from which the apparent permeability was * calculated can be expected to yield an anomalous value that does not correctly characterize the host rock salt. A fracture can Introduce a large surface area for inflow; if this flow is then averaged over the borehole wall area only, the calculated flux and the apparent permeability will be erroneously large. A model that accounts explicitly for flow to both the borehole and a large intercepted fracture should yield a more nearly representative value for the apparent permeability. For example an order-of-Biagnitude estimate of the additional inflow from a 12 m radius crack with a )i6ry small aperture yields an apparent permeability of 10-20 mZ, a value that is in better agreement with the other permeability values hi (9 J*' z E-59 51111 '< o o o z 1 •23.0 -22.5 log(k) I'UlIC Id ■1154 •ISJ '9 Figure 2. Apparent permeabilities based on BSEP data. E-61 •"til; Kill •nil 3.4.3. Permeabilities from Isothermal Mo isture Release Data Before the heaters were turned on in the Instrumented boreholes in Rooms Al and B, moisture was collected in all four holes for a few days [261. The integrated mass flow rates were in the range of 5 to 5 g/^^y» which, averaged over the borehole area, corresponds to a Oarcy flux of 0.8b to 2.6 X 10-H m/s. The product of the flux times the borehole radius, a - 0.38 m, is then in the range: Pa - 3.2 to 9.9 x 10-12 n,Z/, In comparing these values to those calculated from the IT measurements (Table 2), it should be noted that the latter represent flows at much later time (t » to). The apparent permeabilities for the moisture-release holes were calculated in a fashion similar to the aporoach used above, and the resulting values were In the range of 10-21 jn2 to 10 20 pZ In this case, however, the flow rates measured In the pre-heating stage do not reflect very late time, and the asymptotic solution, equation (19), is not accurate Usina the full integral solution (61. the same initial "ondUion. po " 6 ;i06 Pa, and t - 2.1 x 10^ s (8 jonths the observed ranae of fluxes requires permeabilities in the range K - ^.'» lo . S 5 X 10-21 m" These values are quite consistent with those required to represent the IT data (Figure 2), and, again, are consistent with Independently measured in situ permeabilities l31,4Z,4ij. It should be noted that these permeability values are our best estimate so far and represent a significant Improvement over jn interim ttudy [25 In that study. It was assumed that the test boreholes for the W 'iislure'reieale experiments simply intercepted brine fow to he est rooms (WIPP Rooms Al and B). From the scaling relation for the Darcy flux Jo a circular hole or tunnel (equation 15). the permeabllty expected to scale like k - q^a/po, where "a" Is the appropriate length scale. The h scale fo?'the iest rooms Is 3 5m; for the test Jore o es u i 4 m. Therefore, the apparent permeabilities reported in the in^erm study a e about an irder of magnitude larger than the ^PPf J^ P^^^^^^ ^^ ,3 calculated here Here, the length scale used is the test borehole radius of 8 m This scale is appropriate for the model, because the pressure !e?d I the neighborhood of th test room ^^ould change re a ively s owly. and flow to the boreholes should respond primarily to the local Pressure field around the borehole. Time scales for excavations are given in terms of radius and diffuslvlty In equation 16. E-62 :-:-v^.;. ■>4 3.4.4.2. Salt Block II Experiment The Salt Block II experiment [11] was performed some ten years ago in support of the WIPP project. In this experiment, a right circular cylinder of salt, 1 m long and 1 tn in diameter, was obtained from a potash mine near Carlsbad. A 13 cm diameter borehole was located on the axis of the cylinder. An electric resistance heater was placed in the borehole, and the heater power was stepped up over a range of 0.2 to 1.5 kW, with each power level held for a period of several days. The fluid driven to the borehole was collected in a low-pressure dry gas stream and absorbed externally in a desiccant. Temperatures interior to the block were monitored by an array of thermocouples. A one-dimensional idealization of the Salt Block II configuration has been modeled [44] using the "porothermoelasticity" theory described in Section 2. of this report. The block is assumed to be at a constant initial temperature, and the initial excess pore pressure is taken to be zero. The heat flux at the borehole is represented by a linear ramp up to a constant value for each stage of the experiment. The heat flux at the outer boundary is represented by a heat transfer coefficient. The pore pressure at the borehole is taken to be zero, and the outer jacket is assumed to be impermeable, so that the pressure gradient vanishes there. The radial normal stress is zero at both the inner and outer radii. The coupled heat transfer, fluid flow, and solid deformation problem reduces, in this configuration, to a pair of diffusion equations for the temperature and fluid pressure. The equations are nonlinear, because the model allows for temperature-dependent properties, including the thermal conductivity and brine viscosity. The problem is solved numerically by the method of lines. The numerical solver is coupled seeks the set of specified parameters experimental data. In this case, for fitted by the solution to the conduct thermal conductivity and the heat tra boundary. These values are then used flow, with the fluid diffusivity and unknown. Here, the calculated fluid the experimental measurements. to a parameter-estimation code that that results in the best fit to the example, the thermocouple data are ion calculation to determine the nsfer coefficient at the outer in the coupled problem for the fluid a source coefficient considered flux at the borehole is compared to The inverse calculations were carried out for the first three stages of the Salt Block II experiment, at 0.2, 0.4, and 0.6 kW. An excellent representation of the temperature data was obtained, and the inferred properties are consistent with independent determinations. For example, for constant thermal properties, the procedure indicates a conductivity of 5.2 W/m/K, which is typical of measurements for WIPP salt [45]. The result 'i«K 3 E-63 ""I •"111 1 of central Interest here Is that for the brine diffusivity (the cerrr-eability divided by a capacitance and the brine viscosity). The si.r,ul:l1on/wcTe performed for a fixed value of permeability, k - O'^i m^ (1 nanodarcy), and they allowed for a temperature-dependent viscosity. The best fit to the fluid flux data was obtained for a reference (18'C) diffusivity value of c - 0.70 x lO"' m2/s. At 28*C, this ^'-spends to a diffusivity of c - 0.87 x lO"/ m^/s. For a Permeability Ir ;o1l ^2 (1 nanodarcy) and a viscosity of 1.6 x 10-3 Pa-s. this implies a lacltance of 7.2 10-12 pa-1. A previous estimate of the capacitance, bar'd on Independent estimates of the elastic properties of the brine and salt [29] was 5.7 x 10-12 Pa-1, and the corresponding diffusivity for k - 10-21 (1 nanodarcy) was c - 1.1 x 10-7 in2/$. Thus a fit of model calculations to data from the Salt Block II experiment yields a fluid diffusivity only about 25% lower than that comouted from independent estimates of the elastic propert es. This ^Sreement may be regarded as quite good, given the uncertainty in several orthnlterial properties. It might be noted, as well, that one would exoect the apparent diffusivity derived from a one-dimensional model simulation to be less than the apparent diffusivity for the m U S ins ional configuration. The effect of the finite length of the cv iideMs to allow axial losses of heat and pressure and to allow some relaxation of the pore pressure by axial expansion of the solid matrix. Thu the one-dimensional, radial model tends to overpred ct the fluid flux, which must be accommodated In the parameter estimation scheme by reducing the apparent transport coefficients. 3.4.4.3. ]nf^rpnces fr(?m Analyses fff ThermaHv-Drlven Bring Transport TestS Both laboratory and field experiments that measured brine flow rates stimulated by heating of salt from a borehole have been analyzed using a Da cy flow model! XltSough the driving force for the flows is different from those that operate under isothermal conditions, the mechanisms of •XaSe- (or capacitance) and flow resistance are identical. Thus, study of t Jse configurations his a direct bearing on the Isothermal prob ems ?ha ta of mo?e ini^ediate concern at the WIPP. In PJ-^^lcu ar he e experiments offer opportunities to perform Independent model validat on stSdies. and to Infer material properties by matching model calculation and data. Calculations with the Oarcy flow model for WIPPt)rine fit data from the Salt Block II experiment with very good agreement. The Salt BlocK u exoer ment is currently the only transient flow test that has been analyzed croletely n light of the Darcy flow model. Comparisons between the model calc at on and experimental data for the first three stages of the test (0200 6 kW) are excellent. An Inverse calculation yields a wholly emiirkil fluid diffusivity measurement, based principally on the decay rate of the borehole flux. This, when combined with an assumed permeability, provides a direct measure of the capacitance of the salt. ons E-64 The result 1$ only about 25X higher than the capacitance calculated based en the olastlclty wdel and Independent estimates of the properties. The heated borehole experiments at the WIPP also appear to be well represented by the linear, thermoelastlclty model, and the observed cumulative flux Is bracketed by calculations for permeabilities of 10-21 ni2 and 10-^y p£ (1 and 10 nanodarcles), values that are In good agreement with independently-made In situ measurements [31,42,43]. u'> 1 '•a a] E-65 4. PREDICTIONS OF BRINE INFLOW TO WIPP DISPOSAL ROOMS 4.1. Choice of Permpabnitv Va Iups and Othpr Model Parameters The range of 1 to 10 nanodarcles (10-21 to 10-20 m2).was chosen as the experimentally-supported expected permeability range for calculating exoected brine inflow to WIPP TRU waste disposal rooms and for dealized scoping calculations. The experimental support for that range is shown as a histogram in Figure 4. The data cluster very strongly ^n>his range In situ measurements of brini permeabilities in relatively ""^J^turbed WIPP host rock salt and in other rock types such as anhydrite all fall within the chosen range [31,42,431. Explicit relationships between the properties of salt and brine and coefficients appearing in brine flow model relationships are given in APPENDIX A of this report. Also given there are the material properties for WIPP salt that were used in the model. ''11 1,1 I •niili I 4.2. Sro pino Calculations fo r Idealized Geometries The calculations in this section serve to illustrate that the prediction of WIPP brine inflow cannot be divorced entirely from physical Inodels. For example, measurements made in boreholes of roughly the same size reveal nothing about the scaling of brine inflow to larger excavations. Furthermore, one does not know from tests done on a small time scale how to extrapolate brine inflow to much longer times. A model is necessary to translate the brine flow pattern surrounding a test borehole and its evolution in time to the brine flow pattern and time history of flow surrounding a disposal room. These calculations also serve to illustrate the magnitudes of brine inflow that one might expect from a Darcy flow mechanism and the sensitivity of inflow to model variations such as flow geometry and consideration of the transient flow component. 4.2.1. pmindarv and Initial Cond itions and Material Properties It is assumed that the mined room introduces surfaces at atmospheric pressure into a region initially at some uniform pressure yaj'fe. One might exp ct th t the initial pressure is bounded between hydrostatic (for the deph beneath the water table) and lithostatic (for the repository depth). The variation of hydrostatic or lithostatic pressure with depth is negl gib e within a few tens of meters of the repository More dei led S scussion of the initial condition, including the effect o^haUered mean stress field due to the presence of a cavity, is given in [25]. For simplicity, the Initial pressure in the following Sections (4.2.2 - 4.2.5) is taken to be hydrostatic: Po 6.0 X 106 Pa; (21) E-66 Range Implied by moisture release experiments in Rooms Al S B Anhydrite Waste Shaft Halite Facility Halite I ] Brine Inflow -23.0-22.5 -22.0 -21.5 -21. -20.5 -20.0 -19.5 -ig.o -18.5 -18 log (k) '.C „s ;i ills* i9 Z Figure 4. Brine permeabilities derived from in situ experiments, E-67 Si!!' '^l «i the choice of lUhostatk Initial pressure would simply Increase the cakuldted fluxes and volumes by a factor of about two. The cumulative flux is evali-ated at 200 years ; t - 6.31 X 109 $. (22) It has been estimated previously, based on Independent measurements of the mechanical properties of salt [e.g., 45], that the diffuslvity for WIPP salt is c - (1.1 X 10l*)k m2/s (23) where k Is given In units of m2. Permeability (Ic) values in the range of 10-21 to 10-20 nj2 (i to 10 nanodarcy) or lower have been derived for intact WIPP host rocic from independent in situ measurements of brine flow during fluid transport experiments [31,42,43]. It should be stressed that these estimates are subject to improvement from more detailed modeling and field measurements However, they are consistent with the current WIPP data base. k - 10-21 to 10-20 m2 The brine viscosity at 28'C is M - 1.6 X 10-3 pa.s . (24) (25) Equations (24) and (25) were used to calculate the diffuslvity, c, using equation (23). 4.2.2. Radial Flow to an Isolated Tunnel The geometry for a radial flow to an isolated tunnel is shown In Figure 5. This model accounts for flow from above and below the tunnel. It neglects, of course, the effects of the rectangular shape of the room, but those effects damp out for later time. The results for this model geometry have been discussed in a previous report [25]. The flux to the tunnel, q, is given by: |q(a,t)| ' — -J /id n • exp (-u^ct/a^) du (26) where a is the radius, and Jq and Yq are zero-order Bessel functions of the first and second kind, respectively. The total volume of brine is determined by multiplying the flux by the area of the tunnel walls (vertical side walls, floor, and ceiling for an equivalent rectangular room). A calculation for an equivalent waste disposal room follows. E-68 ..9 •M 2a ♦I Figure 5. Geometry for radial flow to an isolated tunnel E-69 The circumference of a reference waste disposal room (33 ft by 13 ft) Is 28 B (92 ft); thus, the effective radius of an equivalent circular tunnel is a - 4.5 B (27) and the appropriate area Is the sum of the side-wall, floor, and celling areas: A2 - 2548 u2 . (28) Equation (26) then gives the following total brine inflow volumes at the end of 200 years , V (for k - 10-21 m2) - 6.7 m3 (29) V (for k - 10-20 m2) - 40.6 m3 (30) 4.2.3. Steady State Flow to a line Sink At sufficiently long time, the pressure field does not relax to zero everywhere as implied by the diffusion model, but approaches a steady-state condition in which the far-field is hydrostatic and there Js/echarge at the water table. See Figure 6 for this geometry. This mode should y eld a smaller brine inflow value, because the higher transient flow at early times is not included. In this case, for a/d « 1, the flux at the room walls, Qwalli Is 9^ven by [25]: IPwalll • — •1 (31) /ia ln(a/2d) and the cumulative flux is obtained simply by multiplying hwalll by the wall area and total time of interest. The WIPP facility horizon is about 600 » below the water table, I.e., d - 600 B . (32) Equation (31), along with equations (27). (28), and (32), then gives the following total brine inflow values at the end of 200 years: V (for k • 10-21 m2) - 2.6 m3 , V (for k - 10-20 ni2) - 26.3 m3 . (33) (34) E-70 y^>C^< m' V Water Table d = 600m »I"J (9 111 Figure 6. Geometry for steady flow to a line sink. E-71 4.2.4. Horizontal Flow (1-D) to an Isolated Room This case represents a situation In which there is no vertical flow, perhaps because of impermeable, horizontal clay or anhydrite seams above and below the disposal room. (See Figure 7.) The flow is allowed to spread ojtward without bound, because adjacent rooms in a panel of rooms are not ^nsidered. This problem is exactly analogous to the cooling of a plane half- ipace, and the pressure profile takes the well-known form: P - Po erf -— ® 2yct (35) where Po is the initial pressure, x is the distance away from the wall, and c is the diffusivity. The flux at the wall, q (e.g., in units of mVs/m^), is determined from equation (35) using Darcy's law: •"tlii' |q(0.t)| kP, ij^ (36) where k is the permeability, and p is the brine viscosity. The cumulative flux, Q (e.g., in units of m3/m2), is obtained from (36) by integration: Q(t) - '^ t>/2 . (37) The cumulative volume of brine is determined by multiplying (37) by the area of the vertical side walls of the room. The vertical side-wall area for the model room is Ai - 728 m3 . (38) Thus, for 1-D flow from an unbounded domain, equation (37) predicts a cumulative volume, V (for k - 10-21 m3) - 0.73 m3 , (39) Y (for k - 10-20 m3) - 2.33 m3 . (40) 4.2.5. Horizontal Flow fl-Dl to a Room in ? Pa nel The next case to be considered is for one-dimensional flow to one room among an array of similar rooms separated by pillars of finite width. See Figure 8. In this case, the pressure disturbance can spread only to the centerline of the pillar, where it must be symmetric because of flow to the E-72 impermeable iii; uiit :;ji3 i a Z Figure 7. Geometry for lateral flow to an isolated room. E-73 SiSll'l •Its ' a' //. 1 k - \ /^r .- — ► m /--^ / / ^ '-—♦' Impermeable '-- \ / rt / Figure 8. Geometry for lateral flow in an array of rooms E-74 "^!!\k°°"''i I?^^ ^J'^^^"" ^^'"P^^' ^°°*^^ l^'^e ^^e cooling of a finite slab and the solution is again well known: * p.p^4 sin Ax , exp (-cA^t) , (41) ^K^'fi*- ^^^^J thickness of the pillar between rooms and Xnl - (2n + 1) The flux at the wall, q, is again obtained from Darcy's law by differentiation of (41) kp |q(0,t)| --° 4 exp (-cA^t) . n»0 The cumulative flux is obtained by integration of (42): Q{t) «^P«L pc 1 - exp (.cA;t) (V)' n«0 (42) (43) ofMieMdTiam""" '' '^''" °^^''"'^ ^^ "multiplying by the vertical area equa tion (38)rand^'°'" * ^'"'^^ '*°"'''" ^'^''''" '°°'"'' "'^"^ A] from above, L - 30.5 n , equation (43) gives: V (for k - 10-21 m2) - 0.37 in3 V (for k - 10-20 n,2) - 0.37 m3 .ii; Mi M" "J i9 E-75 as;- an. •tir I •It': ' •III Si These values are Identical, because the drainage process is essentially complete after 200 years even at the lower diffusivity. This is apparent from evaluation of the characteristic time, (L/2)2/c, which talces the value 2.1 X 105 s (67 years) for k - 10-21 in2 and 2.1 x 108 s (6.7 years) for k - 10-20 m2. Also note that the cumulative flux is significantly less than for the isolated room (unbounded flow region), because there is simply a smaller pressurized region upon which to draw. 4.2.6. Comparison of Results for Idealized Geometries Results from the highly idealized models considered here are collected in Table 4 for ease of comparison. Some observations can be made from these calculated results: Cumulative brine inflow to waste disposal rooms does not scale linearly with host rock permeability. An order-of- magnitude increase in permeability results in significantly less than an order-of-magnitude increase in accumulated brine. This non-linearity occurs, because the characteristic time for the transient component of brine inflow is a function of the permeability. The choice of a flow model has a significant influence on the calculated quantity of accumulated brine in waste disposal rooms. If vertical brine flow is strongly inhibited by bedding planes, brine inflow will be much smaller than for the isotropic flow case, and adjacent rooms in a panel will also cause significantly reduced flow to a disposal room. Bedding planes of unusually high penneability could increase brine inflow. The transient contribution to brine inflow is significant during the first 200 years for a waste disposal room. The expected brine inflow volume to a waste disposal room is to be no more than a few tens of m3 in 200 years, based on this model 4.3. Calculations of Expected Brine Ac cumulation in WIPP Disposal Rooms The WIPP brine flow model was used to calculate, by numerical methods, exoected brine accumulation values for the WIPP reference disposal room ge metry 4 « (13 ft) high by 10 m (33 ft) wide by 91 m (300 t) ong). These calculations yield more accurate estimates of bnne inflow than were obtained from the above scoping calculations for idealized geometries. Transient, two-dimensional numerical analyses were performed for three different disposal room configurations: (1) a room with reference E-76 Model Lateral semi-inf. Lateral finite Radial Line sink Equation Cumulative Volume (m3) 10-21 m2 k = 10-20 m< (37) 0.7 2,3 (43) 0.4 0.4 (26) 6.7 40.6 (31) 2.6 26.3 Table 4. Summary of results for cumulative volume at 200 years. i & •w M a I I Mil Id! .iri{ ;:;3 lis E-77 c.,;:.; 531(1 dimensions placed between adjacent rooms 1n a reference panel conflguratio (30 5 m (100 ft) wide salt pillars between rooms); (2) a reference roo::i sufficiently distant from other rooms so that there are no brine flow interactions with any other excavations; (3) a room that is la[9er than reference in order to simulate, with voidspace, a high-permeability disturbed zone surrounding a reference room Values for model parameters were chosen to represent expected or reasonable ranges. The permeability range of 1 to 10 "^"jdarcies was chosen as described above, as a the expected range for the calculation of br e inf w iuation (18) was used to calculate the diffusivity Two va es for the initial far field (undisturbed) pore Pressure were cho en: hydrostatic pressure (6 HPa) and lithostatic pressure (15 HPa). These pressure valSes are reasonable bounds for the expected undisturbed pore pressure. Brine accumulations were obtained by integrating inflow rates from tl mnrnpnt of excavat on (t - 0). Actual accumulations in WIPP disposal room: Trrpte Jailer Aeca^ water from inflowing brine will e removed by evaporation into ventilation air during early times when the inflow rate is highest. Brine inflow into waste-containing, backfilled WIPP disposal rooms 1 expected to cease within 100 years due to consolidation of room conten s creeo closure [461 and the resulting Increase in pore pressure within the rnnm5 The oresent calculations were carried out to 200 years for rompi;tene!s'and ease of comparison with the scoping calculations present in the previous section of this report. The numerical model constructed for these studies was based on sever simplifications: The variation of hydrostatic or lithostatic pressure with depth was assumed to be negligible within a few tens of meters of the repository. The effect of closure on room geometry was neglected. Closure increases the brine flow path and could decrease brine inflow. Pressure build-up during creep closure due to the compression of room contents is not accounted for. Increasing rorpresrure would decrease the driving force for brine influx. Therefore, neglecting this interaction is conservative. Synmetry of brine flow was invoked to simplify the numerical model. Because of the large geometrical dimensions associated with the model, the specification of impermeable boundaries for a 1 exuTor element boundaries vertically above the repository is a good approximation of the real situation. E-78 Details of the physics, algorithm, model geometry, r.aterlal ?I^S^5oV";.*"^ boundary and Initial conditions are presented elsewhere [47,48]. The rnesh Is a two-dimensional Cartesian finite element mesh that roL?!??nn^'f tJ! ^^' K^T"^ ^^^^ ^^"^^^ ^^^"^"^ ^raphics package. Upo completion of the mesh. It was translated to the equivalent finite difference network. The diffusion equation for pore pressure equation (1), was solved numerically using Q/TRAN [50]. 4.3.1. WIPP Disposal Room In ^ pa noi Brine inflow to a typical waste disposal room In a panel was calculated for hydrostatic and llthostatic initial (undisturbed host rock) pore pressures and permeability values of 1 and 10 nanodarcles. The expected range of brine accumulation in a TRU disposal room Is 4 permeability, to 43 m3 In 00 years for llthostatic Initial pore pressure and 10 nanodarcy permeability. Calculated cumulative volumes are plotted in Figures 9 through 12 for times to 200 years. F'ULiea 4.3.2. Sensitivity to Initial Pore Prp<;«:iirp Because of the linearity of the model, the brine flux and cumulative brine nflow are proportional to the Initial pore pressure. Ill Al^ T^A ^ ^^t analytical results discussed previously (equations 6,36.42), and corroborated by the numerical calculations At a permeability of 1 nanodarcy. the cumulative brine volume in 100 years increases from 4 m3 to 9 m3 when the initial pore pressure Is Increased from hydrostatic to llthostatic (from 6 to 15 Mp!)^ A an d cy ?he cumula ive volume Increases from 17 m3 to 43 m3 fir the same ch ng7 n the nitia pore pressure. Figures 13 and 14 illustrate the sensitivity to the initial pore pressure. -^ :s3 t9 ^•3.3. Sensitivity of Brine Inflo w Host Rnrk PermeabilitY tho K^"*^"^?^!?^ ^t^ ^°r [O'^k permeability from 1 to 10 nanodarcy increases the brine Inflow by a factor that lies between 4 and 5. There Is ^ '"^^''" bpr .l'^^?l!tI°^'^'^''^^y."^'"^^ ^'•^"^ ^^^^^^ ^nd permeability, because the rate at which the transient brine Inflow decays depends upon the permeability. The change in brine Inflow rate Is significant at these permeability values during the first 100 years. These results are illustrated in Figures 15 and 16. results are 'Z 1? ^•3.*- Effect of a High-Per meabil i t y DisturbPd 7nne Surroundinn . Waste Disposal Room ^*-^ wastJS^nnn]°^nn™ScVil'i^^"^''"''''*'^^^^^ disturbed zone surrounding a waste disposal room Is unlikely to cause a significant increase in brine E-79 fiat SiSI Willi lU..' *";! 'lii' m 7.0 6.0 5.0 T 1 r E 4.0 o > > 3.0 3 E 5 2.0 1.0 - 0.0 0.0 T 1 r -1 1 r >-21 -I 1 I + - Permeability K = 1 + - Hydrostatic Po = 6 .OX 1 Pa J L J L 0.4 0.8 1.2 Jl L Time years (•10 ) ■ I L 1.6 2.1 Figure 9. Calculated brine accumulation in a typical waste disposal room in a panel; Po - hydrostatic pressure; K » 1 nanodarcy. E-80 o K) 1.8 1.5 1.2 T r 1 1 r T r T 1 r -> r I 0.9 3 O > ^ 0.6 E Panel Room - Permeobillty K = 10"^' m^ - Llthosfofic Po = 15.0X10* Po 0.0 ' ' ' ' 1— 0.0 0.4 J ■ ■ J 1 L 0.8 J L 1.2 1.6 2.0 Time years ( '10^ ) 1^; I2i Figure 10. Calculated brine accumulation in a typical waste disposal room in a panel; Po - lithostatic pressure: K » 1 nanodarcy. *^ E-81 "laii:: I »r6",:( i 'HI' i 3.0 2.5 O T — I — I — T T r ■1 — I — I — T .-20 + -Permeability K= 10 - + - Hyd-ostotic Po = 6.0X 1 Pa 2.0 E ^^ D O > Z 1.0 o E ^05 0.0 0.0 _1_ 0.4 J I L 0.8 1 1 r T r J L 1.2 Time yeors(*10 ) 1__J L 1.6 Figure 11. Calculated brine accumulation in a typical waste disposal room in a panel; Po - hydrostatic pressure; K . 10 nanodarcy. E-82 7.0 6.0 o 7 5.0 4.0 E o 3.0 > > I 2.0 E 3 o 1.0 T 1 r T r T r r 0.0 Panel Room -Permeobnity K= 10'"m^ - Lithostafic Po = 15.0X10* Pa J L 0.0 0.4 J t I 0.8 1.2 1.6 2.0 I :S :::3 I* Time years ( ♦lO ) Figure 12. Calculated brine accumulation in a typical waste disposal room in a panel; Po - lithostatic pressure; K ■ 10 nanodarcy. E-83 y •>-v t,,;,;' 1^ O 1^ u T r T 1 r 1 Pcrw! Room Anolysis - Lllhosiotlc vs Hydros! otic Permeobinty K= 10'^* m* Lilhosiofic Po = 15.0X10* Po Hy(i-osiatlcPo = 6.0X10* Pa |0.9h D O > o Z 0.6 o D E ^0.3 0.0 J I L J I L. 0.0 0.4 1 J I L 0.8 Time 1.2 T 1 f J L T r yeors(MO^) I i L 1.6 2.0 Figure 13. Sensitivity of calculated brine accumulation to initial pore pressure; typical room in a panel; K ■ 1 nanodarcy. E-84 7X) O • T r T 1 r 1 1 r E 3 O > > 1 ' T T — ' — ' — r Panel Room Arxalysls Pemvjobinty K = 10"" m* Uttwstoflc Po = 15.0x10! Pa Hydrostatic Po = 6.0X10 Pa Time years (• 10 ) s tit :;3 ^0 Figure 14. Sensitivity of calculated brine accumulation to initial pore pressure; typical room in a panel; K - 10 nanodarcy. E-85 3.0 ii;: 2^ O T 1 r Pond Room Anolysis - Sensitivity to Permeobaty L Hyd-ostotlc Po = 6.0X10* Po Permeobllity K = 10" m 2.0 O > ^ 1.0 ^0.5 0.0 0.0 -21 2 PefTTkeobility K = 10 m XHj — L. J L J L J L 0.4 0.8 1.2 1.6 Time years (•10 ) Fiqure 15. Sensitivity of calculated bnne accumulation to host rock permeability; typical room in a panel; Po - hydrostatic pressure. E-86 7^ 5.0 o 7 5.0 4.0 Q) E o 3.0 o S 2.0 E o 1.0 T r T T r T r Panel Room Andysis Lithosfatic Po = 15.0X 10* Pa Permeobinty K= 10"" m' Permeability K= 10"^' m' 0.0 11 J L. 0.0 J L 0.4 J L 0.8 J L. 1^ 1.6 Time years (♦10^) 2.0 § ::3 1119 Figure 16. Sensitivity of calculated brine accumulation to host rock permeability; typical room in a panel; Po - lithostatic pressure. E-87 inflow. The worst-case disturbed zone surrounding a room has infinite permeibility. Such a disturbed zone can be simulated by Mving the atmospheric-pressure boundary into the host rock and calculating room inflow at that boundary. Host rock salt within that boundary is assumed to be hydraulically isolated from the far field; thus the brine that it contains experiences no driving force (pore pressure gradient) for flow. A disturbed zone 10 m thick above and below a room and 5 n thick on either side was simulated by increasing the height of the room by 20 m and the width of the room by 10 m. Calculated results are plotted in Figures 17 and 18 for penneabilities of 1 and 10 nanodarcy. The initial pore pressure was taken to be lithostatic pressure (15 HPa). In this simulation, the disturbed zone increased the calculated 100-year cumulative brine inflow volume from 43 to 52 m3 for the maximum expected permeability of 10 nanodarcies. For 1 nanodarcy, the increase was from 9 n^ to 17 m^. 4.3.5. Effect of Adjarent Rooms in a Panel The effect of adjacent rooms in a panel on brine inflow is to decrease the 100-year cumulative brine volume by approximately 25X when the host rock permeability is 10 nanodarcy (10-^0 m2). This comparison is shown in Figure 19 for hydrostatic pressure as the initial pore pressure and in Figure 20 for lithostatic pressure. The comparison is between the calculated brine inflow to a room far from other rooms and the previously- presented calculated inflow to a room in a panel of rooms. E-88 5.0 T r : 4.0 O m 3.0 - HydrostaficPo = 6.0X10* Pa Permeability K= 10"^° m^ + - Miifiple Room Andysis O - Single Room Andysls 0) E o > 2.0 > E 0.0 0.0 ..•♦*"' .♦•■ 0.4 0.8 U J L 1.6 2.0 Time years (♦ 10 ) Mi -'IS ::3 >; I Figure 19. Effect of adjacent rooms in a panel on calculated bnne accumulation; Po - hydrostatic pressure; K - 10 nanodarcy. E-89 iiiiii ' u 1.0 O « 0.8 I 0.6 jg 0.4 3 0.2 T — I — I — I — I — I — I — I — I I I T PoTMi Room Modd vs tsolotdd Room Modal I I T r 0.0 0.0 Uthostoflc Po = 15.0X10* Pa Permeobinty K= 10"^° m Miitlple Room Andysis bolatad Room Andysis .♦•*'* ..-•••' .<♦' ..«•' ,♦• ,r I I J L J I 1 1 0.4 0.8 1.2 1.6 Time years (♦ 10 ) Figure 20. Effect of adjacent rooms In a panel on calculated brine accumulation; Po - lithostatic pressure; K - 10 nanodarcy. 2.0 E-90 5. ASSESSMENT OF BRINE INFLOW EFFECTS ON WIPP DISPOSAL ROOMS An assessment of brine Inflow effects on disposal rooms is necessary to address the potential consequences of this brine for TRU waste Isolation. It Is desirable to assure that room contents remain in a solid (non-flowing) rather than a fluid state. The final state of rooai contents will depend on the relative rates of brine Inflow and consolidation of room contents by creep closure. Consolidation is expected to be virtually complete within 100 years [46]. It was determined that water-absorbing tailored backfill materials can readily absorb the maximum credible expected 100-year brine accumulations in WIPP disposal rooms without becoming brine-saturated. This assessment was done by coupling expected maximum credible brine accumulations in disposal rooms, the expected maximum reconsolidation time of 100 years [46], and estimated absorption capacities for room backfill materials. The data and calculations that were used are described below. 5.1. Expected Brine Accumulations in WIPP Disposal Rooms Expected accumulations of brine in typical WIPP waste disposal rooms were calculated by numerical methods using a mathematical description for the brine Inflow model. These numerical calculations were given in Section 4J of this report. WIPP disposal rooms filled with waste and backfilled are expected to become virtually completely compacted due to host rock salt creep in about 100 years [46], preventing further accumulations of brine. Therefore, brine accumulations during the first 100 years were used here. For a comparative reference, a typical room has an Initial excavated volume of approximately 3600 cubic meters (950,000 gallons). A sunwary of 100- year brine accumulations from the numerical calculations Is as follows: Host Rock Permeability, Nanodarcies Pre-Excavation Pore Pressure Cumulative Brine Volume In Typical Waste Disposal Room after 100 Years, Cubic Meters, (Gallons), (X of Initial Room Volume) i:3 It! 9 1 1 10 10 Hydrostatic Lithostatic Hydrostatic Lithostatic 4 9 17 43 m3 m3 m3 m3 ( 1000 ( 2400 ( 4500 (11000 gal) gal) gal) gal) (0.11%) (0.25%) (0.47%) (1.19%) Other scoping calculations (in Section 4.2 of this report) for idealized room geometries (long cylinders) provided confirmation of the above results, yielding volumes in the range of approximately 1 to 40 m3. The worst credible case 43 in3 of brine is 1.2% of the Initial room volume, about the same as the quantity of brine in the salt that was removed by raining the room. To gain some visual perspective on the relative magnitude, one can visualize a layer of brine 4.6 cm (1.8 Inches) E-91 IL,. deep on the floor of a 4 m (13 foot) high room as the equivalent of 43 m^ of brine in a typical empty WIPP waste disposal room. It will be shown in the next section that backfill materials such as crushed salt and bentonite clay can readily absorb such a quantity of brine without becoming saturated or degraded. 5.2. Absorotion of Accumulated Brine bv Backfills As-mined (granular) WIPP salt backfill alone can absorb 40 m3 of accumulated brine in a disposal room (93% of the predicted worst case 43 m^), according to conservative estimates of room backfill quantity and water absorption capacity. The absorption capacity is the difference between the measured water content (0.5 wt% or less) of mined WIPP salt backfill material and the water content (2.5 wt%) of mechanically strong blocks pressed from WIPP crushed salt. Details of brine absorption capacity calculations for crushed salt are given in the next section of this report. A tailored backfill material mixture of 30 wt% bentonite in crushed WIPP salt can absorb 120 m3 of accumulated brine. That is about 3 times the predicted worst case 43 m3 in 100 years. This result was also based on conservative estimates of room backfill quantity and water absorption caoacity for bentonite. Bentonite in this WIPP room backfill mixture has the capacity to absorb 90 m3 of water (chemically bound) without becoming water-saturated [51]. This absorption capacity takes into account water that would be pre-absorbed from WIPP air at approximately 70% relative humidity [52], an actual humidity value that is currently being measured by Sandia in WIPP boreholes (ongoing Room D brine inflow and humidity experiments). Details of brine absorption capacity calculations for bentonite/crushed salt mixtures are given In the next section of this report. Tailored backfill mixtures with bentonite as a water absorber have always been considered in WIPP backfill Investigations. Bentonite mixed with 70 wtXWIPP crushed salt is currently being tested in WIPP simulated CH TRU waste technology experiments [53]. The long-term stability of bentonite in contact with WIPP brines is supported by reported Sandia studies [54). 5.3. Capacities of Room Back fill Materials for the Absorption of Brine Absorption capacity values were calculated in the following way. A minimum backfill volume In each disposal room was calculated for a maximum reasonable packing density of waste drums. An empty space two feet thick at the top of each room allows for backfill emplacement with commercially available solids handling and conveying equipment. The water absorption caoacities of crushed WIPP salt and a mixture of 30 wt% bentonite in crushed WIPP salt, both as emplaced backfill materials, were calculated from published data. Then the quantity of accumulated brine that the E-92 backfill In a room can absorb was calculated by combining backfill quantities and absorption capacities with the measured water content of Ml?? brine. WASTE DISPOSAL ROOM VOLUME AVAILABLE FOR BACKFILL Biiii: 33 ft wide by 13 ft high by 300 ft long waste disposal rooms 2 ft diameter by 3 ft tall drums 3 layers of drums (drums stacked 3-hlgh) 150 rows of drums, maximum, In each layer 15 drums, maximum, in each row 2 ft empty gap between emplaced backfill and room back (roof) Calculations : volume of each drum - jr(l)2(3) • 9.4248 ft^ maximum number of drums per room - 15(150)3 - 6750 drums per room maximum volume occupied by drums - 6750(9.4248) - 63,617 ft3 volume of empty gap above backfill - 2(33)300 • 19,800 ft3 volume of disposal room after excavation - 13(33)300 - 128,700 ft^ minimum volume available for backfill - 128,700 - 63,617 - 19 800 - 45,283 ft3 per cent of initial room volume available for backfill ■ 45,283 ♦ 128,700 x 100 - 35% WATER ABSORPTION CAPACITY OF WIPP CRUSHED SALT filiii: as-emplaced water content [55,56] - 0.5 wt% maximum water content in strong crushed salt blocks [57] - 2.5 wt% net allowed water content gain - 2.5 - 0.5 - 2.0 wt% bulk density of crushed salt backfill material [55] - 1300 kg/m3 ^? :.i "I -si IS E-93 Calculations : minimum water absorption capacity - (0.02)1300 ■ 26 kg water/in3 crushed salt volume of disposal room after excavation - 128,700(0.3048)3 - 3644 m3 volume available for backfill - 3644(0.35) • 1276 m3 - 35% of room volume quantity of water that can be absorbed in the crushed salt backfill in a room - 26(1276) - 33,164 kg water absorbed/room WATER ABSORPTION CAPACITY OF A MIXTURE OF 30 WT% BENTONITE WITH CRUSHED WIPP SALT asis: "lyt 1 1 water content of bentonite equilibrated with water vapor In disposal room air [52] - 0.15 g/g bentonite total water capacity of emplaced bentonite [52] - 0.3 g/g bentonite available water gain In bentonite backfill [52] - 0.15 g/g bentonite bulk density of 30 wt% bentonite in WIPP crushed salt [55] - 1300 kg/m3 Calculations : water absorption capacity of bentonite in mixture - 0.15(0.3)1300 - 58.5 kg water/m3 backfill mixture water absorption capacity of crushed WIPP salt In mixture - 0.02(0.7)1300 - 18.2 kg water/m3 backfill mixture total water absorption capacity of backfill mixture - 58.5 + 18.2 - 76.7 kg water/m3 backfill mixture volume of disposal room after excavation - 128,700(0.3048)3 - 3644 m3 volume available for backfill - 3644(0.35) - 1276 m3 - 35% of room volume quantity of water that can be absorbed in the crushed salt/bentonite backfill mixture In a room - 76.7(1276) - 97,869 kg water absorbed/room E-94 ABSORPTION OF WIPP BRINE BY DISPOSAL ROOM BACKFILLS Basis ; maximum expected 100-year brine accumulation - 43 m3 brine/room density of WIPP brines [58] - 1.2 g/cm3 - 1200 kg/m3 water In WIPP brine "weeps" [58] - 0.6877 leg water/kg brine quantity of water that can be absorbed by a room backfill of 100% crushed WIPP salt (see above) - 33,164 kg water/room quantity of water that can be absorbed by a room backfill mixture of 70 wtX crushed salt/30 wt% bentonite (see above) - 97,869 kg water/room Calculation for 100% crushed WIPP salt : quantity of brine that can be absorbed by crushed WIPP salt backfill - 33,164 ♦ ((0.6877)(1200)) - 40 m3 brine/room per cent of 100-year brine accumulation that can be absorbed by WIPP crushed salt room backfill • 40 ♦ 43 - 33% Calculation for 70 wt% WIPP crushed salt/30 wt% bentonite mixtur^ r quantity of brine that can be absorbed by mixture - 97,869 ♦ ((0.6877)(1200)) - 119 m3 brine/room per cent of 100-year brine accumulation that can be absorbed by crushed salt/bentonite room backfill mixture - 119 ■»■ 43 ■ 277% E-95 vsc. 6. SUMMARY AND CONCLUSIONS Water-absorbing tailored backfill materials can readily absorb the maximum expected brine accumulations in WIPP disposal rooms while maintaining mechanical strength and without becoming brine-saturated. Crushed WIPP salt alone can absorb almost all of the maximum expected brine accumulation. Salt creep is expected to virtually completely reconsolidate baclcfilled waste disposal rooms within 100 years, increasing the pore pressure in the room and stopping brine accumulation at that time. The expected 100-year brine accumulations were calculated with a predictive Darcy flow model for the movement of brine to WIPP excavations. The model, data base, expected brine volumes, brine absorption capacity of backfills, and needs for further work are summarized below. 6.1. Brine Inflow Model We have a predictive model for the movement of brine to WIPP excavations from WIPP rock salt. This model is based on well-known physical processes of groundwater flow in granular deposits. All values for model parameters are consistent with independent measurements of brine and host rock salt properties, and brine movements calculated from the model are consistent with the body of existing data for brine accumulations In WIPP underground test boreholes. The details of the model and its applicability to WIPP rooms and test boreholes rest upon a number of assumptions that are being subjected to further testing. Experiments are underway in the WIPP specifically for that purpose [59]. According to the model, brine flows In Intergranular spaces within the polycrystalline host rock sail under the driving force of preexisting hydrostatic (groundwater head of approximately 900 psi, or about 6 MPa ) or lithostatic (overburden pressure of approximately 2200 psi, or about 15 HPa) pore pressure toward the atmospheric pressure at excavation walls. The capability of the host rock salt to allow flow under this driving force, conr-only expressed as a "permeability". Is very small, In the range of 1 to 10 nanodarcies. These permeability values are In good agreement with independent WIPP In situ fluid flow measurements. The Darcy flow process in geologic materials Is well understood, and the describing mathematical formalism Is accepted by the scientific community. 6.2. Brjne Inflow Data Base The range of permeability values for the model, 1 to 10 nanodarcy, was derived from WIPP In situ tests and brine sampling data, and data from moisture release experiments. Permeability values in this range or lower have been derived for Intact WIPP host rock from several Independent in situ measurements of brine flow In the host rock salt and In Interbeds such as anhydrite (e.g., Marker Bed 139). These in situ measurements constitute the most reliable source for the host rock permeability. The measurem.ents were made at the disposal horizon and at Intervals above In the waste- E-96 handling shaft. Permeabilities in the disturbed zone near drift walls were greater than 10 nanodarcy. Darcy flow permeability values calculated from IT Corporation's WIPP brine sampling data were described reasonably well by a typical lognormal distribution with a logarithmic mean of 3.5 nanodarcy. A lognormal distribution of permeability values Is a common observation for other rock types. Permeability values similarly calculated from Sandia moisture release data (Rooms Al and B) are In the range of 2 to 9 nanodarcy. It 1$ our judgement that the uncertainty In permeability is In the order-of-magnltude range. The details of the model and Its applicability to WIPP rooms and test boreholes also rest upon a number of assumptions. For the most part, these assumptions are likely to ^leld conservatively larqe values for long lenn brine Inflow. Critical assumptions concerning flow mechanisms are being tested with ongoing and planned WIPP experiments. Potential Inaccuracies stemming from Idealized geometries are being investigated with more detailed numerical calculations. 6.3. Calculated Brine Accumulations The maximum expected brine accumulation In a disposal room was calculated to be 43 m^. Expected accumulations of brine In typical WIPP waste disposal rooms during 100 years after waste emplacement were calculated by numerical methods using a mathematical description for the brine Inflow model. WIPP disposal rooms, filled with waste and backfilled, are expected to be virtually completely reconsolldated due to host rock creep In about 100 years, preventing further accumulation of brine. Expected cumulative brine volumes were In the range of 4 m3 to 43 m3. Other, less complex calculations for Idealized room geometries (long' cylinders) provided confirmation of these values, yielding volumes in the range of about 1 to 40 m3. The maximum expected accumulation, 43 m3, is 1.2% of the initial room volume, about the same as the quantity of brine in the salt that was removed by mining the room. 6.*. Absorption of Accumulated Brine by Room Backfills Mined WIPP salt backfill alone can absorb 40 m3 of accumulated brine in a disposal room (93% of the expected worst case of 43 m3), according to conservative estimates of room backfill quantity and water absorption capacity. The absorption capacity is the difference between the measured water content (0.5 wt% or less) of mined WIPP salt backfill material and the water content (2.5 wtX) of physically strong blocks pressed from WIPP crushed salt. A tailored backfill material mixture of 30 wt% bentonite in crushed WIPP salt can absorb 120 m3 of accumulated brine, about 3 times the worst credible case of 43 m3. The bentonite in this WIPP room backfill mixture has the capacity to absorb 90 m3 of water without becoming water-saturated This absorption capacity takes into account water that would be pre- :» Id E-97 SKI. ma it'.. Mil *i| absorbed from WIPP air at approximately 70% relative humidity, an actual humidity value that is currently being measured by Sandia in WIPP boreholes (ongoing Room D brine inflow and humidity experiments). Tailored backfill mixtures with bentonite as a water absorber have always been considered in WIPP backfill investigations. Bentonite mixed with 70 wt% WIPP crushed salt is currently being tested in WIPP simulated CH TRU waste technology experiments. The long-term stability of bentonite In contact with WIPP brines is supported by reported Sandia studies. , 6.5. Needs for Further Work Remaining uncertainties in the host rock permeability, in other brine inflow model parameters, and in mechanistic details of the model should be addressed. Experimental work and model development are needed. The following in situ measurements are recommended to reduce uncertainties and test aspects of the existing model: host rock permeabilities to brine throughout the WIPP underground and in all representative strata host rock pore pressures beyond and within the disturbed zone brine inflow rates to excavations of significantly different scale. Including large room-shaped excavations brine inflow rates to identifiably different strata responses of host rock flow properties and pore pressures to changes in stress and strain Scale-up predictions and certain mechanistic assumptions in the model concerning pore pressure and flow paths will be tested with ongoing and planned WIPP in situ tests in small (4-inch) and large (36-inch) diameter boreholes [59]. Laboratory measurements of shear strain and permeability may aid the development of relationships between host rock creep and flow properties. Brine inflow model development is also recommended. Permeability variations that depend on stratum, general location, host rock stress, and host rock creep (disturbed zone development) should be considered in the model The host rock salt is heterogeneous, and, to be complete, the model should be developed further to reflect that heterogeneity. Experimental testing of model assumptions can be guided by sensitivity studies. E-98 APPENDIX A: MATERIAL PROPERTIES The Gxplklt relationships between the properties of salt and brine and the coefficients in the model equation (1) are as follows: The fluid diffusivity, c, Is given by k 2G(1 - v) c 'B^l + y/{\ - 2v) 9(1 - -uH-u ■ -) 1 Ml - K,/KJ - . 1 + *^ L_i. B Kfd - KA ) 3u + B(l - 2v.)(l - K/K^) " 3 - 8(1 - Z.)(\ . K/K^ where 6 is the elastic shear modulus, v is Poisson's ratio under 'drained' (p ■ 0) conditions, to ^s ^^e reference porosity, K is the drained bulk modulus, Kf is the fluid bulk modulus, and Kj is the bulk modulus of the solid, mineral grains. The source coefficient, b', is given by b' 4GB(1 4 u^) 9(1 - -u) 8(1 . ^)(1 + . ) where oj and of are the thermal expansion coefficients for the solid and fluid constituents, respectively. Typical values of these properties for WIPP salt, used in the following calculations, are collected in the following table: \9 19 E-99 Property Symbol Value Units Thermal : CO ■it mtSi ; •ill 111 )i», , Ml ill h Thermal conductivity K Thermal Diffusivity K Elastic: Drained bulk modulus K Shear modulus G Drained Poisson ratio V Fluid bulk modulus Kf Solid bulk modulus Ks Fluid expansivity {28'C) af Solid expansivity as Hydraulic: 1 Permeability k Porosity «o Fluid viscosity {28'C) f* Derived: Fluid diffusivity c Source coefficient Pressure coefficient Undrained Poisson ratio Diffusivity ratio b' B 5.0 2.5 X 10-6 20.7 12.4 0.25 2.0 23.5 4.6 X 10-4 1.2 X 10-4 10-21 to 10-20 0.001 1.6 X 10-3 1.1 X 10-7 to 1.1 X 10-6 1.1 X 106 0.926 0.273 0.042 to 0.419 W m-1 K-1 m2 s-1 GPa_ GPa GPa_ GPa K-1 K-1 m2 - Pa s m2 s-1 Pa K-1 E-100 REFERENCES 11 J. J. Hohlfelder, Salt Block II: Description and Rg^uU.^ . SAN079 Sandia National Laboratories, June 1980. 2226, 25 E. J. Nowak and D, <,» .. . ^' McTigue, Interim Rpsults of Rrinp T rancporf 26 E. J. Nowak, Preliminary Results of Brine Migration Studies in the Waste Isolation Pilot P lant fWIPP) . SAN086-0720, Sandia National Laboratories, Hay 1S86. 28 E. J. Nowak, "Brine Movement in Waste Isolation Pilot Plant (WIPP) Tests," SAN086-I094A, Sandia National Laboratories, presented in th( symposium on: Origin i Evaluation of Brines in the Subsurface terican Chemical Society National Meeting, Anaheim, CA, September 7-12, 1986. ^ ^^ ?: ^. McTig..e and E. J. Nowak, "Brine Transport in the Bedded Salt or the Waste ISO ation Pilot Plant (WIPP): Field Measurements and a Oarcy Flow Model, in: Scientific Basis for Nurlear Waste Han^QPmpnt y , Materials Research Society Symposia Proceedings, Vol. 84, Materials Research Society, Pittsburgh, 1988, accepted for publ ication. 30 0. E. Deal and J. B. Case, IT Corporation, Brine Samolina and Evaluation Program. Pha^P T Rppnrt, DOE-WIPP-87-008. prepared by the Engineering and Technology Department of the Management and Operating Contractor, Waste Isolation Pilot Plant Project for the U S Department of Energy, June 1987. 31 E. W. Peterson, P. L. Lagus and K. Lie, WIPP Horizon Free Field Fluid Transport Characteristics. SAN087-7I64, prepared by S-CUBED A Division of Maxwell Laboratories, Inc., for Sandia National Laboratories, December 1987. 32 J. C. Stormont, E. W. Peterson and P. L. Lagus, Suc-narv of and Observations about WIPP Facility Horizon Flow Measurements thrnnc:h 1286. SAND87-0176, Sandia National Laboratories, May 1987 36 H, A. Biot, "General Theory of Three Dimensional Consolidation," Journal of Aoolied Phvsics . 12, (1941), 155-164. 37 J. R. Rice and M. P. Cle?ry, "Some Basic Stress-Diffusion Solutions for Fluid-Saturated Elastic Porous Media with Compressible Constituents," Reviews of Geophysics and Soace Phvsics . 14, (1976), 227-241. 38 0. F. McTigue, A Linear Theory for Porous Thermoelastic Materials . SAN085-1149, Sandia National Laboratories, September 1985. E-101 REFERENCES nitltj |i 39 40 42 43 44 45 46 47 43 49 50 51 52 53 F HcTigue, "Thermoelastic Response of Fluid-Saturated Porous Rick," J. Geophysical Research , ii, 1986, pp. 9533-9542. J. Crank, The Mathematics of Diffusion . Clarendon Press, Oxford, 1979, p. 87. G J Saulnier and J. D. Avis, Tnteroretati on of Hydraulic Tests Conducted in the Vaste-Handl ino Shaft at the Waste IsolUion Pilot plant (WIPPl Site . SAND88-7001, prepared for Sandia National Laboratories, report in preparation. J C Stormont, Division 6332, Sandia National Laboratories, personal convnunication about in situ brine flow and permeability measurements, reports in preparation. F HcTigue, "Flow to a Heated Borehole in Fluid-Saturated Thenioelastic Rock (abstract)," EOS Transactions of the American Geophysical Union, 68, (1987) p. 1295. R Krieq, Reference Stratigraphy an d Rock Properties for the Waste T^nUtion Pilot Plant (WIPP^ Project . SAN083-1908, Sandia National Laboratories, 1984. H S Morgan, "Estimate of Time Needed for TRU Storage Rooms to Close,' Memorandum to 0. E. Munson, Sandia National Laboratories, Division 1521, June 2, 1987. R Beraun, 'Brine Flow Numerical Modeling for the WIPP Disposal Rooms," Memorandum to Distribution, Division 6332, Sandia National Laboratories, January 22, 1988. R Beraun, 'Fluid Flow-Heat Transfer Equivalence,' Memorandum to Distribution, Division 6332, Sandia National Laboratories, February 29, 1988. PDA Engineering, Inc., p^tpan-C, u.er-s Guide. Volumes I and II, Santa Ana, CA, 1980. F. A. Rockenbach, Q/TRAN . PDA Engineering. Inc., Santa Ana, California, May 1986. R Pusch "Highly Compacted Bentonite - A Self -Healing Substance firN c UarWaste'lsolation/ in: Vientifir pa.i. fpr Nuc ear^ Management . Vol. 3, J. G. Moore, ed.. Plenum Press, New York, 1931. B. M. Butcher. "Bentonite Water Sorption," Memorandum of Record, Sandia National Laboratories, January 6, 1983. M. A. Molecke, T^.t Plan: WIPP ^imnUtgd CH ^nd PH TPII Va^te TestSL TArhnnlQov Exn^rimpnts (TRU TE) . Sandia National Laboratories, Apn 1 1986. E-102 REFERENCES 54 55 56 57 53 59 J. Krumhansi Backfills in Observations RpgarHi nq the Stability of Bentonite-Based ^ High-level Waste R epository in Rock Salt . SAND83-1293, Sandia National Laboratories, June 1984. Material Specifica tions and Reouirements for T. W. Pfeifle, Backfill !^!■■!!P^■l!'^"^^^^^ ^^^^ ^"^ ^^^ ^^^^^ T e chnology FxperifTPnt.. , SAN085-7209, Sandia National Laboratories, July 1987. 0. J. Holcomb and M. Shields, Hydrostatic Creep Consolidation of Crushed $alt with Added Wat^r, SAND87-1990, Sandia National Laboratories, October 1987. J. C. Storoont and C. L. Howard, Development. Implementation and I V f!" ^' ^"^ ^^^^'''^ ^ °^^^^ Sm;^ n.Scale Seal Performance iJlLi, SAN087-2203, Sandia National Laboratories, December 1987. C. L. Stein and J. L. Krumhansi, Chemistry of Brines in Salt from the Vast? holation Pilot Plant fWIPPK Sou theastern New Hptj^-o- a Preliminarr Investigation . SAND85-0897, Sandia National Laboratories, fiarcn 19oo. £. J. Nowak, Test Plan: Brine Inflow and Humidity Experiments RoofO, Sandia National Laboratories, January 1988. in 9 Si E-103 FIGURES 1 Flux to a circular tunnel or borehole. ... 2 Apparent permeabilities based on BSEP data. SZ 4/ Si:: 4 Brine permeabilities derived from In situ experiments. 5 Geometry for radial flow to an Isolated tunnel 6 Geometry for steady flow to a line sin)c 7 Geo-.etry for lateral flow to an isolated room 8 Geo-etry for lateral flow In an array of rooms 10 11 12 13 14 15 16 Calculated brine accumulation In a typical waste disposal roc- in a panel; Po - hydrostatic pressure; K • 1 nanodarcy Calculated brine accumulation In a typical waste disposal roo- in a panel; Po - lithostatic pressure; )C ■ 1 nanodarcy. Calculated brine accumulation In a typical waste disposal roc- in a panel; Po - hydrostatic pressure; K • 10 nanodircy Calculated brine accumulation In a typical waste disposal roc- in a panel; Po - lithostatic pressure; K • 10 nanodsrcy Sensitivity of calculated brine accumulation to Initial pore pressure; typical room In a panel; K ■ 1 nanodarcy Sensitivity cf calculated brine accumulation to initial pore pressure; typical room In a panel; )C - 10 nanodarcy Sensitivity of calculated brine accumulation to host rock permeability; typical room In a panel; Po - hydrostatic pressure Sensitivity of calculated brine accumulation to host rock permeability; typical room In a panel; Po • lithostatic pressure ^7 ^? 7/ 73 7f 8C SI 21 83 85- %k 87 E-104 FIGURES 19 Effect of adjacent rooms In a panel on calculated brine accumulation; Po - hydrostatic pressure; K - 10 nanodarcy , 20 Effect of adjacent rooms In a panel on calculated brine accumulation; Po • llthostatic pressure; K - 10 nanodarcy ffl ^0 K ^ E-105 ■mm* -r : i» TABLES 1 Observed flow rates for WIPP boreholes [30] S^ 2 Observed flow rates for WIPP boreholes and apparent permeabilities based on eq. (19) ^^ 4 Suiri^-ary of results for cumulative volume at 200 years 'i E-106 E.4 WIPP HORIZON GAS FLOW MEASUREMENT RESULTS SUMMARY THROUGH 1986 This subsection of Appendix E contains information and data on the WIPP facility horizon in situ flow tests and measurements conducted through 1986. Flow measurement tests can be grouped into three categories: 1984 tests, N1420 drift tests and first storage panel tests. The results of these tests are briefly summarized in the following excerpt. More detail on the 1984 and N1420 tests can be found in this SEIS Appendix E and Subsections E.5 and E.6. This subsection Is provided to support near-field permeability rates defined in the text. This subsection is excerpted from Appendices B and C from Stormont et al 1987 Summary of and Observations About WIPP F acility Horizon Flow Measurements through Ifll E-107/108 ^ ^ :ri ! II 'I E.5 1984 GAS FLOW MEASUREMENT TEST RESULTS This subsection of Appendix E contains Phase I test results of in situ gas flow measurement results collected in 1984. A summary of this test and its results can be found in this SEIS Subsection E.4. This subsection is presented to support near-field horizon permeability rates detailed in the text. This subsection was excerpted from Chapters 4 and 5 of Peterson et al., 1985, WIPP Horizon In-Situ Permeability Measurements . ' ' I E-133/134 * Seismic Methods A series of seismic tomography and refraction studies were conducted underground at WIPP (Skokan et al., 1988). The first study was to set up a tomographic array on an older pillars at the site, as bdicated by oosshatches m Figure. 1 An older pillar would have the more extensive ffaaures. This survey showed that the bterior of the pillar was homogeneous with respect to seismic velocities (east to west raypaths, 4570 m/s), which suggests that fracture zones have not developed withm the pillar. A skm of low velocity material (4350 m/s) up to 1 m deep has developed around the pillar. The physical process that produces this skb is not well understood, but b general, fracture density and the degree of saturation affect attenuation and velocity b fractured rock (O'ConneU and Budiansky, 1974). Al WIPP, this skb of slower velocities develops b response to a combbation of microfractur- bg, daantancy and dehydration adjacent to the excavation. These processes will have the similar ef- fect of baeasing resistivity, hence, this skb of slower velocities may correlate to the zone of higher resistivities observed b the electromagnetic surveys. In addition to seismic tomography, wc have utilized seismic refraction (one study b the pillar used above and the other along pillars between the oldest rooms of the facility (300 to 500 ft west of the test pillar described above. Figure. 2). The refraction surveys detected planar vertical boundaries withb the pillar parallel to the rib face. These boundaries represent fractures developing wilhb the DRZ. The University of Texas at Austin, Dept. of Civil Engbeering, completed the analysis of the Spectral Analysis of Surface Wave (SASW) testbg at WIPP. The SASW method is based on the analysis of surface wave velocities determbed between two pobts at varying distance along the ex- cavation surface. The varied distances allow the velocity of the wave at different wavelengths to be determbed. As the wavelength bcreascs, the wave bteracts with rock more distant from the excava- tion; hence, a depth profile of velocity and rock moduli can be calculated. Analysis of three surveys along different excavation surfaces showed a systematic bcrcase b values of shear modulus and seis- mic velocity with depth bto the surrounding wall rock. The major baease b velocity and modulus occurs at approximately 1 m depth. E-288 FRACTURES Figure 1. Observed Fracture Pattern for 4 x 11 m Room 10* liftxxaTM Xr HIGH RESISTIVITY SALT u; I .«?•«€/ ■< "»«Tly^ — ■ : „ .. ■ f--— . .■-.wJ__'_'^ L~I[G[[": 75 RESISTIVITY SALT MEASUREMENTS AT WIPP SITE WITH EM 31 WITH EM 34 SAL2HAL0E RONNENBERC 10.0 Hgure 2. Location of EM-31 and EM.34 Traverses (In Black) and Seismic Surveys (crosshatched) 0.001 0.01 0.1 1.0 WEIGHT % HjO Figure 3. Apparent Resistivity versus Water Content •see detailed dcscripUon In the footnote below 3 :i ■• ■I n I • Explaaation of Figure 3. This figure dispUys the reUtionship between apparent resistivity and water content for different factors of cementation or consolidation (m) for Archies Law ( - 15 to 175 for Asse saJtfKessels et al., 1985]), crosshatched ranges are for resistiviUes of both Asse salt (high and low resistivity salt) and salt tailings pile (Salzhunde Ronnenbcrg) in which the water content was deter- mined indepcndcnUy, stippled ranges are apparent resistivities of salt at the WIPP facility horizon the water content of WIPP salt can be extrapolated from the intersection of the WIPP resistivities 'with the lines for the different consolidation factors E-289 63 60 1^ '9 14 16i 1 f_ 4M8r- 2655|' 229 9 i 2121 ' 101 3 V 6«0 191 "■t] O . n n I! in n^y ^r 1 110 30 A'^Vs'^V ■f\_f\- nJuLn 1 685S 2223 30S2 3032 31 7T 1^137 90 199 10 122 70 1^ 183 70 3069lf~l 2S9 5 2S5 4. HYDRAUUC TESTING 4.1 Gat-Flow Testing Gas flow tests at the WIFP site were con* ducted from horizontal, vertical, and angled boreholes drilled from the WIPP drifts (Slormonl ct al., 1987). Nitrogen was bjccted into the test mtcr."al, which was isolated by the packer system. The tests were cither constant pressure flow tests or pressure decay tests conducted from sbgle boreholes. The principal data from these tests arc flow rates from the test interval into the formation. For comparisons of flow rates measured during different condi- tions (test pressure and test interval size), flow rates have been normalized to a 0.07 MPa (10 psig) working pressure and a test interval of 1 m length and 13 cm diameter (Stormont et aL, 1987). The characteriza- tion of the DRZ based on the gas flow tests is as follows: E I*! ' ! Figure 4. Apparent Resltivlty Measurements In Ohm- Meters ' Within 2 m of the excavation, the DRZ is a zone in which baeascd flow rates are observed rela- tive to the far-Geld host rock. ® The inCTcase m flow rate within the DRZ appears to be a function of rocktype, size of the excava- tion, and age of the excavation. For example, an array of holes was drilled radially around a 3i x 6 m drift to a depth of about 10 m. Gas flow tests were conducted in numerous mtervals along each hole. Figure 6 presents values of nor- malized gas flow rates and the distribution of apparent resistivities around the NllOO drift at 4 years after cxca\-ation. The contours gas flow within the halite suggest a circular or elliptical pattern centered on the drift with flowrates deaeasbg radially outward from the excavation. Stormont et al. (1987) postulated that a partially-saturated dilatant zone surrounds the WIPP excavation. In par- ticular, a dilatant DRZ could account for gas flow m a formation that is thought to be brine- saturated in the undisturbed condition. The dilatant zone could include brine-saturated pores of sufficient size that their entry pressures are very low, permitting gas flow in our tests. An alternative explanation is that this zone is not completely brine-saturated, and the gas flows through the acces- sible gas-filled pore space. When the dilatant zone is created, accessible brine may be drawn into pore spaces by capillary pressure. Evaporation of pore brine, enhanced by mine ventilation , will create, maintain, or expand a partially-saturated zone. 4.2 Brine Injection Tettlng The shortcomings of gas flow measurements for determining permeability for a fully or partially saturated rock were recognized (Stormont ct al, 1987), and a limited number of brine bjcction tests vere conducted to compare/contrast with gas flow tests (Peterson et al., 1987). These tests were per- formed in two 10 cm diameter boreholes: a horizontal borehole with the test interval located in a relatively pure halite bed at a distance of 9 m from the nearest excavation, and a borehole angled 45° with respect to horizontal with the test interval located in an 1 meter thick anhydrite layer (Marker Bed 139). This angled borehole intercepts MB139 12 meters from the nearest excavation. Prior to the brine bjcction tests, 20- hr duration eas flow tests were conducted. Gas flows was measured which corresponds to permeabilities of 10 darcy and 10 darcy for the test btervals comprised of halite and anhydrite, respectively, assumbg the flow paths were gas-saturated. Subsequent brbe testbg b both holes lasted 220 days. In borehole DPHOl, a 3 J MPa bjcction pressure was held essentially E-290 "to I 10 M •OUNCEI / (H) M 5» to 41 40 JJ M Jl 2oX| 10 S ' ■ ■ ' 'II I 1 / ' , r Ve ' 4.3 km/t to C140 Ve « 4.$ km/t @) Sandia Natana) Laboraiones Hgurt 5. S«lsmlc Tomography RayPaths In the North End of the Trial PUIar, CompressionaJ Velocities are 4 J kin/j within a 1 m thick outer sUn of the pillar. coostant for 13 days. The test region was then shut-in for foUowing 13 days. The test region was sub- sc^uentlythcn shut-in for the remainder of the test, and the lest region pressure inaeascd to almost J?* cfX^ consistent with a brine-saturated formation with a pore pressure of 82 MPa, r.T. ^% i • "^J ".c cxcavaUoo remains open). The fractures that axe observed around the openings could contHbulc approximately an additional 2 can of vertical closure and 1 cm of horizontal closure. 6. SUMMARY ';"hc strucJuxes developed witliiii the DRZ are characterized by me$o«opic and miaoscopic frac- turicg b both the halite and anhydrite interbeds at the facility horizon m response to stresses developed during excavation or excessive strain mduced by ceep and the release of pore pressure. The rock salt m the ribs develops nearly vertical fractures parallel to the drift due to the low radial streiscs near the ribs. These fractures may become extensive enough to result in spalling. Within the •Active Opening,* slratigraphic layers that have undergone stress relief will respond by beam buck- ling to aeep of the rocksalt outside the zone of streis relief. The magnitude of the structures developed within the DRZ appears to be a function of both the size and the age of the opening. E-294 REFERENCES Baar, C. A^ 1977. AppUcd SaJt-Rock Mechanics 1; The in-situ behavior of saJt rocki, Elsevier Amslcrdam, 294 p, ' ^ "AfSi^'' ^^' ^^^^" ^^^ ^^ ^ ^^"^ °^ DriUcore from a Systematic Array. Saadia Report, Eonu, D. J., and J. C Stormont, 1988a. An interim report on excavation effect studies at iJbe Waste IsolaUon PUo( Plant: The delineation of the disturbed rock rone, Sandia Report, SAND87-U75 Bonis, D. J, and J. C. Stormont, 1988b. DUalancy and Fracturing in the Disturbed Rock Zone and us contribution to the total observed closure in WIPP excavations, memo to distribution, Sandia Nauonal Labs, Oaobcr 7, 1988 Brace, W. F, B. W. Paulding, and C. Scholz, 1966. DUatancy in the fracture of crystalline rocks. J Geophys. Res., 71:3939-3953 / uc«,v. Brady. B. H. C, and E. T. Brown, 1985. Rock Mechanics for Underground Mining. George Allen and Unwin, London, 527p Coatei, D. F, 1981. Rock Mechanics Principles, CANMET, Energy Mines and Resources Canada. Ottawa, Monograph 874 Dusseault, M. B, L Rothenburg, and D. Z. Mraz, 1987. The design of openings using mulUpIe mechanism viscoplastic, 28th US Symposium on Rock Mechanics, 633-642 Franckc, C, 1987. Excavation Effects and Fracture Mappbg, input to 1987 GFDAR, memorandum to R. McKinney, IT Corporation, WIPP site, August 18, 1987 Holcomb, D. J, 1988. Crossholc measurements of velocity and anenuation to detect a disturbed zone in salt at the Waste Isolation PUot Plant, in Cundall, P. A^ Sterling, and StarCeld, A. M., cds. Rock Mechanics: Proceedings of the 29th U. S. Symposium, 633-640 Kcsscli, W, I. nentge, and H. Kolditz. 1985. DC geoelectric sounding to determine water content in the salt mine ASSE (FRG): Geophysical Prospecting. 33:436-446 Mraz, D, 1980. Plastic behavior of sail rock utilized b designing a mining method, CIM Bulletin. 73:11-123 Munson, D. E, T. M. Torres, and R. L Jones, 1987. Ps«udostrain representation of multipass ex- cavauons in salt, in Fanner, I. W. ci aL, cds, 28th U. S. Symposia on Rock Mechanics, Tuscon. 29 Juae-UaJy 1987, 85^862 «,i««n, ^°T?^^^.'** ^' ^' ^^"^^^ "<* R- Beraun, 1988. Brine Inflow to WIPP Disposal Rooms: Data. Modelmft and Assessment, Sandia Report, SAND88-0112 O'ConncH, R. J, and B. Budiansky, 1974. Seismic Vcloddcs in dry and saturated cracked soUds. Journal of Geophysical Research, 79-.5412-5426 Peterson, E. W, P. L Lagus, and K. Kie, 1987. Ruid now Measurements of Test Series A and B of tte Small-Scalc Seal Performance Tests, Sandia Report, SAND87.7041 iV' »; ^"^ ^^^' Multicomponent Underground DC Resistivity Study at the Waste IsoUtion cv i ^ T ^ Southeast New Mexico, M. S. Thesis, Colorado School of Mines, T-3372 bto\aa, C J. Starrett, and H. T. Andersen, 1988. Fmal Report: Feasibility study of seismic tomog- raphy to momtor underground piUar integrity at the WIPP site. Sandia Report, SAND88.7096 btoraiottl, J. C Peterson, E. W, and Lagus, P. U 1987, Summary of and Observations about WIPP racmty Horizon flow Measurements through 1986, Sandia Report, SAND87^)176 I E-295/296 ■CI antilil •■-1**! .Jf»() lll'9' I E.9 SEAL DESIGN AND EVALUATION This subsection of Appendix E evaluates the design concepts for the tunnel and shaft seals required for the WIPP, as they are presently envisioned. The principal design strategy involves the use of salt as the primary structural seal material, relying on creep closure of the surrounding host rock to compress this salt into a low-permeability plug. Key elements of the supporting experimental program are also outlined. This subsection consists of Stormont (1988), entitled Preliminary Seal Design Evaluation for the Waste Isolation Pilot Plant. E-297/298 S3 liKi SAND87-3083 Unlimited Release Printed March 1988 Distribution Category UC-70 PRELIMINARY SEAL DESIGN EVALUATION FOR THE WASTE ISOLATION PILOT PLANT J. C. Stormont Experimental Programs Division Sandia National Laboratories ABSTRACT s^alinrof' tlTtV'c.^ preliminary evaluation of design concepts for the eventual - ifv TK ''' "^''J'"' '"^ boreholes at the Waste Isolation Pilot Plant ty The purpose of the seal systems is to limit the flow ^of water into 'Ugh, and out of the reoositorv tHp nr;n.:«oi A^.-.r, _.. .''/'"'"' Faci hrn ai' 7 P^^Po^^ ot the Seal systems is to limit the flow ^of water into through, and out of the repository. The principal design strategy involves' he consohdat.on of crushed or granular salt in response To the closure of th m ur r^and" ^^•^•Other candidate seal materials are bentonite, cementitiou studL7 '.r P^^'^'y ^^Phalt. Results from in situ experiments and modeling ence .;/ ^a\ '!, ^^^^^'^'^'V materials testing and related industrial experi non W..7. "".' ^ fo develop seal designs for shafts, waste storage panel entryways men^aTnrop ^^^^'T' -f-'I''^ '"' boreholes. Key elements of the ongoing exper^ mental program are identified. & & ^ I- E-299 Aw* CONTENTS Page 1. PURPOSE OF THE REPORT 1 2. SITE STRATIGRAPHY 2 3. FACILITY DESIGN ^ 4. FUNDAMENTAL SEALING CONCEPTS 6 5. SEAL FUNCTIONS AND REQUIREMENTS 7 5.1 Seal Functions and Requirements Inferred from Site Performance Assessment Scenarios ^ 5.1.1 Undisturbed Scenario ^ 5.1.2 Human Intrusion Scenarios * 5.2 Discussion of Seal Functions and Requirements 9 5.3 Working Criterion • '^ 6. CANDIDATE SEAL MATERIALS ^2 6.1 Salt '2 6.2 Bentonite ^^ 6.3 Cementitious Materials J^ 6.3.1 Grouts ° 6.3.2 Concretes ' ' 6.4 Asphalt '^ 7. DESIGN EVALUATION OF SHAFT SEALS ^1 7.1 Shaft Sealing Strategy ^^ 7.2 Shaft Seals in the Rustler ^^ 7.2.1 Bentonite Design ^^ 7.2.2 Concrete Design zl 7.2.3 Sealing the Rustler Rock ^' TO 7.3 Shaft Seals in the Salado fr 7.3.1 Salt Seals ^^ 7.3.2 Bentonite Design 7.3.3 Concrete Design ^ 7.3.4 Sealing the Salado Rock ^^ 7.4 Design Options Including Asphalt -^ E-300 CONTENTS (Continued) Paee 8. DESIGN EVALUATION OF PANEL SEALS 35 8.1 Panel Sealing Strategy 35 8.2 Panel Seal Design 3^ 8.2.1 Salt Seal Design ''''"^^'''^'^'"^^^^^ 36 8.2.2 Salt/Bentonite Seal Design 38 8.2.3 Rock Sealing 3g 8.3 Design Options Including Concrete 42 9. DESIGN EVALUATION OF NON-WASTE ROOM SEALS 45 9.1 Non-waste Room Sealing Strategy 45 9.2 Non-waste Room Seal Design 45 10. DESIGN EVALUATION OF BOREHOLE SEALS 46 10.1 Borehole Sealing Strategy 4^ 10.2 Borehole Seal Design 4^ 10.3 Design Options Including Crushed Salt 45 11. CONCLUSIONS .n 49 11.1 Materials Development 5q 11.2 Formation Hydraulic Properties 50 11.3 Seal Tests ,^ 11.4 Seal Design and Modeling ^0 1 1.5 System Integration ^j 12. REFERENCES I E-301 LIST OF FIGURES Page k,.ii> mtUU ••'•ji If 2.1. Generalized WIPP Site Stratigraphy 3 3.1. Plan View of the Proposed WIPP Facility 5 3.2. WIPP Facility Stratigraphy ^ 6.1. Permeability Versus Fractional Density for Two Consolidation Tests on Wetted Crushed Salt ^^ 7.1. Schematic of Design Concepts for Sealing the Rustler 23 7.2. Possible Shapes for Concrete Seals (from Sitz, 1981) 24 7.3. Finite Element Mesh and Stratigraphy for the Nonsalt Seal (from Van Sambeek, 1987) 26 7.4. Lift Temperatures in the FWC Nonsalt Seal (from Van Sambeek, 1987) 26 7.5. Contact Radial Stress in the FWC Nonsalt Seal (from Van Sambeek, 1987) 27 7.6. Design Concepts for Sealing the Salado 29 7.7. Sensitivity of salt consolidation in the WIPP shafts to brine inflow from overlying water-bearing zones (from Nowak and Stormont, 1987) 31 7.8. Configuration of Modeled Concrete Seal in Top of Salado (from Van Sambeek, 1987) 32 7.9. Lift Temperatures in the ESC Salt Seal (from Van Sambeek, 1987) 32 7.10. Contact Radial Stress in the ESC Salt Seal (from Van Sambeek, 1987) 33 8.1. Tentative Locations of Panel Seals ^^ 8.2. Cross-Sectional View of Panel Seals 36 8.3. Fractional Density with Time for Seal Emplaced at 0.5 Years After Excavation (from Arguello and Torres, 1987) 37 8.4. Fractional Density with Time for Seal Emplaced at 10 Years (from Arguello and Torres, 1987) 38 8.5. Gas Flow Rates in Halite Test Intervals (from Stormont, Peterson and Lagus, 1987) 8.6. Flow Rates in Interbed Layers Within 2 m of WIPP Drifts When Tested at the Drift Center or Near or Just Removed from the Drift Edge (from Stormont, Peterson, and Lagus, 1987) 8.7. Drift- Width vs. Flow Rate From Tests on MB 139 (from Stormont, Peterson, and Lagus, 1987) 8.8. Nl 100 Drift Flow Rate (SCCM) Contours (from Borns and Stormont, 1987) '*' 8.9. Idealized Excavation Effects in a 4m x 10m Room from (Borns and Stormont, 1987) "*' 8.10. Fractional Densities of the Crushed Salt Core as a Function of Time After Emplacement (from Arguello, 1988) '*3 10.1. Backfill Density for a 3.66-m-Diameter Shaft (initial density = 0.60; from Torres, 1987) ^' E-302 LIST OF TABLES Pa«e 7.1. Concrete Seal Length Criteria (from Garrett and Pitt, 1958) 25 » E-303 PRELIMINARY SEAL DESIGN EVALUATION FOR THE WASTE ISOLATION PILOT PLANT 1. PURPOSE OF THE REPORT ,-••''' 'II This report is a preliminary evalua- tion of design concepts for the sealing of penetrations (shafts, drifts, and boreholes) at the Waste Isolation Pilot Plant (WIPP) Facility. This evaluation is a product of the Plugging and Seal- ing Program (PSP), an experimental program conducted by Sandia National Laboratories (SNL) for the Department of Energy (DOE). The goal of the PSP is to develop the design concepts, bases, and criteria for the effective, long-term sealing of the WIPP Facility. A final conceptual design evaluation providing all input and information from the PSP is required to support the 1993 DOE decision whether to convert from pilot plant status to an operating repository. This preliminary evaluation will Allow the application of results and experience to update the design concepts for the WIPP o Provide direction for the ongo- ing experimental program o Provide input for the decision for the first receipt of waste, presently scheduled for October 1988. This preliminary evaluation draws information and data principally from the PSP, although other sources have been utilized when appropriate. These other sources include other facets of the DOE's program investigating the suitability of the WIPP as a nuclear waste repository, other experimental programs for sealing nuclear waste re- positories in salt and other geologic media, and mining-related research and experience. Although substantial in- formation is available to support this evaluation, many data and models are not presently available or adequately understood. Thus, estimates, extrapola- tions from limited data, inferences, and judgements have been used in this evaluation. W E-304 2. SITE STRATIGRAPHY Sealing activities for the WIPP will be largely directed at the Rustler and Salado Formations (the generalized WIPP site stratigraphy is given in Figure 2.1). Some existing boreholes in the vicinity of the WIPP site will penetrate the formations underlying the Salado, and will therefore require sealing the Castile Formation. The Dewey Lake Red Beds above the Rustler and the Delaware Mountain Group below the Castile are zones in which a seal adds little to restricting transport because the zones themselves are relatively permeable compared to salt (Christensen, Gulick, and Lambert 1981). Ground Surface Feel Meiers 500- - 200 1000- - - 400 1500 - 2000- 2500 - 3000- 4000- 4500- - 1000 1200 1400 r 600 Figure 2.1. Generalized WIPP Site Stratigraphy. Mapping the shaft walls prior to liner installation provided a good record of the lithology of the Rustler Formation from 168 to 257 m below the surface. The Rustler lithology is very diverse, being composed of carbonates, sulfates (gypsum, anhydrite, and poly- halite), clastic rocks, and halite (US DOE, 1983; US DOE, 1984). The Rustler contains the 8 m thick, water- bearing Magenta and Culebra dolomite beds at 186 and 220 m below the sur- face, respectively. The Culebra is considered the most transmissive unit in the Rustler, with t r a ns mi s s i vi - tie 10 sin the range of 2 x m / s ( Me r c e r , 1 < 10 983) to 1 The transmissivities of the Magenta vary from 4 X 10"' to 6 x 10""^ m^/s (Mercer, 1983). In addition, the Rustler/Salado contact is transmissive in some locations in the vicinity of the WIPP site (Haug et al., 1986), but has not produced water in the WIPP shafts (US DOE, 1983; US DOE, 1984). Mechanical properties of the Rustler rocks have not been determined, but can be estimated from generic properties such as those compiled by Lama and Vuturkuri (1978) and Callahan (1981). The Salado Formation, from the base of the Rustler to 850 m below the sur- face, is primarily halite, but also includes thin beds of anhydrite, poly- halite, clay zones, and in some areas, potash minerals. Numerous excavations and boreholes have provided a detailed and extensive characterization of the WIPP Facility horizon stratigraphy (e.g., US DOE, 1986). Krieg (1984) pre- sents a reference stratigraphy and rock properties for the facility horizon, including the reference constitutive model for time-dependent salt deforma- tion (creep). Permeabilities of the Salado rocks calculated from tests made from surface wellbores are generally in the range of 10"'^ m^ or lower E-305 "• 'l tti ,iP' (Mercer, 1986; Peterson et al., 1979). Numerous gas permeability measurements have been made from the facility hori- zon. At locations well removed from the excavations, the inferred oermea- bilities are very low (<10~ ^ m"^); close to the excavation the permea- bility can increase 3 orders of magni- tude or more (Stormont, Peterson, and Lagus, 1987). Brine testing from the facility horizon indicates permeabili- ties consistent with the gas test values, but also a substantial pore or formation pressure (Peterson, Lagus, and Lie, 1987a). The Castile extends from the bottom of the Salado to 1220 m below the sur- face. It consists principally of thick anhydrite beds with some interbedded halite. Pressurized brine has been en- countered several times in the upper- most Castile anhydrite, and is believed to be contained within localized, iso- lated reservoirs that are chemically and hydraulically in equilibrium with their environment (Popielak et al., 1983). E-306 3. FACILITY DESIGN The WIPP waste emplacement rooms are being developed approximately 650 m below the surface in the Salado Forma- tion, about 400 m below the Rustler/ Salado contact, and 200 m above the Salado/Castile contact. The three shafts which currently afford access to the repository are: The Construction and Salt- Handling Shaft (C&SH), 3.7 m drilled diameter The Waste Shaft, 6.1 m slashed diameter The Exhaust Shaft, 4.6 m slashed diameter. All three shafts are lined through he overlying Dewey Lake Red Beds and he Rustler Formation, and have a liner :ey or foundation located nominally 8 m into the Salado. The liners in the /aste and exhaust shafts are concrete, -hiie the C&SH shaft incorporates a teel liner cemented in place. Details egarding the design, construction, and laintenance of the shafts are given in tie Final Design Evaluation Report (US 'OE, 1986). A fourth shaft, nominally •0 m diameter, is presently being con- ructed by up-reaming. Excavation at the facility horizon egan in 1982. With the exception of le Cc&SH shaft station, which was exca- ated by drilling and blasting, exca- Jtions have been created by mining achines (continuous miners). A plan ew of the underground development is ven in Figure 3.1, and can be divided to the experimental area to the north the shaft stations and the waste orage panels to the south. Within e experimental area, the drifts are two levels: to the west drifts have en developed at the disposal horizon- the east drifts have been developed their floor is about 2 m above the roof of the disposal horizon. The stra- tigraphy associated with the two levels is given in Figure 3.2. Information on the underground construction, completed and planned, is given by the US DOE (1986). Area ^-JUl rJ) Jf C»SH Shan Wail* Sh«(l m storage Am DODDD U D. DD ma DQD DDD _JDDD ODDDD ZDDDD. DDDD[ Eiperlmtnltl Ar«a TMU Ridloacllv* TttI Ar** Plannad - Exhaust Shaft DOGD OOODOODDDDDD DaQOOODaDDOOQ Figure 3.1 Plan View of the Proposed WIPP Facility. Waste to be stored/disposed at the WIPP will be contact handled trans- uranic (TRU) wastes in 55-gal drums and remote handled TRU waste in canisters. It is presently planned that the con- tact handled waste will be packaged and handled in groups of seven drums which will be stacked three high within the storage rooms. The remote handled waste will be emplaced in 91 cm diam- eter horizontal boreholes in the ribs of some of the contact handled TRU stor- age rooms. E-307 I r3 Ml-- Clear to grayish orange-pink halile, trace of dispersed polyhalile and intercryslalline clay. T Clear to grayish orange-pink halite, trace polyhallte and discontinuous clay stringers. Clear to moderate reddish-brown halile, trace to some polyhalite and Iraccol clay. Clear to moderate reddish-brown- halite, trace some polyhalite, anhydrite stringers near bottom of unit. II I I I I I I I I' Anhydrite underlain by clay seam (anhydrite "a Xv\ Clear to moderate reddish-brown halite, trace to some polyhalite and discontinuous clay stringers. T Typical 5.6 m High Simulated-Waste Experimental Room Clear to grayish orange-pink halite. Anhydrite underlain by clay seam (anhydrite ■■b").v Clear to moderate reddish-brown to medium gray halile, trace polyhalite and some clay stringers. /• Gray clay seam locally with anhydrite \ V\\\\\\\\\ ^ i 2.7 Qm 0.21 m t^%%K\\\\% Clear to reddish-orange halite, trace polyhalite. =//^/A^/A- =r/^ Clear halile, trace argllaceous material. Clear to reddish-brown argillaceous '— halite with discontinuous clay partings in upper half. Clear to reddish-orange halite, trace polyhalite. --/- >^ Reddish-orange halite, trace y^ polyhalite. Reddish-brown to bluish-gray / argillaceous halite. V Typical 4.3 m High Test Room -^z^ -^/A^/^ Clear to reddish-orange and reddish-brown halite, argillaceous in upper part, trace polyhalite. Clear to reddish-orange polyhalitic halite, locally grading downward to polyhalite. Clear to gray and reddish-orange halite, trace polyhalite and argillaceous material. 1.79 m ' 0.06 m « i 0.66 m f 2.12 m 2.12 m /,: • Top of orange band 1.82 m 1.39 m 0.76 m Figure 3.2. WIPP Facility Stratigraphy. E-308 4. FUNDAMENTAL SEALING CONCEPTS The basic goal of the sealing sys- tem is to minimize the release of radio- nuclides from man-made penetrations in the WIPP by limiting fluid migration in, through, and out of the repository (Stormont, 1984). The most challenging aspect of seal^ design is the requirement that it be effective for hundreds to thou- sands of years, greatly exceeding common engineering and construction demands or experience. Processes will have to be modeled well beyond periods of time for which direct observations can be made. Therefore, extrapolations of relatively short-term data using time-dependent models (whether simple or sophisticated) is inherent in the seal design process. There are a number of factors which serve as the foundation on which designs are based. These include (Stormont, 1984): The consolidation of crushed or granular salt. This material, which is a by-product of mining the repository, is expected to consolidate into a mass compar- able to intact salt under appro- priate conditions, restoring the excavation to a condition ap- proaching its undisturbed state. Seal designs incorporate crushed salt as the principal long-term seal material. The time-dependent plastic be- havior of the host salt. Salt Seals are not implied, defined or considered to be completely impervious structures. "Perfect" seals may not be practical or attainable, they are not verifiable, and they will undoubt- edly not be necessary--as indicated by previous consequence assessments (Stormont, 1984). creeps or flows under deviatoric stresses, which results in the time -dependent closure of the excavations and the "healing" of fractures under some conditions. Intact salt also possesses a low permeability. Emphasis on the chemical and me- chanical compatibility between the host formation and the seal in order to increase long-term stability of the seal system, reduce the burden on predictive modeling, and add confidence to long-term waste isolation. The use of crushed salt maximizes compatibility in the salt forma- tion. Multiple component seal systems. A multiple component seal design allows individual seal compo- nents to serve different func- tions, to be effective over different time spans, and to exist in different locations and formations in order to ensure sufficient redundant barriers are in place at all times. Designs are to be practical. Some of the seal system will be emplaced by commercial contrac- tors and the chance for success will be increased by the simplic- ity of the designs, and by uti- lizing modifications of and extrapolations from current in- dustrial capabilities. The seal systems for the WIPP can be grouped into shaft seals, panel entryway seals, non-waste room back- fill, and borehole seals. Because the requirements, functions, and designs of these subsystems differ, they are con- sidered as separate entities in this report. The backfill for the waste- containing rooms is presently not con- sidered part of the sealing system. ill E-309 5. SEAL FUNCTIONS AND REQUIREMENTS I i Kg, 1 In order to develop a rational seal design, quantitative requirements for seal system performance must be known. At this stage of assessment, specific requirements for sealing the WIPP have not been established. In this chapter, a perspective for how well the WIPP needs to be sealed is developed from estimates of seal functions and require- ments. The required performance of the seal system and its components must ultimately be developed from the perfor- mance assessments of the WIPP site sys- tem. These assessments will evaluate the system response of the repository to various scenarios and conditions and will compare the predicted radioactive releases to the applicable environmen- tal standards. As these performance assessments are not yet available, the preliminary designs and design concepts considered here have been developed in the absence of quantitative performance requirements. However, some insight may be gained by considering the sce- narios which have been developed for possible site performance assessments. Other factors, including the design phi- losophy for long-term waste isolation and binding agreements with the State of New Mexico, also contribute to pre- sent estimates of seal functions and re- quirements. In order to quantify seal performance requirements, a working criterion pertaining to effective salt consolidation has been developed. This criterion allows relevant seal design analyses to be conducted for both salt and nonsalt seal components. 5.1 Seal Functions and Requirements Inferred from Site Performance Assessment Scenarios Following Hunter's scenario devel- opment work for the WIPP site (Hunter, 1987), seal performance may be consid- ered in the context of two classes of scenarios: the "undisturbed" scenario and various human intrusion scenarios. The undisturbed scenario involves the predicted response of the disposal system without disruption by human intrusion or unlikely natural events. The human intrusion scenarios involve the disposal system response to the drilling of exploratory boreholes at the repository site, some of which provide fluids for the dissolution of waste or a pathway for the transport of radioactivity to the biosphere. 5.1.1 Undisturbed Scenario The undisturbed scenario involves numerous time-dependent processes which will impact the performance of the en- tire repository, and the seal system in particular. Predominant processes iden- tified to date include: o Closure of the excavations in the halite formations. This clo- sure tends to densify and con- solidate backfills, and induces buildup of stresses in the vicin- ity of stiff seal components. Brine influx from the host rock salt into the excavations in halite formations. This natural ly- exist i ng brine seeps into excavations and, given enough time, will accumulate in the void spaces remaining in the excavations. The brine may affect backfill consolidation, corrode waste packages, and, if present in discrete pockets, become pressurized in response to closure. Water inflow from the water- bearing zones overlying the salt vertically down the shafts. The amount and rate of water inflow presently observed is governed by the performance of the shaft liners, but ultimately will be E-310 dependent on the performance of the shaft seal system. In addi- tion to the possibly deleterious effects of brine mentioned above, this water, being unsatu- rated in halite, may dissolve substantial quantities of salt. o The creation of a disturbed zone surrounding excavations. The po- tential for flow in these zones can be significantly greater than in the undisturbed rock, and seal bypass can occur. Gas generation by the waste. This gas may accumulate in waste rooms, potentially slowing room closure and backfill consolida- tion. The undisturbed scenario includes these processes, and their synergism and extrapolation to long periods of time. There is presently sufficient uncertainty associated with this sce- nario that a wide range of site perfor- mances can be postulated depending basically upon the efficacy of the seal- ing systems. Hunter (1987) proposed an undisturbed scenario that will be eval- uated for its potential to provide a radioactive dose to members of the pub- lic. First, water from leakage through the shafts or from the Salado forma- tion into the repository is postulated. Water in contact with the waste then dissolves radioisotopes, producing a solution of radioactive brine that occupies the remaining available void space in the repository. The continued closure of the excavations may then pressurize the brine pockets if they exist and force fluid from the reposi- tory through available paths to the biosphere. Possible paths are through the host rock (salt and/or clay and anhydrite seams) and the shaft seal system. In such a sequence of events, if the sealed shafts are substantially more permeable than the formation, they may be preferential paths for water movement. The amount of wafer they can allow in and out of the repository and not violate the applicable standards will be the subject of the eventual performance assessments. Effective panel seals separating volumes of waste from one another and from the shafts will also provide substantial resis- tance to flow through the repository and therefore represent another signif- icant barrier to waste release. Non- waste drift backfills would eventually serve a function similar to that of panel seals after sufficient reconsoli- dation. A variation of the undisturbed sce- nario that could result in the release of radioactivity from the repository in- volves existing boreholes, which would serve as the source of water and/or the path for contaminated brine. No exist- ing boreholes penetrate the repository, so they are inconsequential unless they facilitate introduction of water to the repository. To do so, the flows estab- lished in the boreholes must dissolve the salt that separates the borehole and the repository. In boreholes that intersect the water-bearing strata only above the repository there is no circu- lation of water, consequently the disso- lution is controlled by diffusion and proceeds so slowly as to pose no threat to the WIPP even if the boreholes re- main open (Stormont, 1984). Boreholes connecting water-bearing strata above and below the repository can dissolve salt faster because of the circulation established between them. Conservative calculations reveal that open boreholes of this type 300 m horizontally from the bounds of the repository will not intercept the repository for millions of years (Stormont, 1984). Sealing will further slow or prevent flow in the boreholes. 5.1.2 Human Intrusion Scenarios Numerous scenarios that involve human intrusion can be postulated. The 3 E-311 I s I I' Environmental Protection Agency (EPA) regulations suggest that future inadver- tent intrusion by exploratory drilling for resources can be the most severe scenario assumed in a performance as- sessment (US EPA, 1985). These explor- atory boreholes, assumed to be drilled between 100 and 10,000 years after the decommissioning of the repository, can be combinations of holes that may or may not penetrate the repository and/or intercept pressurized brine reservoirs or other water-bearing strata under- lying WIPP. The boreholes that do not penetrate the repository may serve as shortened flow paths between the reposi- tory and the overlying water-bearing zones (Hunter, 1987). When a borehole intercepts the re- pository, the conditions and properties existing in the repository at the time of intrusion will govern the resulting response. Depending on the condition of the waste and surrounding backfill, radioactive cuttings and drilling mud may be released to the surface, or the penetration may go undetected. Con- tinued drilling into underlying strata may allow the introduction of pressur- ized brine into the repository. Panel seals will limit effect to one panel and will isolate this panel from sub- sequent intrusions in other panels. In addition to the possibility of future boreholes penetrating the affected waste panel(s) and allowing releases, contaminated brine leaving the reposi- tory through the shaft seal system must be considered. In the human intrusion scenarios, the panel seals may have a role in lim- iting the consequences of intrusions by isolating volumes of waste from one another. The shaft seals have not been explicitly included in these particular scenarios. However, because the condi- tion of the repository is in part depen- dent on the performance of the shaft seals, their performance is implicitly included in all the scenarios. 5.2 Discussion of Seal Functions and Requirements Consideration of hypothetical sce- narios that may result in radioactive releases to the public provides some concept of seal requirements. Shaft seals may be required to limit the vol- ume of water introduced to the reposi- tory from the overlying water-bearing zones. A further requirement may be to limit the amount of contaminated brine that could move up the shaft to either the surface or the overlying water- bearing zones. Regardless of human intrusion scenarios, shaft seals will require a relatively rigorous design because: (1) shafts serve as a direct connection between the overlying water- bearing zones, the surface, and the repository; (2) shaft seals have to per- form immediately after installation to limit the inflow; (3) shaft seals expe- rience decreasing benefit from creep in the upper portions of the shafts; (4) shaft seals must be effective in the di- verse geologic conditions through which the shafts pass; and (5) there is a lirn- ited opportunity for full-scale experi- mental design validation. The role of the panel seals is less obvious. Their performance may not be required unless the repository is breached by future exploratory bore- holes, at which time they will serve to isolate volumes of waste from one another and the shafts. However, their contribution as a redundant barrier should not be overlooked, especially because the results of direct experimen- tation and observation are available. The necessity for non-waste drift backfill is not immediately obvious from the previously considered scenar- ios. However, regardless of the sce- nario, non-waste drift backfill will serve as a redundant barrier to fluid migration, limit damage around excava- tions, shorten the time until the repository is returned to a condition E-312 comparable to intact rock, limit sub- sidence and its accompanying effects, and serve as a disposal location for mined salt. The requirements for seals in exist- ing boreholes are expected to be mini- mal. Conservative calculations reveal that even open existing boreholes would not be expected to result in a signifi- cant radiological dose to the public basically because they do not penetrate the repository. Because the performance assessment activity is not complete, new scenarios or processes may conceivably arise that place more or different requirements on one or more component of the seal sys- tem. Furthermore, regardless of the outcome of the performance assessments, the WIPP should be sealed to the extent deemed effective and practical at the time of decommissioning because (Stormont, 1984): A cautious and conservative approach is appropriate when public health and safety are involved Sealing will add confidence in the long-term isolation of waste, and reduce public concern regarding long-term hazards o Sealing the penetrations is con- sistent with the multiple bar- rier approach mandated by EPA standards. Finally, the Department of Energy and the State of New Mexico have en- tered into a binding agreement that requires the inclusion of certain seals in the WIPP: "DOE shall use both engi- neered and natural barriers to isolate the radioactive waste after disposal in compliance with the Environmental Pro- tection Agency standards. The barriers shall include, at a minimum, properly designed backfill, plugs, and seals at the drifts and at panel entries, and plugs and seals in the shafts and drill holes" (US DOE and State of New Mexico, 1981). Thus, there is a legal DOE commitment to seal the WIPP in addition to technical considerations and EPA standards. 5.3 Working Criterion In order to conduct meaningful anal- yses in the absence of final perfor- mance requirements derived from site performance assessments, a preliminary "working" criterion is required. For this purpose, the preliminary design criterion has been defined as the re- quirement for effective crushed salt consolidation at panel entries and in portions of the shafts. A crushed salt criterion was selected because it is the fundamental element of the long- term sealing strategy: if salt con- solidates to a condition comparable to the intact salt, the result is con- sidered to be the ultimate long-term seal. Further, salt consolidation analyses embody many of the reposi- tory's time-dependent processes and will provide an opportunity to model these processes. A consolidated salt seal design criterion also permits requirements for other seal components to be estimated by determining the time and degree of isolation from water necessary to allow crushed salt to con- solidate effectively. The criterion is considered satis- fied when the porosity of the crushed salt decreases to 5 percent or less. Available data suggest that as the po- rosity decreases to about 5 percent, the permeability of the crushed salt is reduced to s ubmic rodarcy values (Holcomb and Shields, 1987; IT Corp., 1987). Such a low permeability makes the crushed salt a relatively good barrier to fluid flow. In fact, at this porosity the permeability of the crushed salt approaches that of intact salt, Thus, this preliminary criterion E-313 I s Ci '2' is probably conservative, as future requirements cannot reasonably require a seal system to have a lower permea- bility than the intact host rock. The 5 percent specification has an impor- tant practical application. The pres- ent constitutive model for crushed salt consolidation indicates that the crushed salt will offer very little resistance to the continued closure of the excavations until the porosity of the crushed salt decreases to 5 percent or less (Sjaardema and Krieg, 1987). Thus, as an analytical convenience, drifts and shafts containing crushed salt backfill can be modeled as open drifts until they become effective seals. Obviously, the estimated time re- quired for the crushed salt to achieve satisfactory consolidation is of in- terest. Also important is the time and condition (porosity) at which the crushed salt becomes saturated with water liberated from the intact salt or with that flowing along the penetra- tion. The water in the pore space could resist or retard further consoli- dation, and if the porosity is greater than 5 percent, significant connected porosity may persist. If so, the par- tially saturated crushed salt could become a preferential flow path, degrad- ing a component of the long-term seal- ing strategy. A design criterion that involves salt consolidation allows requirements for other seal components to be in- ferred. The principal function of most non-salt seals is to limit the amount of water that reaches the crushed salt while it's consolidating. As will be subsequently discussed, present esti- mates of times to achieve effective salt consolidation are <100 years at the disposal horizon and in the lower portions of the shafts. Given the present inability to predict durability or longevity for seal materials other than salt, limiting the timeframe for the required performance of these seals to periods within reasonable engineer- ing experience is crucial for the credi- bility of the design. In addition to the need for other seal components to protect salt during consolidation, other seal materials are included in seal designs because: (1) crushed salt will not be consolidated by creep clo- sure in those portions of the shafts which pass through nonsalt formations; (2) crushed salt is not an effective short-term barrier; (3) other seal materials may have desirable properties not possessed by crushed salt; and (4) redundancy in the design can be provid- ed by including other seal materials. Ill E-314 6. CANDIDATE SEAL MATERIALS Following is a discussion of the various candidate seal materials: salt, bentonite, cementitious materials, and asphalt. The best possible seal mate- rial would return a penetration to a condition comparable to its undisturbed state within a predictable period of time. 6.1 Sajt Salt has the potential to be an ef- fective, simple seal material. Experi- mental evidence suggests that granular or crushed salt consolidates under cer- tain conditions, resulting in a poros- ity and permeability that decrease toward values comparable to intact salt. For crushed salt emplaced in an opening in a rock salt formation, the consolidation is driven by the creep closure of the adjacent host rock. The time-dependent properties of crushed salt have been measured by numerous laboratory researchers. At a given stress, the single most impor- tant parameter in the consolidation of crushed salt is the presence of a small amount of water. Small amounts of water accelerate consolidation and the accompanying permeability decreases in comparison with dry crushed salt (Holcomb and Shields, 1987; IT Corp., 1987; Shor et al., 1981; Pfeifle and Senseny, 1985). The effects of other variables, such as particle size, are secondary and not as obvious. The dependence of salt consolida- tion on added water can be illustrated by considering the experimental results of Holcomb and co-workers (Holcomb and Hannum, 1982; Holcomb and Shields, 1987). The 1982 tests were conducted on dry (no additional water) crushed salt, whereas the 1987 tests involved small amounts (<3% w) of additional water. The volume strain data, dV/V , ably described by (Holcomb and Hannum, 1982; Holcomb and Shields, 1987) dV/V^ = a log t + b (I) where a and b are fitting constants and t is time in seconds. The constant, b, is a measure of the initial condition of the sample (Holcomb and Shields, 1987). To compare times to achieve the same volumetric strain for tests under similar initial and loading conditions. Equation (I) can be rewritten as t2^V3l) = t,. (2) The constant, a, for wet test data is five to ten times greater than from a comparable dry test. Therefore, for dry crushed salt to experience the same strain under similar test conditions requires a time five to ten orders of magnitude greater than that for the wet sample. Sjaardema and Krieg (1987) devel- oped and implemented a constitutive relationship for the consolidation of crushed salt based on the laboratory data of Holcomb and co-workers. Numeri- cal calculations of wet crushed salt consolidation in WIPP shafts and drifts were then conducted to determine the in- fluence of the presence of the crushed salt on the closure of the shafts and drifts. Up to a fractional density of 0.95 (the extent of the laboratory data the model was based on), the results indicate that no substantial backstress (resistance) develops in the crushed salt. That is, the closure is largely unaffected by the presence of crushed salt. As expected, as consolidation pro- ceeds, the permeability of the crushed salt decreases. In general, permeabili- ty values for samples with a fractional density of 0.85 or less are milli- ^ O..U... wcid, UT/ r^, uciibiiy ui V.6D or less are mi 1 1 1 - from both sets of data can be reason- darcy or greater values (10"'^ m^ E-315 I s 1 a" •ei. or greater). Between fractional densi- ties of 0.85 and 0.95, however, the per- meability drops dramatically. By 0.95 fractional density, the permeability of the crushed salt is on the order of that of intact salt. Figure 6.1 shows permeability versus fractional density for two tests that proceeded to high fractional densities (Holcomb and Shields, 1987; IT Corp, 1987). A simi- lar trend of a dramatic permeability decrease at 0.95 fractional density has been observed in experiments on calcite to simulate the alteration of permea- bility and porosity of rocks by plastic flow processes (Evans, 1983). E n ra 0) E a o ■11 -12 -13 -14 -15 -16 -17 -18 -19 -20 -21 IT Corp. (1987) - L Holcomb -^ ~ and Shields (1987) J L 0.75 0.80 0.85 0.90 0.95 Fractional Density Figure 6.1. Permeability Versus Fractional Density for Two Consolidation Tests on Wetted Crushed Salt. The exact mechanism(s) of consoli- dation are not understood. Clearly, water plays some important role. Yost and Aronson (1987) dismiss dislocation mechanisms of creep as a primary mecha- nism of consolidation of wet salt, and suggest pressure solution and/or the Joffe effect as the dominant mecha- nism(s). Holcomb and Shields (1987) discuss the possibility of a pressure solution mechanism for consolidation in view of their experimental data, and conclude that further investigation is required. Post-test analyses were conducted on consolidated samples (IT Corp, 1987), and it was concluded that water played an important role in salt consolidation (and the accompanying permeability decrease) by facilitating pressure solution. Zeuch (1987) adapt- ed a model for isostatic hot-pressing to the consolidation of nominally dry crushed salt, and found good agree- ment between the model and Holcomb and Hannum's laboratory data. Interesting- ly, this model predicts consolidation approaching intact salt densities over periods of less than 50 years under approximate repository conditions, in contrast to simple extrapolations of laboratory data. The model is present- ly being expanded to include the influ- ence of water. While small amounts of water have been determined to benefit consolida- tion, larger amounts may be detri- mental. It is conceivable that if the salt becomes saturated while substan- tial porosity remains, further con- solidation could be impeded by the low compressibility of the entrapped brine (Nowak and Stormont, 1987). Previous tests by Baes et al., (1983) indicate that brine can be readily squeezed out of salt so as to not impede consoli- dation even to low permeabilities. Preliminary results by Zeuch (1987) suggest that saturated crushed salt consolidates similarly to crushed salt with much less water. However, these laboratory tests have been on vented samples; it is not obvious to what degree brine in large emplacements will be expelled during consolidation. E-316 Another advantage of crushed salt is its availability and low cost. Gran- ular salt is a by-product of the exca- vation of the WIPP Facility, and is therefore in plentiful supply. Future operations may wish to consider under- ground stockpiling to limit handling of the mined salt. Because the time required for crushed salt to become an effective seal is dependent on its initial den- sity, the emplacement method can have a large impact on the sealing function. The options for emplacement include dumping, dumping with compaction via vi- brating tampers or rubber-tired trucks, pneumatic stowing, or the placement of pre-compacted blocks. With the ex- ception of the blocks, commercially- available equipment exists for these emplacement techniques. Based on adobe technology, Sandia has developed a pro- totype machine that presses blocks of salt (and other materials) for use as a seal material (Stormont and Howard, 1987). For the possible emplacement techniques mentioned above, a reason- able range of fractional densities is from 60 to 85 percent. The 60 percent fractional density was obtained from crushed salt poured into molds in the laboratory (Holcomb and Hannum, 1982). The 85 percent fractional density is achievable with the Sandia Block Ma- chine (Stormont and Howard, 1987). In- terestingly, it was necessary to add 1 to 3 percent water to produce coherent blocks. Block properties are given by Gerstle and Jones (1986) and Stormont and Howard (1987). An alternative to crushed salt as a seal material is intact or quarried salt blocks. These intact blocks have higher fractional densities than pre- compacted blocks of granular salt, and the time required for them to become an effective seal is correspondingly reduced. The permeability decrease expected in a quarried salt seal as the adjacent rock tends to creep in may be similar to the "healing" of salt sam- ples brought to the laboratory from the field. Initial permeabilities are relatively great due to sampling dam- age; after application of hydrostatic pressure for only a short period of time, permeabilities decrease to low values (Sutherland and Cave, 1980). The interfaces between blocks may heal readily, as evinced by fracture healing studies in salt (Costin and Wawersik, 1980; IT Corp, 1987). Salt is easy to cut and machine, and blocks have already been fashioned from 41 cm di- ameter cores simply using a band saw. Seals constructed of intact salt blocks require stock material, and block ma- chining would be labor intensive; there- fore these alternatives are presently envisioned for limited applications where time to effect a salt seal must be minimized. 6.2 Bentonite Clays have found many applications as fluid barriers in underground excava- tions (e.g.. National Coal Board, 1982; Sitz, 1981), as components of earth dams (e.g., Sima and Harsulescu, 1979), and in containment of hazardous wastes (e.g., Johnson et al., 1984; Leppert, 1986). In particular, sodium bentonite is under consideration as a seal mate- rial for geologic nuclear waste reposi- tories (e.g., Pusch, 1987; Stormont, 1984; Lopez, 1987; Ke 1 s a 1 1 et al., 1982). Bentonites are composed princi- pally of montmori lloni te, a smectite mineral responsible for their charac- teristic swelling. Bentonite mixed with filler or ballast material is being considered as a seal material as a matter of economy, as well as to mini- mize the loss of the bentonite through small fractures or cracks. Sitz (1981) found that the sand in a bentonite/sand mixture stopped bentonite losses through fractures with a maximum width of 2 to 4 mm. The permeability of mixtures of ben- tonite and various filler materials has been measured by numerous investigators E-317 I 1 s < >■ in the laboratory (e.g., Radhakrishna and Chan, 1985; Wheelwright et al., 1981; Peterson and Kelkar, 1983; Stroup and Senseny, 1987). There is consider- able variability in the data due to differences in test methods, sample den- sity, working fluids, etc. In general, the permeability of the mixtures to water and brine was found to fall off to microdarcy or lower values somewhere between 25 to 50 percent bentonite by weight, probably coincident with the bentonite becoming the continuous phase of the mixture (Nowak, 1987). Pusch (1987) determined that the permeability of bentonite to brine is about an order of magnitude greater than that to fresh water. Another important property for mix- tures containing bentonite is the swell- ing pressure developed when the mixture is confined and saturated with water. Swelling is expected to fill voids and heal fractures within the bentonite seal and perhaps to a limited degree in the adjacent host rock. The average swelling pressure of confined 100 per- cent bentonite in salt water was given by Pusch (1980) as Ps = e _ ^lI.5(rho-1.87) (MPa) (3) where p^ is the swelling pressure and rho is the bentonite bulk density between 1.8 and 2.1 g/cc. Gray, Cheung, and Dixon (1984) demonstrated that swelling pressures of bentonite mix- tures are dependent on the effective clay density, that is, the mass of the bentonite divided by the volume of the bentonite and any voids. Thus, the sand or other filler material is merely an inert filler. Bentoni te/sand or bentoni te/salt mixtures could be emplaced in much the same way as crushed salt: mechanical- ly, pneumatically, or in pre-compacted blocks. Blocks of 50 percent benton- ite/50 percent salt and small amounts of water have been pressed to a dry den- sity of about 1.97 g/cc, and an effec- tive clay density of 1.6 g/cc (Stormont and Howard, 1987). At these condi- tions, a swelling pressure of about 2 MPa and a brine permeability of about 10" m are expected. Drift em- placements of bentonite mixtures in the Stripa Facility were accomplished with vibrating tampers and a robotic pneu- matic machine (Pusch, 1987). Bentonite has also been emplaced and tested as a borehole seal (Pusch, 1987; South and Daemen, 1986; Kimbrell, Avery, and Daemen, 1987). Bentonite slurries have been suggested as a rock mass grouting material (Meyer and Howard, 1983). Soil structures (including clays) can fail in the presence of seepage by erosion along pre-existing cracks or piping (internal retrogressive ero- sion). The predominant factors in- volved in failure by both mechanisms are (Resendiz, 1976) loosening of inter- particle coherent forces upon satura- tion (dispersivity), permeability, and swelling potential. The risk of fail- ure is increased as the first two fac- tors increase and the third decreases. Clays rich in mont mor i 1 1 ini te (e.g., bentonite) are generally too expansive to permit cracks to remain open and too impervious to allow seepage veloci- ties large enough to induce piping (Resendiz, 1976). Further, bentonite is relatively plastic and can withstand considerable deformation prior to fail- ure. The tendency for erosion or piping failures is increased at the interface between the clay and dissimilar mate- rials (Penman and Charles, 1979), i.e., the seal/rock interface. Pusch, Borgesson, and Ramquist (1987) demon- strated the effectiveness of bentonite in effecting a tight interface by swell- ing. Pusch (1983) investigated the possibility of the migration of benton- ite into rock fractures, and the sub- sequent erosion of the bentonite by flowing groundwater. He concluded that bentonite will migrate a few tenths of meters into fractures wider than 0.1 mm over the course of thousands of years, and should not be significantly eroded E-318 by groundwater. Because swelling is a time-dependent phenomenon, the rate of introduction of water prior to satura- tion may be significant. Stormont and Howard (1987) emplaced and tested 50 percent bentonite/50 percent crushed salt seals in 1 m diameter boreholes in the WIPP Facility. Failure by erosion was observed when water was introduced rapidly to one face of the seal; a rela- tively low permeability seal was estab- lished in a similar seal configuration when the water was introduced at a slower rate to permit a gradual uptake of water. Clays exist naturally in geologic formations, including bedded salt, and are therefore appealing as long-term seal components. Clay sealants have been used by man for long periods of time; Lee (1985) documented the effec- tiveness of a clay sealant for periods as long as 2100 years. While bentonite alteration to other clays does occur under some conditions, at non-elevated temperatures bentonite transformations are expected to be very slow, on the order of millions of years (Meyer and Howard, 1983; Roy, Grutzeck, and Wakeley, 1983). Krumhansl (1984) found from experiments in WI PP-spec i f i c aqueous solutions that bentonite is expected to maintain its desirable min- eralogic characteristics indefinitely. 6.3 Cementitious Materials Cementitious materials have been considered as a candidate repository seal material because (Lankard and Burns, 1981): (I) cementitious mate- rials possess favorable seal properties such as low permeability and adequate strength; (2) there is a historical precedent for sealing penetrations with cementitious materials; (3) much physi- cal and chemical properties data exist; and (4) construction with cementitious materials is an established practice with a large number of equipped, quali- iTied and available commercial contrac- tors. Since 1975, cementitious seal materials have been developed and stud- ied for the WIPP. Early work focused on development of grouts for borehole sealing, with more recent research being devoted to concretes for sealing shafts and drifts. Research on rock fracture grouting has been initiated for the WIPP. 6.3.1 Grouts Cementitious grouts have been uti- lized for many years to seal surface- drilled wellbores for disposal of chemical and toxic wastes and to seal abandoned oil and gas boreholes. Typi- cally, few problems are encountered but quantative measures of seal effective- ness are generally not available (South and Daemen, 1986). Emplacement technol- ogy for borehole sealing with cementi- tious grouts is available (e.g.. South, 1979). Recent testing has provided more information about the effective- ness of cementitious borehole seals. The Bell Canyon Test, conducted in bore- hole AEC-7 near the WIPP site, involved the placement of a 2-m-long grout seal at a depth of 1370 m in anhydrite host rock, isolating the upper portions of the borehole from the 12 MPa Bell Canyon aquifer. The plug reduced the production of the aquifer by five orders of magnitude, and analyses indi- cated that the predominant flow occur- red through the plug/borehole interface region (Christensen and Peterson, 1981). In situ tests in granite show that cementitious plugs placed with conventional methods reduce the hydrau- lic conductivity of the we II bo re to or less than that of the host rock (Kimbrell, Avery, and Daemen, 1987). Laboratory tests by South and Daemen (1986) indicate the effective- ness of cementitious grouts as a seal material in basalt, granite and tuff. Large flows along the interface have been observed during a laboratory test on a grout-sealed hole in anhydrite (Bush and Lingle, 1986); the sealing E-319 rM I 1 £ < effect of a grout plug in rock salt was considered to be much better in a companion test (Bush and Piele, 1987). Gulick and Wakeley (1987) provide the reference formulations and proper- ties for candidate grouts for use in sealing the WIPP. Both a freshwater (BCT-IFF) and saltwater (BCT-IF) grout have been selected. A saltwater-based grout is necessary in the host rock salt to preclude dissolution of adja- cent rock during hydration. The prop- erties of the fresh water grout are considered somewhat more favorable. The BCT-IFF has been emplaced in the Bell Canyon Test, in portions of the C&SH shaft liner, in the upper portions of borehole B-25 on the WIPP site, and in an underground test bank for curing candidate seal materials (the Plug Test Matrix). The BCT- 1 F mixture has been emplaced in borehole B-25 and in the Plug Test Matrix. The properties of the BCT- IF and BCT-IFF grouts have been determined under a range of condi- tions, and are summarized by Gulick and Wakeley (1987). Subsequent to the development of the BCT grouts, modi- fications have been proposed (e.g., Wakeley, Walley, and Buck, 1986; Buck, Boa, and Walley, 1985; Buck, 1985; Buck et al., 1983; Wakeley and Roy, 1985). However, because there is no identifi- able deficiency of the BCT grouts and the advantages of the other formula- tions have not been shown, the BCT grouts remain the reference materials for the WIPP. Another potential use of cementi- tious grouts in sealing the WIPP is grouting fractures in the host rock. Grouts for this application are expect- ed to be thinner than the BCT grouts. Control of inflow to the existing WIPP shafts has been attempted in part by rock grouting with cementitious mix- tures. Rock grouting with cementitious mixtures has been used to control inflow to shafts (e.g.. Hart, 1983), in conjunction with establishing concrete seals in shafts and drifts (e.g., Auld, 1983; Garrett and Pitt, 1958; Garrett and Pitt, 1961), and with dams. The complicated system of a curing grout injected into poorly characterized fractures has generated a technology laden with empiricism (e.g., Dept. of the Army, 1984) and controversy over techniques and claims of effectiveness. Rock fracture grouting may be detri- mental in some instances: fractures may propagate from injection pressures, and water pressure buildup from seal- ing drainage paths may be sufficient to further fracture the host rock. Schaffer and Daemen (1987) considered rock fracture grouting technology for repository sealing applications, and concluded that "considerable and well- recognized uncertainty exists about the actual performance of grouting." 6.3.2 Concretes Concrete has historically been used as a seal and shaft liner material because of its availability, relatively low cost, and familiarity among con- tractors and mine operators. Further- more, properties of standard concretes such as strength and permeability are generally understood and considered adequate for typical seal applications (National Coal Board, 1982; Auld, 1983). Unfortunately, there is little documentation of the design and perfor- mance of concrete seals. The few refer- ences to concrete seals in the mining industry must be considered in the con- text of their application: these seals are usually emplaced in response to an inrush of water, and a substantial reduction in leakage is considered suc- cess. In what is believed to be the only documented tests on experimental full-sized drift seals, Garrett and Campbell Pitt (1958, 1961) demonstrated the effectiveness of concrete seals as fluid barriers in quartzite host rock. Auld (1983) cites examples of the suc- cessful placement and performance of concrete seals in a sandstone and a gypsum and marl deposit. Sitz (1981) provides a summary of German experi- ences with concrete seals in various rock types, describing both successes E-320 and failures of concrete seals. Con- crete seals have been successfully uti- lized in tuff as containment structures for underground testing at the Nevada Test Site (Gulick, 1987). The single consistent conclusion from historical experience is that con- crete itself is relatively impermeable, and that observed leakage is predomi- nantly attributable to the concrete/ rock interface and the near-field rock. Probable causes for flow at the inter- face are concrete shrinkage, poor rock quality, and interaction between the concrete structure and the host rock. In non salt host rock, there are two potential remedies to ensure a tight interface: the use of an expansive con- crete and contact or interface pressure grouting. Expansive concretes have been developed in the laboratory (e.g.. Buck, 1985); however, experience with placement of numerous full-size drift seals in tuff with supposedly expansive concretes is inconclusive with regard to net expansion (Gulick, 1987). Pres- sure grouting along the concrete/rock contact has been demonstrated to be effective in substantially reducing the leakage along the interface, and is considered standard practice in the placement of concrete seals (Garrett and Pitt, 1958; Garrett and Pitt, 1960; Auld, 1983; National Coal Board, 1982; Gulick, 1987; Defense Nuclear Agency). In halite, creep of the adjacent host rock may result in a tight rock/ concrete interface. Reference formulations and proper- ties of candidate concretes for the WIPP are given by Gulick and Wakeley (1987). A saltwater-based concrete (ESC) and a freshwater concrete (FWC) were selected. The ESC is an expansive (in laboratory tests), salt-saturated concrete which has been emplaced in two seal tests in the WIPP (Stormont, 1986; Stormont and Howard, 1986) and in the Plug Test Matrix. The performance of the ESC material has been adequate structurally (Stormont, 1987; Labreche and Van Sambeek, 1987) and exceptional as a fluid barrier (Peterson, Lagus, and Lie, 1987b) in the field tests. Its properties have been extensively tested in the laboratory and are given in Comes et al. (1987), Wakeley and Walley (1986), and Wakeley (1987). The FWC is based on an expansive concrete developed by Buck (1985) for nonsalt host rock applications. A thermomechanical model for the ESC was developed based on the results of the in situ seal tests (Van Sambeek and Stormont, 1987; Labreche and Van Sambeek, 1987). The model results show excellent agreement with the measured temperature changes from hydration and fair agreement with the measured strains and stresses in the seal and the adjacent rock. The assumed expan- sivity of the concrete was found to be the parameter that influences the short-term model results the most and is the least well understood. Numeri- cal modeling of panel seals has uti- lized the elastic properties of the ESC (Arguello, 1987; Arguello and Torres, 1987); both the ESC and FWC time- dependent properties have been applied to numerical studies of shaft seals' structural interactions and stability (Van Sambeek, 1987). Large volume pours of concrete will be required for drift or shaft seals. This existing emplacement technology uses standard commercial equipment and techniques (e.g., Defense Nuclear Agency). In situ seal tests conducted at the WIPP have successfully employed gravity-feed by tremmie for small-scale shaft seals and pumping into a formed interval for small-scale drift seals (Stormont, 1986; Stormont and Howard, 1986). A principal concern regarding the use of cementitious materials as a seal material for nuclear waste repositories are their durability or longevity, Ce- mentitious materials will not be in chemical equilibrium with their environ- ment (Lambert, 1980a). Potential miner- alogic phase changes could manifest E-321 I 1 (J) £ < S3 19 themselves as: (1) the formation of a soluble, friable, or permeable phase in the plug or nearby rock; (2) shrinking or degradation of adhesion, opening the interface between the seal and the rock (Lambert 1980a, Lambert, 1980b). On the other hand, there is evidence for the longevity of cementitious materials in certain environments. Evaluation of some ancient cementitious materials reveals they have survived in appar- ently good condition for centuries (Malinowski, 1981; Monastersky, 1987). Research on the durability of cementi- tious mixtures applicable to the WIPP is generally favorable with regard to expectations or speculations about the maintenance of long-term properties (Buck, 1987; Wakeley, 1987b; Burkes and Rhoderick, 1983; Wakeley and Roy, 1986; Roy, Grutzeck, and Wakeley, 1983). Yet it is known that concrete is suscepti- ble to degradation, especially in envi- ronments with high sulfate contents (Lea, 1971) such as Culebra formation water (Mercer and Orr, 1979). Hart (1983) reports that concrete liners which pass through the formations above salt mines in the northeastern U.S. degrade or corrode from formation water leaking through the liner, resulting in a reduction of the concrete thickness of about 3 mm per year. An examination of a 20-year old shaft liner in the Carlsbad potash district suggests that the concrete liner has appreciably deteriorated from sulphate attack (D' Appolonia, 1981). Heimann et al. (1986) demonstrated that the presence of clay accelerates the dissolution of some cements. There is presently no comprehensive model of the complicated system of cementitious materials, the host rock, the formation water, and their inter- actions sufficient to make reliable predictions of long-term (thousands of years), time-dependent performance. In- deed, the problem is so multi-faceted, large, and diverse (involving kinetics, thermodynamics, and chemistry) that resolution of all issues seems remote. Therefore, reliance on cementitious materials as long-term seal materials should be minimized. Emphasis on con- solidated salt as the long-term seal will relieve the requirement for con- crete effectiveness to perhaps a few hundred years. 6.4 Asphalt Asphalt is a bituminous material produced by the distillation of crude oil. In the construction industry, asphalts have a wide variety of applica- tions because they are durable, highly waterproof, strong, and highly resis- tant to the action of most acids, alka- lies and salts (Herubin and Marotta, 1977). Bacterial degradation requires microorganisms and moisture; even if these conditions are present, the deg- radation is expected to be very slow (ZoBell and Molecke, 1978). Many prop- erties of asphalt, including density and viscosity, can be tailored by the distillation process and by the addi- tion of weighting materials and blend- ing and dissolving agents. Liquid asphalt has been utilized as a key component in the construction of waterproof liners in strata overlying salt and potash deposits (Hart, 1983; Wegener, 1983). A method successfully employed in German mines is described by Wegener (1983). A precast concrete block liner is fixed to the rock con- current with shaft construction. A steel cylinder is then emplaced in the shaft so as to leave a gap or annulus between the concrete blocks and the steel. A reinforced concrete liner is then cast on the interior of the steel cylinder. Finally, asphalt with a spe- cific gravity 30 to 40 percent greater than water is poured into the annulus up to the surface, so asphalt tends to move out into the formation rather than formation water tending to move into the shaft. Asphalt is added at the sur- face to replace that which moves into the formation. Wegener (1983) reports that two such shaft liners recently E-322 installed are ". . . absolutely imper- meable to the water from surrounding strata." Such a liner design is being used in the shafts of Germany's pro- posed radioactive waste disposal facili- ty at Gorleben. Sitz (1981) describes the use of asphalt as a component of an elaborate seal for an underground gas storage facility in domal salt. Over- pressure of the asphalt is achieved by pipes from the surface in contrast to an open volume of asphalt. Solid asphalt, or asphalt cement, has also been used in waterproof liners and drift seals. The liner key is often located in the saliferous forma- tion, and it is imperative that water does not flow behind it or the entire shaft liner may fail by washout or dissolution. Special care is taken to seal the liner at the key, includ- ing the use of asphalt cement (e.g., Wegener, 1983; D'AppoIonia, 1981). Solid asphalt has also been used in con- junction with drift and shaft seals in salt or potash mines in Germany (Sitz 1981). Previous WIPP seal concepts have not included asphalt, and the experimen- tal program has not evaluated asphalt as a candidate seal material. However, a large experience and data base exist from applications at other facilities and could be readily applied to the WIPP situation. Asphalt warrants con- sideration as a possible seal material based on its successful applications, especially in Germany. Its present role in WIPP seal concepts is as a po- tential redundant component. E-323 7. DESIGN EVALUATION OF SHAFT SEALS I I 4 Shaft sealing strategy and designs are considered separately for the Rustler and Salado formations. Benton- ite and concrete are the principal seal materials in the Rustler, where treat- ment of the disturbed rock zone may be the most difficult sealing problem. In the Salado, salt is the principal long- term seal material. 7.1 Shaft SealinR Strategy The fundamental strategy for seal- ing the WIPP shafts is to maximize the amount of consolidated salt between the repository horizon and the top of the Salado Formation. In this way, the long-term seal is essentially identical with the host rock, and the otherwise very difficult issue of seal longevity is averted. Shaft seal performance can then be evaluated in the context of salt consolidation; that is, the time to achieve satisfactory consolidation can be used to estimate the type, num- ber, and required performance of other seal components. Furthermore, effec- tive salt consolidation achieved prior to 100 years after decommissioning is independent of breach scenario assump- tions. To ensure effective consolidation, unacceptable amounts of water must be prevented from accumulating in the crushed salt. There are three possible sources of water: the overlying water- bearing zones, the host rock salt, and the repository. Water, if present, could be forced up the shafts from the repository horizon by closure or by some breach event. This suggests a seal at the base of each shaft to elimi- nate a preferential flow path up the shafts prior to effective salt consoli- dation. Water influx from the host rock salt will be difficult to limit along the entire length of the shaft in the Salado Formation. An annular seal may limit the flow into the crushed salt, but it would be at odds with the fundamental strategy of monolithic salt as the long-term seal. The crushed salt could be protected from the over- lying water-bearing zones by seals in the top of the Salado, seals in the lower portions of the Rustler, or both. Placing seals in the lower portions of the Rustler is intuitively obvious, because these seals would be as close as possible to the source of water (the Culebra and Magenta dolomites, and pos- sibly the Rustler/Salado contact). How- ever, the Rustler lithology is very diverse, being composed of carbonates, sulfates (gypsum, andhydrite, and poly- halite), clastic rocks, and halite (US DOE, 1983; US DOE, 1984). Such vari- ability may be troublesome if, for example, seal design requires a certain length of seal in the same rock type, or if a detailed understanding is need- ed of the interaction between the seal material and multiple host rocks. Some of the weaker rocks in the Rustler may be adversely affected by the excavation and subsequent redistribution of stress- es, resulting in seal locations which are weak and a potential source of by- pass. Further, some of the clastic rocks such as siltstones and sandstones in the Rustler are susceptible to ero- sion, which could result in relatively large flow along the seal/rock inter- face. The Salado formation may be a more favorable environment for seals, because it has a more uniform strati- graphy and the stratigraphic units are thicker than the ones in the Rustler. The predominant rock type is halite, which has many properties considered favorable for sealing (low permeabili- ty, fracture healing, and plastic de- formation). Moreover, the experience and data base for salt is large, be- cause the vast majority of the seal tests are being conducted with halite as the host rock. The principal con- cern with sealing in the Salado For- mation is the solubility of halite. The water of the Culebra and Magenta E-324 dolomites is not saturated with respect to NaCI and is therefore able to dis- solve salt. Even brine which is satu- rated at standard conditions may be capable of dissolving salt due to the pressure and temperature dependence of salt solubility. Concern that the initial seepage behind WIPP waste and exhaust shaft liners could progress enough to threaten the stability of the liner keys (located in the top of the Salado) has led to remedial grouting programs in these shafts. In the waste shaft, drill holes that penetrated the liner/salt contact produced an esti- mated 0.03 m^/hr (Sauliner and Avis, in preparation). In the exhaust shaft, pre-grouting activities indicated some fluids at the concrete/salt contact (US DOE, 1987). Salt dissolution behind liners in US Gulf Coast mine shafts requires more than half of all shafts to undergo maintenance (principally grouting) to preclude unacceptable in- flows (Hart, 1983). Sitz (1981) reviews attempts to seal salt and potash mines in Germany, and concludes that "due to the solubility of the saliferous sys- tem, the greatest problems occur in the construction of plugs and dams in potash and rock salt mining." The preceding discussion indicates that there are advantages and also prob- lems to overcome in sealing either the Salado or the Rustler to limit inflow down the shafts into the crushed salt seals. A prudent approach is to not place total reliance on either system, but to include seals in both regions. This approach is consistent with the concept of multiple barriers. 7.2 Shaft Seals in the Rustler A simple model of flow through seal systems in the Rustler was constructed by Stormont and Arguello (in prepara- tion) to provide information relevant to shaft seal design. The model pro- vides one-dimensional flow through the seal material, the seal/rock interface. and the adjacent rock (the so-called disturbed zone) at 14 intervals between the Magenta and the top of the Salado. Concrete and bentonite-based materials were input as the seal components. Also input were various cases of seal mate- rial and rock performance (principally permeability) estimated from available measurements. Combinations of seals with varying seal and rock performance were examined via the model, and the flow rate through the seal system was compared with estimates of allowable flow into the lower portions of the shaft in the Salado (the allowable inflow was based on a study of salt consolidation in the lower portions of the shafts, and is discussed further in Section 7.3). The analysis provided the following conclusions: o The quality (essentially the per- meability) of the rock adjacent to the seal is the single most important factor in maintaining a low flow rate through the seal system. Even with perfect seal- ing of the shaft itself, large flows bypassing the shaft seals through the adjacent rock are possible, especially if vertical- ly persistent fractures exist, o A very small gap at the con- crete/rock interface can allow substantial flow through con- crete seal systems. The assumed degradation of con- crete seals may render concrete structures ineffective as flow barriers even when their initial permeability is low. Including bentonite in the seal design can obviate the above con- cerns over concrete seals if the bentonite is located in a low permeability host rock and does not appreciably degrade with time. E-325 I s S3 The conclusions from this study sug- gest that emphasis should be placed on establishing seals of low permeability and long-term durability against rock which has little potential for vertical flow or seal bypass. This approach is consistent with the undisturbed state of the Rustler: the rocks between the water-bearing zones and the top of the Salado have low vertical permeabilities (Sauinier and Avis, in preparation). Thus, the intent for seals in the Rustler is to reestablish the natural low permeability of portions of the formation. Bentonite-based seals, if adequately confined, should be satisfac- tory. Anhydrite and claystone are two potential rock types in which such a seal can be located. The low vertical permeability of the Rustler has been attributed in part to anhydrite (Barr, Miller, and Gonzalez, 1983). Anhydrite is a strong rock, and its disturbed zone may be limited and well defined. Claystone is more similar to the seal material than is anhydrite, and there- fore increases compatibility. Low permeabilities have been measured in Rustler claystone within 2 m of the shaft wall (Sauinier and Avis, in preparation). A schematic of the design concepts for sealing the Rustler is given in Figure 7.1. The principal seals are constructed from bentonite-based and cementitious materials. Above the top of the Magenta dolomite, the shaft with the existing liner left in place is filled with locally plentiful material, including a clay fraction to reduce the permeability of the mixture, if desired. Owing to the relatively high t ra ns missi vi t y of these strata, there is little motivation to establish a low permeability seal in this location. Be- tween the Magenta dolomite and the top of the Salado are three bentonite-based seals that abut against anhydrite and claystone layers. These are the prin- cipal fluid barriers in the Rustler. Concrete in the shaft between the ben- tonite seals confines the bentonite, provides structural strength for the 200 m 260 m Rutller Seal System Magenta Dolomite Antiydrlte General Fill to Surface Perhaps Grouted to Reduce Settlement ■'-'I^mW^^ --------------- Salado Seal System Begins Figure 7.1. Schematic of Design Concepts for Sealing the Rustler. system, and acts as a redundant flow barrier. Grouting of the concrete/rock interface or contact is specified. For- mation grouting is provided in some locations to seal disturbed zones adja- cent to the shafts, where possible. It is expected that the shaft liner will have to be removed at locations adja- cent to the bentonite seal locations to permit removal of damaged rock and pre- vent the degraded liner from becoming the predominant flow path through the seal system. Whether or not the rest of the liner can remain in place will depend on the function and shape of the adjacent seal and the condition of the E-326 liner. Note that there is no intent to establish a tight seal in the water- bearing zones themselves because this would require very extensive and diffi- cult treatment (grouting the rock), which would probably divert the water around the seals into lower portions of the shafts. 7.2.1 Bentonite Desig n The lengths of the bentonite-based seals are more than 4 m, and exceed an empirical guideline for a minimum length of 2 m for clay seals (National Coal Board, 1982). The shape of the seals is expected to be cylindrical, with a diameter determined by the removal of fractured host rock. The bentonite will be mixed in approxi- mately equal proportions with a filler material to increase its strength and limit losses through cracks or fractures. The filler could resemble Pfeifle's (1987) silica sand used as a filler with bentonite. A permeability of 10" m was shown to dramati- cally reduce the flow through a model seal system in the Rustler (Stormont and Arguello, in preparation), and should be achievable for such a mix- ture. An in place density of about 1.8 g/cc for a 50/50 mixture should result in a swelling pressure of less than 3 MPa, limiting the potential for damage to the confinement (rock or concrete) and the bentonite's propensity to mi- grate from the seal interval through fractures in the rock or along the rock/concrete interface. The water content should be on the order of 10 percent, to reduce the likelihood of piping failure and to limit drying shrinkage. 7.2.2 Concrete Desig n Two obvious design considerations are the shape and length of concrete seals. As shown in Figure 7.2, there are many possible shapes for concrete seals in the Rustler, from simple cylin- drical or parallel shapes to multiple element, truncated-cone shapes. For strength considerations, the parallel shape is generally considered adequate (National Coal Board, 1982; Auld, 1983) and is the most often employed. How- ever, Sitz (1981) argues that parallel shape will result in an unfavorable stress state upon loading sufficient to cause failure of the seal. In fact, he attributes some noteable failures to the parallel shape. Nevertheless, more recent and complete analyses have not confirmed his results (Van Sambeek, 1987). Parallel Abulmenli t»* Slnglt-Parnllel Interlocked Abutmenit Truncaled-Cone-Shaped Abutmenit Slnglr-Truncst»d-Cen«- Shaped Mulllplf-Truncalad'Con*- Shapad Calotte Stielts Figure 7.2. Possible Shapes for Concrete Seals (from Sitz, 1981). Seal length can be determined by means of leakage or structural con- siderations. From their tests on drift seals, Garrett and Pitt (1958; 1961) regard length as governed by leakage, rather than by structural considera- tions. Garrett and Pitt developed concrete seal length criteria to estab- lish the point at which leakage becomes excessive, based on allowable pressure gradient across the seal and given as a E-327 I ra function of the contact (interface) and adjacent rock grouting associated with the concrete seal (see Table 7.1). Garrett and Pitt stressed that these criteria are applicable only to the particular rock conditions under which the test was conducted (relative- ly strong and intact quartzite). They recommended that safety factors of at least four and up to ten be applied to these criteria to account for uncertain- ties in rock conditions and the design function of the seal. They concluded that the principal factor in a seal's performance is the condition of the host rock. This is borne out by the dramatic increase in the allowable pres- sure gradient across a seal when the host rock is extensively grouted. Auld (1983) recommends grouting at pressures up to one and one-quarter times the hydrostatic pressure for contact with weaker rocks. The most important point is the dramatic influence of interface contact and adjacent rock grouting on concrete seal performance. The concrete seal has to be able to support the imposed axial load, which will be a combination of the weight of overlying seal materials and water, and possibly the load generated by expan- sive bentonite seals directly adjacent to the concrete seals. Simple formulae for determining the necessary length which assume a frictional contact along the interface or direct bearing on the inclined surfaces of asperities along the contact are of limited practical value, as they bear little resemblance to the actual state of stress in the concrete and adjacent rock (Sitz, 1981). Numerical studies offer the potential for a more rigorous treat- ment of the strength and stability of concrete shaft seals. Van Sambeek (1987) conducted a nu- merical analysis of an unsupported 10 m long, 7 m diameter concrete shaft seal located at the base of the Rustler. The host rock for the seal was assumed to be sandstone, and neighboring layers of anhydrite and salt were included (Figure 7.3). The modeling of the con- crete (FWC, see Chapter 6) accounted for the time-dependent elastic modulus, thermoelastic expansion, time-dependent chemically induced expansion, and creep of the concrete. The general model of the concrete behavior was based on (9 Table 7.1. Concrete Seal Length Criteria (from Garrett and Pitt, 1958). ^ minimum p/I ratio where p is hydraulic pressure and 1 is seal length MPa m"' (lb in"^ ft"') No grouting of interface or adjacent rock 0.21 (10) Interface only grouted at hydrostatic pressure 4.72 (228) Interface grouted at hydrostatic pressure, and adjacent rock grouted at twice hydrostatic pressure 8.28 (400) E-328 4 m [• 121 m mj i^ ^ll±J-^ 7h ^ ^ dr Depth (m) 181.7 ■- 189 1 T 215.4 221.9 2272 •- 234.4 ■- 258.0 ■ 310 6 •- 315.2 Dolomllr Anhydrite Dolomite Anhydrite Salt Slllstone Salt Anhydrite Figure 7.3. Finite Element Mesh and Stratigraphy for the Nonsalt Seal (from Van Sambeek, 1987). laboratory data that had shown fair agreement with in situ test results (Van Sambeek and Stormont, 1986; Labreche and Van Sambeek, 1987). Refer- ence properties were used for the rock (Krieg, 1984) or estimated from avail- able literature. Thermal analyses were first used to calculate the temperature rise in the concrete and the adjacent host rock resulting from the exothermic hydration of the concrete. The peak temperature for the concrete was esti- mated to be about 60° C (Figure 7.4), and the maximum penetration into the rock of the 3° C contour was about 4 m from the seal edge. Thermome- chanical analyses were then conducted to determine the state of stress and strain in the concrete and the adjacent rock from thermal expansion/contraction of the rock and concrete, chemical expansion of the concrete, and creep of the salt rock and the concrete. Radial and shear stresses at the contact ind 0.0 0.1 0.1 0.2 0.2 0.3 0.3 0,4 0.4 0.5 0.5 Time Alter Seal Emplacement (yr) Figure 7.4. Lift Temperatures in the FWC Nonsalt Seal (from Van Sambeek, 1987). tensile stresses within the concrete we re satisfactory with respect to preliminary criteria to judge the E-329 I IP effectiveness of the concrete seal. For example, the radial stress at the interface was compressive from emplace- ment on, indicating a tight interface (Figure 7.5). The concrete seal was then exposed to a 10 MPa axial load (simulating the swelling pressure of an adjacent bentonite seal), and the seal remained stable. However, when the assumed expansion of the concrete was neglected, the seal was not stable, even without the axial load. -I — II I iiii| I I I iiii| 1 — I I Min| I M m il A. -I I I 1 1 II 0.01 0.1 1 10 Time Alter Seal Emplacement (yr) 100 Figure 7.5. Contact Radial Stress in the FWC Nonsalt Seal (from Van Sambeek, 1987). It is apparent from the preceding discussion that the nature and condi- tion of the contact between a concrete seal and the host rock is an important factor in the strength and stability of a concrete seal in rock. If a good con- tact is provided (that is, if a substan- tial normal stress exists across the interface), then substantial strength in response to axial loads will be de- veloped. This conclusion has been sub- stantiated by laboratory push-out tests on borehole seals in rock (Stormont, 1983) and in intermediate, in situ seal tests (Stormont, 1987; Labreche and Van Sambeek, 1987). Further, as previously discussed, flow through concrete seal systems is reduced when a good contact has been provided. A satisfactory con- tact in these non-creeping host rocks could be provided by an expansive con- crete, extensive interface contact grouting, or constructing the seal in some favorable shape. The condition of the adjacent rock is another significant consideration in the performance of concrete seals. In addition to the influence of the adja- cent rock as a significant flow path which can bypass the concrete seal, the strength of the seal system may be developed by direct bearing on the inclined surfaces of asperities along the contact. Thus, the strength of the host rock may be a factor in the stability of a concrete seal. Keying or recessing the concrete seals into the rock may provide a better seal by removing heavily damaged (fractured) rock, to provide a stronger bearing surface if required, or even to create a more favorable seal geometry if desired. Limitations of such second- ary excavation are given in the next section. In summary, the first choice for the shape of the concrete seal remains a cylindrical or parallel seal shape. The shapes of concrete seals shown in Figure 7.2 are conceptual, to indicate that the shape may be something other than cylindrical, if necessary. The lengths of the concrete seals (>I0 m) are well within Garrett and Pitt's cri- terion with a safety factor of 10, and should be adequate if the concrete is expansive or the interface is grouted. Creating bearing surfaces by secondary excavation is a further option. 7.2.3 Sealing the Rustler Rock Water seepage into the WIPP shafts in the Rustler has been observed to a varying degree essentially from con- struction on (US DOE, 1986). The upper range of these rates tends to be be- tween 1000 and 2000 m^/year (US DOE, 1986; Haug et al., 1986). It has been estimated that these observed inflow rates would have to be reduced by a E-330 factor of up to 1000 to limit the satu- ration of crushed salt seals in the Salado so as to not impede consolida- tion (Nowak and Stormont, 1987). If a substantial portion of the observed inflow is through the damaged zone of the adjacent rock, effective sealing will require limiting flow through this damaged zone. In other words, no matter how well the penetration or shaft open- ing itself may be sealed, the potential for flow in the adjacent rock must still be addressed to limit flow to the top of the Salado. This conclusion is consistent with the shaft seal model study of Stormont and Arguello (1987), as well as with case studies of effec- tive sealing (e.g., Garrett and Pitt, 1958). Today's approach for reducing fluid seepage from the water-bearing strata into the shafts through the existing liners has been the application of ce- ment and chemical grouting. However, there may be difficulties and limita- tions for grouting applications with present technology in support of the eventual sealing of the WIPP shafts. Validated techniques for remote identi- fication of fractures and positive con- firmation of grouting effectiveness do not presently exist. Experience, notably at the WIPP, has shown that grouting often has to be repeated to obtain or maintain effectiveness. Finally, grouting may have to be effec- tive for up to 100 years, well beyond the currently designed longevity for typical materials and applications. Alternatives for the sealing of this region of rock include large cut- outs and overpressure systems. A suf- ficiently large cut-out would remove the damaged rock, and replace it with a material such as concrete. This con- cept has several difficulties, includ- ing determination of the distance into the rock such a structure should ex- tend, the actual construction if the damaged zone is large, and assuring that the excavation for the cut-out does not just extend the damaged zone farther into the rock. An overpres- sure system involves placement of a fluid in the shaft that is at a greater pressure than the water; flow is then from the shaft out into the rock, rath- er than the other way. These systems, employing viscous asphalt in the annu- lar space between the rock and liner, have been used with success in German salt mines. Limitations of this method for long-term sealing applications in- clude assuring that the overpressure is maintained and that an adequate supply of the sealing fluid is available, as it will flow out into the rock. 7.3 Shaft Seals in the Salado The design concepts for sealing the Salado are given in Figure 7.6. Most of the shaft will be filled with crushed salt consistent with the long-term shaft sealing strategy of maximizing the amount of consolidated salt be- tween the repository and the top of the Salado. Other seal materials are con- crete and bentonite/salt mixtures. At the top of the Salado, salt/bentonite fill is to be placed as a flow barrier and to saturate water moving down the shaft with salt. Salt/bentonite mix- tures are also to be placed against the few layers which are predominantly anhy- drite and thicker than 3 m because these intervals will not close from creep and crushed salt would not con- solidate in these intervals (if axial consolidation is ignored). Salt/benton- ite mixtures will act as a redundant flow barrier, and will perhaps seal the fractures which may result along the contact between halite and anhydrite, where large differential strains are expected. Salt/bentonite mixtures could also be placed to limit downward drain- age of water added to the large volumes of crushed salt if necessary. While salt/bentonite mixtures are expected to be an excellent shaft fill material, their applications are limited to select locations because consolidated salt will be an even better long-term seal and to preclude a substantial continuous phase other than monolithic E-331 Salado Seal System Begins I .J I s 300 m SCO m 600 m 700 m Mudslon* HallU Anhydrite Haiti* Anhydilte Halite Polyhalll* Halite Polyhallle Halite Polyhallle Halite Clayslone Halite Polyhallta Halite Polyhallte Halite Anhydilte Anhydilte Halite Polyhallte Halite Polyhallle Halite Polyhallte Halll* Polyhallle Halite Anhydilte Halite Anhydrite Halll* Salt/Benlonlle Mlituie ,,jV'?i;.\ Concrete ■•'>''^,,<'.j';-:-'. Bentonlle-Based Quarried Salt or Crushed Salt Blocks Benlonite-Based Crushed Salt Sall/Benlonlte Mlaluies Crushed Salt • Conceptually Similar >- to Composite Seal Centered at 304.8 m SSSSJSm^J^SSSS: Halite - - . . Anhydrite i^^P^tW Crushed Salt Salt/Bentonlle Mlitur* Salt/Bentonlte Mlalure Satt/Benlonlte Mlalure Figure 7.6. Design Concepts for Sealing the Salado. E-332 salt in the shaft. There are two bulkhead-type or composite seals located within the Salado. The first is located nominally 15 m into the Salado, and it is the Salado counter- part to the seals in the Rustler; that is, its principal function is to limit the flow of water down the shaft from the overlying water-bearing zones. The composite seal consists of concrete, be nt oni te/salt mixtures, and a salt component. The salt component may be quarried or intact machined salt to hasten its return to a state comparable to intact salt. A similar composite seal is located approximately 150 m above the repository horizon. This seal separates the crushed salt that is estimated to consolidate in 100 years or less (and is therefore independent of breach scenarios) from the overlying crushed salt, which will require longer periods of time. This depth is based on a study of salt consolidation dis- cussed in Section 7.3.1. A seal struc- ture is located at the base of the shaft to preclude substantial settle- ment or movement of the overlying back- fill. In addition to concrete, the base seal will have other components to restrict flow either up the shaft or down into the repository horizon. 7.3.1 Salt Seals Scoping model calculations of crushed salt consolidation in the WIPP shafts conducted by Nowak and Stormont (1987) utilize the working criterion for salt consolidation given in Section 5.3. The model couples simplified and idealized representations of shaft clo- sure, salt consolidation, brine influx from the host rock, and inflow from the overlying water-bearing zones. The model predicts the porosity decrease of the crushed salt due to closure con- current with the filling of the poros- ity from brine influx and inflow down the shafts. As a worst case , consolida- tion was assumed to cease when the salt became saturated. This assumption was made to allow simple, conservative calculations. In fact, it is expected that the greater rate of closure at depth may force fluid upward, so water saturation will not necessarily pre- clude consolidation. Future experi- mental studies will address this issue. Effective consolidation was assumed to be achieved when the poros- ity of the crushed salt decreased to 5 percent or less. The model provides conservative estimates of the final condition of crushed salt (saturated porosity) and the corresponding time needed to achieve its final condition as a function of depth. The representa- tions of closure, brine influx, inflow from the overlying water-bearing zones, initial porosity of the crushed salt, and time of emplacement after excava- tion were varied in order to assess the sensitivity of the model to these param- eters. Over and above revealing the sensitivity of the model to parameters such as closure and brine influx, con- clusions regarding the shaft seal design were reached. First, a prelimi- nary criterion for the allowable or target flow from the overlying water- bearing zones was developed. Figure 7.7 reveals the influence of inflow down the shafts on the length of the effectively consolidated salt column at the bottom of the shaft. Baseline values representing best estimates were selected for the other paranieters. If the flow is limited to 1 m /year or less, at least 100 m of salt will reach 5 percent or less porosity within 100 years. Another conclusion from this study is that the initial density of the crushed salt in the shafts should be as great as practicable to minimize the time to consolidation. In fact, the 100 m of consolidated salt in 100 years requires an initial density achievable only by salt blocks. It should be emphasized that the model is believed to be conservative; that is, the actual amount of consolidation is expected to be greater than the model predicts. However, it provides quanti- tative results that can be used to provide guidance for the experimental program, as well as design-relevant information. E-333 I II ;;r' a o 200 = 1 1 1 1 -■ Inflow to Sliafts from Overlying Water-bearing Zones (mVyr) 300 - y//\ Top of Salado — Effective Salt ^^ / / 1 Consolidation ^y^ / / 1 Workinc 1 Criterion ^^ y^ / / 400 = 0-;? //V 500 ^/O = 3 y/ / y/^ 0= 10 / 600 /// / y O = 100 / 1 1 Bottom of Shaft Q"= 1000 , , 700 1 1 0.00 0.05 0.10 0.15 0.20 0.25 Brine-Saturaled Void Fraction In Reconsolldated Crusfied Salt Figure 7.7. Sensitivity of Salt Consolidation in the WIPP Shafts to Brine Inflow from Overlying Water-Bearing Zones (from Nowak and Stormont, 1987). 0.30 There are presently no estimates of the time required for quarried salt to achieve its final condition, but it is assumed to be less than that for crushed salt blocks. 7,3.2 Bentonite Design The bentonite mixtures in the Salado could contain salt or sand as the filler material, A 50/50 mixture of bentonite and a filler with an in place density of about 1.8 g/cc will result in a low permeability seal which generates moderate swelling pressures. The minimum seal length should be 4 m. Emplacement of bentonite mixtures in block form offers good control over the in place properties. The discus- sion regarding shape and length of bentoni te-based seals from Section 7.2.1 applies to the bentonite compo- nent in the composite seal. For the bentonite mixtures at the top of the Salado and against anhydrite layers, confinement by crushed salt blocks rather than concrete is specified. The pores in the crushed salt blocks have been sufficiently small to prevent sub- stantial loss of bentonite in intermedi- ate size tests in the WIPP (Stormont and Howard, 1987). 7.3.3 Concrete Design Results from the Small-Scale Seal Performance Tests have been very favorable with regard to the establish- ment of tight, stable concrete seals in salt. Test Series A involved the placement of six concrete seals in vertically-down boreholes, and thereby simulated shaft seals in halite host rock (Stormont, 1986). Three different E-334 sizes were emplaced: 15.2 cm dia, 30.4 cm length; 40 cm dia, 61 cm length; 91 cm dia, 91 cm length. The concrete was the ESC mixture. Measurements of strains and stresses in the concrete seals and the adjacent rock revealed that the strains and stresses are com- pressive in nature, and are tending toward equilibrium. Creep of the adja- cent host rock was identified as the predominant mechanism for the develop- ment of stresses and strains in the con- crete (Stormont, 1987). The stability of the seal system was not threatened by permeability measurements, which imparted a 2 MPa axial gas pressure on one face of the seals, implying that the concrete/rock interface has substan- tial strength. Fluid flow measurements indicate that the seals are excellent barriers to fluid flow (Peterson, Lagus, and Lie, 1987b). Both brine and gas flow tests determined that five of the six seals had permeabilities of less than 10"'^ m^. There was no breakthrough of brine during a 140-day test at 3.5 MPa driving pressure on the 60 cm long seal (Peterson, Lagus, and Lie, 1987b). As part of the numerical analyses of shaft seals described in Section 7.2.2., Van Sambeek (1987) evaluated a 10 m long, 7 m diameter concrete shaft seal located in the top of the Salado. The seal was slightly recessed into the formation to account for the removal of the shaft key and any remnants of the chemical seal material that had been behind the key (Figure 7.8). The con- crete was modeled as the ESC, using the best available representations of the concrete properties. The temperature rise in the concrete was 72°C (Figure 7.9), and the maximum penetration of the 3 C temperature rise contour was about 5 m from the seal edge. The sub- sequent thermomechanical modeling of the seal system accounted for the time- dependent properties of the salt, as well as those of the concrete. The model results imply that the seal sys- tem is structurally stable. Consider the modeled radial stress at the rock/ Figure 7.8. Configuration of Modeled Concrete Seal in Top of Salado (from Van Sambeek, 1987). 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Time atler Seal EtnplacemenI (yr) Figure 7.9. Lift Temperatures in the ESC Salt Seal (from Van Sambeek, 1987). concrete interface given in Figure 7.10. The stress buildup within the first 0.2 year is a result of the E-335 i Z3 1 1 ■ 1 -1 V with Slld«Mn*t Y —— " Bonded ConlAcIt - a. Z -2 \\ M ec -3 ■ ^^^^^^'"'"''''""X - -4 - - -5 1 1 1 _ 0.01 0.1 1 10 100 Time aMer Seal Emplacement (yr) Figure 7.10. Contact Radial Stress in the ESC Salt Seal (from Van Sambeek, 1987). chemical expansion of the concrete and thermal stresses resulting from the heat liberated during hydration. The subsequent decrease in radial stress is due to the cooling of the concrete and the salt. The stress increase after about one year is a result of creep of the host rock; eventually, the radial stress would approach the lithostatic value of 6 MPa. Axial loads of 10 MPa to simulate the swelling of adjacent bentonite-based seals produced tensile stresses in the concrete that were largely attributed to the artificial modulus of the salt (one-twelfth of the measured value) used to improve creep closure calculations (Van Sambeek, 1987). Due to the dominant effects of salt creep, when the concrete was modeled without expansion the stresses in the seal system tended toward the same values as with expansion. Both the in situ tests and the nu- merical study indicate that the biggest advantage of concrete seals in salt is the tendency of the rock to creep in on the seal to effect a tight, stable interface that results in an early, positive seal without waiting for exten- sive salt creep. The shape previously given in Figure 7.6 is conceptual, to indicate that a shape other than a simple cylinder may be required. The length of approximately 10 m should be adequate. 7.3.4 Sealing the Salado Rock Placing seals in halite will in time reduce permeability in the sur- rounding formation and the interface. When a relatively stiff inclusion (such as concrete immediately after emplace- ment and crushed salt after it appre- ciably consolidates) is located in an opening in rock salt, the tendency of the rock to creep will cause the radial and tangential stresses in the vicinity of the inclusion to approach the litho- static stress. These stresses are expected to reverse the disturbance (including a decrease of permeability) in the adjacent rock by literally forc- ing the rock back together. Further, the stresses at the seal/rock interface are expected to become great enough to render the often-troublesome interface tight. Thus, emplacing certain seals may not only seal the excavation, but may also return the adjacent rock to a near pre-excavation condition. "Disturbance reversal" as described above has been observed in laboratory testing of halite, and has been re- ferred to as "healing." When samples of salt are brought from the field (in a disturbed condition), their permea- bilities are usually great. After application of hydrostatic pressure, permeabilities decrease to a low value and remain fairly insensitive to stress changes (Sutherland and Cave, 1979). Healing is generally attributed to de- creased porosity from plastic deforma- tion at the grain boundaries. Another type of healing that may occur in halite is macroscopic fracture healing. Limited tests of fracture toughness suggest that fractures in halite heal appreciably when subjected to moderate pressure and temperatures (Costin and Wawersik, 1980). IT Corporation (1987) E-336 found that confining pressure and ele- vated temperatures reduced the permea- bility of fractures in salt with time to a level comparable to that prior to fracturing. Healing has also been observed in in situ tests. Test Series B of the Small-Scale Seal Performance Tests involved 1-m-long horizontal concrete seals emplaced in 1-m-diameter bore- holes (Stormont and Howard, 1986). Ap- proximately 30 days after seal emplace- ment, tracer gas and flow measurements indicated that while the volumetric flow rates were quite small, very fast travel times ( 5 S u U 4 M i ' * u. * .? 1 -T ' 1 1 1 IMailmiim Mtatuitd Vilu* Minimum Mtatuitd Vilut T Test Interval Depth (m) Figure 8.5. Gas Flow Rates in Halite Test Intervals. (from Stormont, Peterson and Lagus, 1987), s o o w S m 6 S 1^ /::(s 4 • 1 — 1 . • McHurtd Vtlu* (Tttt Inltrval Includti MB13«) ▼ Matlmum Mtaturtd Valut (Tttt Inttrval Includti MB13S) A Minimum Mtaturtd Valut (TttI Inltrval Includti MB 139) i o u. « N 1 z o 3 2 1 -1 -2 o i : O Mtaiurtd Valut (TttI Inltrval Includti Stam B) V Mailmum Mtaiurtd Valua (Ttit Inltrval Includti Stam B) & Minimum Mtaiurtd Valut (Ttil Inltrval Includti Stam B) - E n 2 o ▼ 1 — J 1 7 OrIM Center Near Drift Edge or Removed From DrH< Figure 8.6. Flow Rates in Interbed Layers Within 2 m of WIPP Drifts When Tested at the Drift Center or Near or Just Removed from the Drift Edge (from Stormont, Peterson, and Lagus, 1987). o u 6 • Meatured Valut ' p — 1 ■■- 1 A — 1 5 - ▼ Mailmum Meatured Value A ^ Minimum Mtaiured Value • w 4 - i o u. 3 - ■o 2 1 • re f - b o 1 ■. -• z • ■ -1 T E ■c T T O) • ■ o -1 -' . , 1 2 4 6 8 10 12 Width of Drift (m) Figure 8.7. Drift-Width vs. Flow Rate From Tests on MB139 (from Stormont, Peterson, and Lagus, 1987). E-343 la Seam B ^ - Reading Below Resolution ol Equipment (less than 0.1 SCCM) s Figure 8.8. NllOO Drift Flow Rate (SCCM) Contours (from Boms and Stormont, 1987). Rock Bolts "^^^m^^^ \\\\V^^''''^^"''' K\\\\\\1.\ mmmm Figure 8.9. Idealized Excavation Effects in a 4m x 10m Room from (Borns and Stormont, 1987). E-344 damage. However, the limitations of overexcavation should be recognized: it will entail additional costs, there is no experience with or data from the overexcavated drift configuration, and as the effective diameter of the open- ing increases, the disturbed zone may propagate further into the rock. The overexcavation should be done just prior to seal emplacement if pos- sible, to minimize disturbance and uti- lize the increased creep rates of the host salt just after excavation. It is likely that portions of Marker Bed 139 and Seam B will have to be removed. The most favorable shape may be elliptical, rather than rectangular. Limited grout- ing of Marker Bed 139 in the vicinity of the overexcavation may be required to fill large voids. 8.3 Design Options Including Concrete Many alternative designs for panel seals can be generated by including concrete as a component. Concrete com- ponents are presently not thought nec- essary to establish effective panel seals, but they are retained as a secon- dary design option. Concrete could be used for many reasons: (I) if salt con- solidation assumptions are not substan- tiated; (2) for confinement of salt or bentonite-based seal components; (3) to reverse formation disturbance; (4) as a short-term flow barrier; (5) as a redun- dant component. Due to the length of time that they will be open, the seals in the main entries may require con- crete components. The shaft base seal (introduced in Chapter 7) is likely to include concrete. Test Series B of the Small-Scale Seal Performance Tests involved three 92 cm diameter, 92 cm long concrete (ESC) seals emplaced horizontally in the rib of Room M (Stormont and Howard, 1986). These seals simulate panel seals in an idealized (circular) geometry. Two seals contain thermal/structural instrumentation, and one is uninstru- mented. The concrete was pumped into place, and the resulting seals had an excellent contact with the host rock. Structural results indicate that the seals are stable, even when axially loaded to 2 MPa during flow testing (Labreche and Van Sambeek, 1987). Trends indicate that axial stresses may become tensile about two years after emplacement, perhaps eventually result- ing in fracture. As with Test Series A, the creep of adjacent rock was the predominant mechanism of stress and strain development in the concrete and the adjacent rock after transient effects diminished. The results of fluid flow tests of these seals indi- cated that they were excellent barriers to fluid flow and become more effective with time, due to the healing effect (Peterson, Lagus, and Lie, 1987b) dis- cussed in Section 7.3.4. Arguello and Torres (1987) conduct- ed analyses of concrete panel seal com- ponents. The model was identical to that previously used for their analyses of the crushed salt component (Section 8.2.1), except the crushed salt was replaced with a concrete seal. The con- crete was modeled as linearly elastic, and the material constants were those for the ESC (Gulick and Wakeley, 1987). The analyses were carried out to 50 years, where the seal system response is assumed to be dominated by salt creep. Previous modeling by Van Sambeek (1987) and Van Sambeek and Stormont (1987) suggests that the effects of hydration and expansion diminish with time for a concrete seal in salt be- cause of the dominant long-term effect of salt creep. When the concrete was loaded by the creep of the adjacent rock, the calculations predicted that essentially no tensile stresses devel- oped in the concrete, and the com- pressive stresses were well below the strength of the concrete. Tensile stresses which exist in the rock prior to seal emplacement (which indicate potential locations for fractures) were E-345 predicted to disappear and become com- pressive soon after seal emplacement; within five years after seal emplace- ment no tensile stresses exist. Thus, a concrete component is expected to generate a stress field in the adja- cent rock that is conducive to healing or tightening of the halite host rock. Results from Test Series B that substan- tiate this prediction are discussed in Section 7.3.4. A composite panel seal consisting of a central crushed salt core with con- crete end caps was analyzed by Arguello (1988). A two-dimensional, axisym- metric geomechanical model was used to estimate the effect of a finite length composite seal and the influence of the stiff concrete end caps on the consoli- dation of the central core (bridging). The concrete seals were 5.4 m in diam- eter and 5.3 m long, and the salt core was 5.4 m in diameter and 19.8 m long. The composite seal was assumed to be emplaced two years after excavation. Fractional densities of the crushed salt core as a function of time after emplacement are given in Figure 8.10 for an initial salt fractional density of 0.8. The effect of the concrete is largely confined to within 2 m of the concrete/crushed salt transition. Data from Test Series B imply that the clo- sure of a borehole is largely unaffect- ed by a concrete seal within one-half I g t9 Si; i •«■■•:.•:■•.■.••.••••••■•■•.••'■'■:■•»'■•.■'■•••■•••.'.'•:■•::•.■■• .•;r. Crushed Salt Core ■;V:'.':;/! ••"•:'■: 1.0 0.9 3 w c o c _o U « 0.7 T r T 1 r T 1 r Concrete Cap 1 r X Effective Salt Consolidation — Working Criterion Time After Seal Emplacement Syr 10yr 20 yr 30 yr 40 yr 50 yr Po ^ 0-80 0.6 L X X X X X 0.0 2.0 4.0 6.0 Distance from Seal Centeriine (m) 8.0 10.0 Figure 8.10. Fractional Densities of the Crushed Salt Core as a Function of Time After Emplacement (from Arguello, 1988). E-346 of a hole diameter away from the con- crete face (Labreche and Van Sambeek, 1987), and are therefore consistent with this modeling study. Arguello also predicted that unacceptably high axial tensile stresses develop in the concrete soon after seal emplacement, and suggested that reinforcement or seal geometries other than simple cylin- drical ones may be necessary. Test Series B data indicate that tensile strains develop in the concrete, but there is uncertainty over the stress measurements (Labreche and Van Sambeek, 1987). E-347 W ■:■■ 9. DESIGN EVALUATION OF NON-WASTE ROOM SEALS I < The non-waste rooms are to be sealed by backfilling with crushed salt. 9.1 Non-Waste Room Sealing Strategy Non-waste rooms or drifts are all excavations not presently dedicated to eventual waste disposal, including the experimental areas. It is presently planned to seal non-waste rooms by back- filling with crushed salt. The purpose of backfilling these areas is to pro- vide a redundant barrier to fluid migra- tion, limit the damage around these excavation, shorten the time until the repository is returned to a condition comparable to intact rock, and serve as a disposal location for mined salt. 9.2 Non-Waste Room Seal Design The non-waste rooms should simply be backfilled with crushed salt. No secondary excavation is anticipated. Pneumatic stowing or backfilling may be the emplacement method of choice because: (1) relatively high initial densities can be achieved; (2) there is good control over the consistency of the emplacement; (3) it is relatively inexpensive; (4) emplacement can be achieved remotely to avoid regions of possible danger. The time to achieve effective con- solidation can be inferred from anal- yses for consolidation of panel seal components (e.g., Arguello, 1988; Arguello and Torres, 1987). Based on these analyses, consolidation should be complete in less than 200 years. Be- cause of the time the excavations will be open prior to sealing and the lack of preparation or treatment of the adja- cent disturbed zone, substantial addi- tional time may be required to reverse the adjacent formation damage. S^i E-348 10. DESIGN EVALUATION OF BOREHOLE SEALS Cementitious grouts will be used to seal boreholes in the vicinity of the WIPP, probably without removing the hole casing. Present analyses suggest that crushed salt may not be effective for borehole sealing. 10.1 Borehole Sealing Strategy Previous assessments have indicated that open existing boreholes in the vicinity of the WIPP pose little or no threat to the public (Intera, 1981; Christensen, Gulick, and Lambert, 1981; Stormont, 1984), principally because no existing boreholes penetrate the WIPP Facility and salt must be dissolved in the boreholes before penetration could occur. Such dissolution is calculated to proceed slowly, and the requirements for borehole sealing are therefore ex- pected to be minimal. Because concerns regarding long-term performance are alleviated for borehole seals, cementi- tious mixtures can be used as the prin- cipal seal material. Cement-based materials (grouts) are preferred as borehole seal material for their emplacement characteristics. Bore- hole sealing entails remote emplace- ment, and confidence is required that the sealing material completely fills the borehole and makes good contact with the borehole wall. This may be particularly important in boreholes penetrating rock susceptible to substan- tial washouts (Christensen, Statler, and Peterson, 1980). Cement grouts have known flow properties and estab- lished emplacement techniques that allow good rock/seal contact. Even if the grout degrades into its constitu- ents (principally sand), adequate resis- tance to flow should exist (Stormont, 1984). 10.2 Borehole Seal Design The saltwater BCT- 1 F mix (Section 6.3.1) should be placed in the salt zones to preclude dissolution of the host rock by the cement water. On the other hand, the freshwater BCT- IFF mix is preferred in nonsalt zones because of its slightly better performance char- acteristics. A fundamental issue concerning bore- hole sealing is whether or not the cas- ing should be removed prior to sealing. Iron casing will corrode over long pe- riods of time, leaving a more permea- ble conduit through the seal (Tremper, 1966; Tonini and Dean, 1976). How- ever, because all boreholes which pene- trate the Salado are unlined below the Rustler contact with the exception of ERDA-9, a seal of substantial length which has a good bond with the host rock will be emplaced even if the hole is left cased. Very short borehole seals emplaced in salt in Test Series A of the Small-Scale Seal Performance Tests have exhibited permeabilities to gas and brine of less than 10''^ m^ (Peterson, Lagus, and Lie, 1987b). The Bell Canyon Test demonstrated the effec- tiveness of short grout borehole seals in anhydrite host rock (Christensen and Peterson, 1981). Given the probable minimum sealing requirements for bore- holes, it is believed that adequate seals can be achieved with the casing left in place above the Salado. 10.3 Design Options Including Crushed Salt While it would be desirable to achieve a salt seal in boreholes, pres- ent concerns regarding emplacement tech- niques and brine saturation prior to achieving high fractional densities do not make it a first choice material for borehole seals. Bridging of a granular material during remote emplacement in a relatively small diameter is possible. Concerns over bridging and the complete filling of the borehole can be partial- ly alleviated by first screening the salt to eliminate large grains, perhaps to a distribution similar to sand, and then emplacing it through tubing that E-349 is withdrawn during filling. Screening may also help to obtain the highest pos- sible initial density of the crushed salt. Even if the crushed salt could be emplaced effectively, the consolidation may be impeded by saturation of the crushed salt by brine from the host rock salt. Salt consolidation calcula- tions for shafts (Torres, 1987) can be related to salt consolidation in a borehole by applying the "pseudostrain concept" (Munson, Torres, and Jones, 1987), which in essence states that closure in homogeneous salt is directly proportional to its diameter. The frac- tional density of a crushed salt seal with time is therefore independent of the opening diameter, as long as the initial fractional density is the same. Thus, the results of Torres (1987) show- ing the change in fractional density with time for an initial fractional den- sity of 0.60 (see Figure 10.1) apply to salt consolidation in boreholes, as well as in shafts. The present best estimates of brine influx, however, pre- dict that once short-lived transients diminish, the volumetric flow rate is independent of excavation diameter, or stated differently, that the flux is 200 I •d § zi 300 100 m long seal of consolidated salt at the base of the shafts can be expected if exces- sive water from the overlying water- bearing zones is ruled out. Finite element modeling of salt consolidation in panel seal components shows that effective salt consolidation is ex- pected in less than 100 years at these locations. An alternative for emplac- ing salt seals is the use of quarried salt blocks. This technique, while not yet as advanced as crushed salt consoli- dation, could substantially reduce the time required to achieve an effective, long-term salt seal should that be desirable. The design evaluation reveals that the host rock is expected to signifi- cantly influence the adequacy of seal systems. The adjacent rock can be the predominant flow path through a seal system, as demonstrated by model stud- ies and in situ test results. In the shafts, rock in the Rustler Formation may be of particular concern due to its diversity and relative inaccessibility. At the disposal horizon, the anhydrite/ clay interbed has been observed to contribute substantially to a disturbed rock zone. In halite, the tendency for the host rock to creep has positive ben- efits. First, the closure consolidates the seal material. Once the seal mate- rial resists continued closure, the rock stresses increase and tend to tighten the seal/host rock interface. This effect, known as healing or distur- bance reversal, has been simulated nu- merically and observed during in situ tests. Bentonite-based seals have many favorable properties, including low per- meability and moderate swelling poten- tial. Model results suggest bentonite's importance in effecting adequate shaft seals in the Rustler. Bentonite can be tailor-mixed with other materials, such as sand or crushed salt, and emplaced in many different ways, including blocks. Initial results from in situ tests are favorable for bentonite/salt mixtures as barriers to fluid flow. Bentonite-based materials are being used in other experimental programs throughout the world, and the data base is therefore growing rapidly. The long- term physical and chemical stability of bentonite in WIPP environments is prom- ising, but largely unsubstantiated. Cementitious materials have been developed for placement in different WIPP environments (salt and nonsalt) with adequate material properties and emplacement characteristics. These ma- terials have been emplaced and tested in situ in numerous configurations, and have demonstrated exceptional sealing ability. Structural and thermal mea- surements have been used to improve numerical models of concrete/salt E-352 interaction. Numerical simulations of concrete seals in shafts and panel seal locations have indicated that stable seals should be achievable, but indica- tions of tensile stresses in concrete seals emplaced in halite have been noted. There is presently no known funda- mental obstacle to effectively sealing the WIPP by implementing the design con- cepts contained herein. Therefore, the present long-term sealing considera- tions support waste isolation at the WIPP. The designs, however, are not complete. Much work will be required to confirm these design concepts prior to the WIPP conversion from a pilot plant to a repository in about 1992. Key elements of the ongoing experi- mental program are given below. 11.1 Materials Development Laboratory testing of salt consoli- dation and permeability will continue, with emphasis on developing a mecha- nistic model of consolidation. The recently developed constitutive model will be tested against new laboratory data, and applied to potential seal configurations. The quarried salt concept will be evaluated to determine its feasibility. Emplacement technology, laboratory test- ing, and model simulations will be de- veloped if warranted. The long-term physical and chemical stability of bentonite-based seal mate- rials will be investigated. Basic properties and efficacy of asphalt seals will be obtained from existing literature. for formation grouting in the shafts will be pursued if necessary. 1 1.2 Formation Hydraulic Properties Measurements of permeability, pore pressure, and brine influx will be made in the soon-to-be-excavated Air Intake Shaft. Measurements to characterize the time-dependent development of the dis- turbed rock zone surrounding excava- tions at the facility horizon are being conducted. These measurements include gas permeability and dye injection tests. Further tests to determine brine influx size effects and the pressure regime in the vicinity of the WIPP Fa- cility are being implemented. A drift- scale, test is planned. 1 1.3 Seal Tests The Small-Scale Seal Performance Tests have provided a wealth of practi- cal information and data in return for a modest investment. The existing test series will be maintained to provide data on time-dependent effects, and future test series are being designed and implemented to simulate shaft seal components. A full-size test of a seal compo- nent will be required to provide rea- sonable assurance that the concepts developed on a relatively small-scale can be extrapolated to their intended application. Present plans are for a test of crushed salt block and quarried salt seals to be installed in conjunc- tion with a drift-size brine influx experiment. 1 1.4 Seal Design and Modeling Continued testing of laboratory sam- ples of cementitious material will con- tinue, with emphasis on indications of The models of crushed salt con- long-term stability. Grout development solidation in the Salado portion of the E-353 I i shafts and the flow through the seals in the Rustler portion of the shafts are being coupled to provide a more realistic model for the progressive consolidation and saturation of the crushed salt column in the shafts. Various loading conditions and geom- etries will be investigated to develop a stable concrete seal design for salt and non-salt host rocks. The necessity of concrete expansivity will be investi- gated. 11.5 System Integration An adequate and defensible design will require the integration of labora- tory data, in situ data, and modeling results. Results from other experi- mental programs must also be consid- ered, including performance assessment activities. A system analysis approach for the entire seal system and its sub- systems will be implemented to ensure that the design is adequate. E-354 12. REFERENCES Arguello, J. G. ( 1988). WIPP Panel Entrywav Seal - Numerical Simula- tion of Seal Composite Interaction for Preliminary Seal Design Evalua- tion . SAND87-2804, Sandia National Laboratories, Albuquerque, NM. Arguello, J. G. and T. M. Torres (1987). WIPP Panel Entrywav Seal - Numerical Simulation of Seal Compo- nent/Formation Interaction for Preliminary Seal Design Evaluation . SAND87-2591, Sandia National Lab- oratories, Albuquerque, NM. Auld, F. A. (1983). Design of Under- ground Plugs, International Journal of Mining Engineering . Vol 1. Baes, C. F. Jr., L. O. Gilpatrick, F. G. Kitts, H. R. Bronstein, and A. J. Shor (1983). The Effect of Water in Salt Repositories: Final Report . ORNL-5950, Oak Ridge Na- tional Laboratory, Oak Ridge, TN. Barr, G. E. , W. B. Miller, and D. D. Gonzalez (1983). Interim Report on the Modelling of the Regional Hydraulics of the Rustler Forma- tion . SAND83-0391, Sandia National Laboratories, Albuquerque, NM. Borns, D. J. and J. C. Stormont (1987). Delineation of the Disturbed Rock Zone Around Excavations in Rock Salt, Geological Society of America Abstracts vt^ith Programs . Vol 19, No 7. Buck, A. D. (1987). Effect of Salt- Saturated Mixing Water on Long-Term Durability of Hvdraulic-Cement Mix- tures , in preparation. Buck, A. D. (1985a). Development of a Sanded Nonsalt Expansive Grout for Repository Sealing Application . SL-85-6, Office of Nuclear Waste Isolation. Buck, A. D. (1985b). Development of Two Candidate Concrete Mixtures (Salt. Nonsalt) for Repository Seal - ing Applications . SL-85-8, Office of Nuclear Waste Isolation. Buck, A. D. , J. A. Boa, and D. M. Walley (1985). Development of a Sanded Expansive Salt Grout for Repository Sealing Application . SL-85-1I, Office of Nuclear Waste Isolation. Buck, A. D., J. E. Rhoderick, J. P. Burkes, K. Mather, R. E. Reinhold, and J. A. Boa (1983). Modifica- tion of Bell Canyon Test (BCT) l-FF Grout . SAND83-7097, Sandia National Laboratories, Albuquerque, NM. Burkes, J. P. and J. E. Rhoderick (198 3). Petrographic Examination of Bell Canyon Tests (BCT) l-FF Field Grouts over a Three-Year Pe- riod . SL-83-12, Waste Isolation. Office of Nuclear Bush, D. D. and R. Lingle, (1986). A Full-Scale Borehole Sealing Test in Anhydrite Under Simulated Down- hole Conditions. Vol. I . BMI/ONWI- 581(1), Office of Nuclear Waste Isolation. Bush, D. D. and S. Piele (1986). A Full-Scale Borehole Sealing Test in Salt Under Simulated Downhole Con- ditions. Vol. 1 . BMI/ONWI-573(l), the Office of Nuclear Waste Isola- tion. Callahan, G. D. (1981). Inelastic The r momec ha nical Analysis of a Generic Bedded Salt Repository . ONWI-125, Office of Nuclear Waste Isolation. E-355 Christensen, C. L., C. W. Gulick, and S. J. Lambert (1981). Sealing Con- cepts for the Waste Isolation Pilot Plant . SAND81-2195, Sandia National Laboratories, Albuquerque, NM. Franke, C. (1987). Excavation Effects and Fracture Mapping, input to 1987 GFDAR, Memorandum to R. McKinney, August 18, 1987, IT Corp, WIPP Site, Carlsbad, NM. I L9 S O t Christensen, C. L. and E. W. Peterson (1981). The Bell Canvon Test Sum- mary Resort . SAND80-1375, Sandia National Laboratories, Albuquerque, NM. Christensen, C. L., R. D. Statler, and E. W. Peterson (1980). Downhole Television Applications in Borehole Plugging . SAND80-0459, Sandia Na- tional Laboratories, Albuquerque, NM. Comes, G. D., L. D. Wakeley, and E. F. O'Neil (1987). Properties of Labo- ratory-Tested Specimens of Concrete from Small-Scale Seal Performance Test at the Waste Isolation Pilot Plant . SL-87-9, Sandia National Laboratories, Albuquerque, NM. Cos tin, L. S. and W. R. Wawersik (1980). Creep Healing of Fractures in Rock Salt . SAND80-0392, Sandia National Laboratories, Albuquerque, NM. D'Appolonia Consulting Engineers (1981). Sealing Consideration for Repository Shafts in Bedded and Dome Salt . ONWI-255, Office of Nu- clear Waste Isolation. Defense Nuclear Agency Concrete and Grout Technology for Underground Nuclear Effects Tests (n.d.). Department of the Army, U. S. Army Corps of Engineers (1984). Grout- ing Technology . Engineer Manual No. EM 1110-2-3506. Evans, B. (1983). Alteration of Po- rosity and Permeability by Plastic Flow Processes in Rocks, No. 27664, Geologic Society of America Ab- stracts with Programs . Vo I 15, No 6. Garrett, W. S. and L. T. Campbell Pitt (1961). Design and Construction of Underground Bulkheads and Water Barriers, III, Transactions of the Seventh Commonwealth Mining and Metallurgy Congress . South African Institute of Mining and Metallurgy. Garrett, W. S. and L. T. Campbell Pitt (1958). Tests on Experimental Bulk- heads for High Pressures, Journal of South African Institute of Min- ing and Metallurgy . 59:123-143. Gerstle, W. H. and A. K. Jones (1986). Mechanical Properties of Crushed Salt/Bentonite Blocks . SAND86-0707, Sandia National Laboratories, Albu- querque, NM. Gray, M. N., S. C. H. Cheung, and D. A. Dixon (1984). The Influence of Sand Content on Swelling Pressures and Structure Developed in Stati- cally Compacted Na - Bent oni t e . AECL-7825, Atomic Energy of Canada Limited, Pinawa, Manitoba. Gulick, C. W. (1987). Oral communica- tion to J. C. Stormont, September, Sandia National Laboratories, Albu- querque, NM. Gulick, C. W. and L. D. Wakeley, Refer - ence Properties of Cement-Based Plugging and Sealing Materials for the Waste Isolation Pilot Plant (WIPP) . SAND87-2817, Sandia Nation- al Laboratories, in preparation, Albuquerque, NM. Hart, D. E. (1983). Current Design and Maintenance Practice in Salt Mine Shafts, Proceedings of the Sixth International Symposium on Salt . E-356 Hang, A., V. A. Kelley, A. M. LaVenue, and J. F. Pickens (1986). Modeline of Ground-Water Flow in the Culebra Dolomite at the Waste Isolation Pi- lot Plant . SAND86-7167, Albuquer- que, NM. Kelsall, P. C, J. B. Case, D. Meyer, and W. E. Coons (1982). Schematic Designs for Penetration Seals for a Reference Repository in Bedded Sail, ONWI-405, Office of Nuclear Waste Isolation. Heimann, R. B., M. A. T. Stanchell, and R. D. Hooton, (1986). Short-Term Dissolution Experiments on Various Cement Formulations in Standard Canadian Shield Saline Solution in the Presence of Clav . AECL-9059, Atomic Energy of Canada Limited. Herubin, C. A. and T. W. Marotta (197 7). Basic Construction Mate- rials . Res ton Publishing Co., Reston, VA. Kimbrell, A. F., T. S. Avery, and J. J. K. Daemen (1987). Field Testing of Bentonite and Cement Borehole Plugs in Granite . NUREG/CR-4919, US Nu- clear Regulatory Commission. Krieg, R. D. (1984). Reference Strati- graphy and Rock Properties for the Waste Isolation Pilot Plant fWIPP) Project . SAND83-1908, Sandia Na- tional Laboratories, Albuquerque, NM. Holcomb, D. J. and M. Shields (1987). Hydrostatic Creep Consolidation of Crushed Salt with Added Water . SAND87-1990, Sandia National Lab- oratories, Albuquerque, NM. Holcomb, D. J. and D. W. Hannum (1982). Consolidation of Crushed Salt Under Conditions Appropriate to the WIPP Facility . SAND82-0630, Sandia Na- tional Laboratories, Albuquerque, NM. Hunter, R. L. (1987). Memorandum to Distribution . July 10, Sandia Na- tional Laboratories, Albuquerque, NM. IT Corp. (1987). Laboratory Investiga- tion of Crushed Salt Consolidation and Fracture Healing . BMI/ONWI-631 , Office of Nuclear Waste Isolation. Intera Environmental Consultants, Inc. (•981). Consequence Assessment of Hydrological Communications throug h Borehole Plugs, SAND8I-7I64, Sandia National Laboratories, Albuquerque, NM. Johnson, A. I., R. K. Froebel, N. J. Cavalli, and C. B. Pettersson, Edi- tors (1984). Hydraulic Barriers in Soil and Rock, ASTM Special Techni- cal Publication 874. KrumhansI, J. L. (1984). Observations Regarding the Stability of Benton- ite Backfill in a High-Level Waste ( HLW) Repository in Rock Salt . SAND83-I293, Sandia National Labora- tories, Albuquerque, NM. Labreche, D. A. and L. L. Van Sambeek (1987). Analysis of Data from Ex- pansive Salt-Water Concrete Seals in Series B Small-Scale Seal Perfor - mance Tests . SAND87-7155, Sandia National Laboratories, Albuquerque, NM. Lama, R. D. and V. S. Vutukuri (1978). Handbook on Mechanical Properties of Rocks: Volume II. Trans Tek . Clausthal, Germany. Lambert, S. J. (1980a). Recommended Analysis Plan for the Borehole Plugging Program Potash Core Test . SAND80-08I2, Sandia National Laboratories, Albuquerque, NM. Lambert, S. J. (1980b). A Strategy for Evaluating the Long-Term Stability of Hole-Plugging Materials in Their Geologic Environments . SAND8 0- 0359C, Sandia National Laborato- ries, Albuquerque, NM. E-357 Lankard, D. R. and F. L. Burns (1981). Mercer, J. W. and B. R. Orr (1979), Borehole Plugging and Shaft/Tunnel Sealing With Portland Cement-Based Materials As Related to the Geolog- ic Isolation of Radioactive Waste , presented at the American Ceramic Society 83 Annual Meeting. Interim Data Report on the Geohv- drology of the Proposed Waste Iso- lation Pilot Plant Site. Southeast New Mexico . US Geologic Survey Wa- ter Resources Investigations Re- port, 79-98. Lea, F. M. (1971). The Chemistry of Cement and Concrete . Chemical Pub- lishing Co. Lee, C. F. (1985). A Case History on Long-Term Effectiveness of Clay Sealant . TR-338, Atomic Energy of Canada Limited. Meyer, D. and J. J. Howard, Editors (1983). Evaluation of Clays and Clay Minerals for Application to Repository Sealing . ONWI-486, Of- fice of Nuclear Waste Isolation. Monastersky, R. (1987). Heirs to cient Air, Science News . Vol. 132. An I s Leppert, D. (1986). Environmental Uses of Natural Zeolites and Bentonite, Proceedings of the 22" Symposium on Engineering Geology and Soils Engineering . Boise, ID. Lopez, R. S., Editor (1987). The Vault Sealing Program: Proceedings of the Workshop Held 1983 December 1 - 2 in Toronto. Ontario . TR-339, Atomic Energy of Canada Limited. Malinowski, R. (1981). Ancient Mortars and Concretes - Durability Aspects . Goteberg, Sweden. Mercer, J. W. (1983). Geohvdrology of the Proposed Waste Isolation Pi- lot Plant Site. Los Medanos Area. Southeastern New Mexico . US Geo- logical Survey Wa ter Re sources Investigation Report 83-4016, Mercer, J. W. (1986). Compilation of Hydrologic Data from Drilling the Salado and Castile Formations near the WIPP Site. Southeast New Mexi- co., SAND86-0954, Sandia National Laboratories, Albuquerque, NM. Munson, D. E., T. M. Torres, and R. L. Jones (1987). Pseudostrain Repre- sentation of multipass excavation.s in salt. Proceedings of the 28 U.S. Rock Mechanics Symposium . th National Coal Board (1982). The Treat- ment of Disused Mine Shafts and Adits . Mining Department, National Coal Board. Nowak, E. J. and D. F. McTigue (1987). Interim Results of Brine Trans- port Studies in the Waste Isolation Pilot Plant (WIPP) . SAND87-0880, Sandia National Laboratories, Albu- querque, NM. Nowak, E. J. and J. C. Stormont (1987). Scoping Model Calculations of the Reconsolidat ion of Crushed Salt in the WIPP Shafts . SAND87-0879, Sandia National Laboratories, Albu- querque, NM. Nowak, E. J. (1987). Oral communication to J. C. Stormont, November, Sandia National Laboratories, Albuquerque, NM. E-358 Penman, A. D. M. and J. A. Charles (1979). The Influence of Their In- terfaces on the Behaviour of Clay Cores in Embankment Dams, Inter- national Congress on Large Dams. 1 3^ '^. Tr ansae tions . Vol. 1^ New Delhi, India. Peterson, E. W., P. L. Lagus, and K. Lie (1987a). WIPP Horizon Free Field Fluid Transport Characteris- tics . SAND87-7164, Sandia National Laboratories, Albuquerque, NM. Peterson, E. W., Lie (1987b). ments of Test P. L. Lagus, and K. Fluid Flow Measure- Series A and B of the Small Scale Seal Performance Tests . SAND87-704I, Sandia National Labo- ratories, Albuquerque, NM. Peterson, E. and S, Laboratory Tests Kelkar (1983) to Determine Hy- draulic and Thermal Properties of Bentonite-Based Backfill Materials . SAND82-7221, Sandia National Labo- ratories, Albuquerque, NM. Peterson, E. W., Broce, and K. P. L. Lagus, R. D. Lie (1979). In Situ Permeability Testing of Rock Salt . SSS-R-79-4084-2, Sandia National Laboratories, Albuquerque, NM. Pfeifle, T. W. (1987). Mechanical Properties and Consolidation of Potential DHLW Backfill Materials: Crushed Salt and 70/30 Bentonite Sand . SAND85-7208, Sandia National Laboratories, Albuquerque, NM. Pfeifle, T. W. and P. E. Senseny (1985). Permeability and Consoli- dation of Crushed Salt From The WIPP Site . Topical Report RSI-0278, Sandia National Laboratories, Albu- querque, NM. Popielak, R. S., R. L. Beauheim, S. R. Black, W. E. Koons, C. T. Ellington, and R. L. Olsen (1983). Brine Reservoirs in the Castile Formations. Southeastern New Mexico . TME 3153. Pusch, R. , L. Borgesson, and G. Ramquist, (1987). Final Report of the Borehole. Shaft, and Tunnel Sealing Test - Volume II: Shaft Plugging . Stripa Project Report 87-02. Pusch, R. (1983). Stability of Ben tonite Gels i n Crystalline Rock Physical Aspects . Report 83-04. Stripa Project Pusch, R. (1980). Swelling Pressure of Highly Compacted Bentonite . KBS80- 13, Stockholm. Pusch, R. (1980). Workshop on Sealing Techniques. Tested in the Stripa Project and Being of General Poten- tial Use for Rock Sealing . Stripa Project Report 87-05. Radhakrishna, H. S. and H. T. Chan (1985). Strength and Hydraulic Con- ductivity of Clay-Based Buffers for a Deep Underground Nuclear Fuel Waste Disposal Vault . TR-327, Atom- ic Energy of Canada Limited. Resendiz, D. (1976). Relevance of Atterbere Limits in Evaluating Pi ing and Breaching Potential . 79 fn Annual Meeting of ASTM, Dispersive Clays, Related Piping and Erosion in Geotechnical Projects, ASTM-STP- 632. Roy, D. M., M. W. Grutzeck, and L. D. Wakeley (1983). Selection and Dura- bility of Seal Materials for a Bed- ded Salt Repository: Preliminary Studies . ONWI-479, Office of Nucle- ar Waste Isolation. Saulnier, Jr., G. J. and J. D. Avis (1988). Interpretation of Hydrau- lic Tests Conducted in the Waste- Handling Shaft at the Waste Isolation Pilot Plant (WIPP) Site . Sandia National Laboratories, Albu- querque, NM in preparation. E-359 I s S :3 P I; Schaffer, A. and J. J. K. Daemen (1987). Experimental Assessment of the Sealing Effectiveness of Rock Fracture Grouting . NUREG/CR-4541, US Nuclear Regulatory Commission. Shor, A. J., C. F. Baes, and C. M. Canonico (1981). Consolidation and Permeability of Salt in Brine . ORNL-5774, Oak Ridge National Laboratory, Oak Ridge, TN. Sima, N. and A. Harsulescu (1979). The Use of Bentonite for Sealing Earth Dams . International Association of Engineering Geology, Bulletin 20. Sitz, P. (1981). Cross-Sectional Seals of Underground Cavitites Involving Dams and Plugs . Translated by Lan- guage Services for Sandia National Laboratories, Albuquerque, NM, 1984. Sjaardema, G. D. and R. D. Kr i e g (1987). A Constitutive Model for the Consolidation of WIPP Crushed Salt and Its Use in Analvses of Backfilled Shaft and Drift Con- figurations . SAND87-1977, Sandia National Laboratories, Albuquerque, NM. South, D. L. and J, J. K. Daemen (1986). Permeameter Studies of Water Flow Through Cement and Clav Borehole Seals in Granite. Basalt and Tuff . NUREG/CR-4748, Nuclear Regulatory Commission. South, D. L. (1979). Well Cementing , topical report prepared for US Nu- clear Regulatory Commission. Stormont, J. C. (1987). Small-Scale Seal Performance Test Series "A" Thermal/Structural Data Through the Stormont, J. C. and J. G. Arguello (n.d.). Model Calculations of Flow through Shaft Seals in the Rustler Formation . SAND87-2859, Sandia Na- tional Laboratories, Albuquerque, NM, in preparation. Stormont, J. C. and C. L. Howard (1987). Development. Implementa- tion, and Early Results: Test Series C of the Small Scale Seal Performance Tests . SAND87-2203, Sandia National Laboratories, Albuquerque, NM. Stormont, J. C, E. W. Peterson, and P. L. Lagus (1987). Summary of and Observations about WIPP Facility Horizon Flow Measurements through 1986 . SAND87-0I76, Sandia National Laboratories, Albuquerque, NM. Stormont, J. C. and C. L. Howard (1986). Development and Implementa- tion: Test Series B of the Small- Scale Seal Performance Tests . SAND86-1329, Sandia National Labo- ratories, Albuquerque, NM. Stormont, J. C, ed. (1986). Develop- ment and Implementation: Test Series A of the Small-Scale Seal Performance Tests . SAND85-2602, Sandia National Laboratories, Albu- querque, NM. Stormont, J. C. (1984). Plugging and Sealing Program for the Waste Iso- lation Pilot Plant (WIPP) . SAND84- 1057, Sandia National Albuquerque, NM. Laboratories, Stormont, J. C. (1983). Mechanical Strength of Borehole Plugs . Thesis, University of Arizona. 180 TTT Day . SAND87-0178, Sandia Na- tional Laboratories, Albuquerque, NM. E-360 S t r oup, D. (1987). Content E. a nd P. E. Se ns e ny, Influence of Bentonite on Consolidation and Per- meability of Crushed Salt from the WIPP . RSI-0309, Sandia National Laboratories, Albuquerque, NM. Sutherland, H. J. and S. P. Cave (1980). Argon Gas Permeability of New Mexico Rock Salt Under Hydro- static Compression, International Journal of Rock Mechanics. Mining Sciences, and Geomechanics Ab- stracts . Vol 17. Tonini and Dean, eds (1976). Chloride Corrosion of Steel in Concrete . ASTM Publication STP 629. Torres, T. M, (1987). Design Evalua- tion: Structural Calculations for the Construction and Salt Handling Shaft and the Waste Handling Shaft at the Waste Isolation Pilot Plant (WIPP) . SAND87-2230, Sandia Nation- al Laboratories, Albuquerque, NM. Tremper, B. (1966). Corrosion of Rein- forcing Steel. Significance of Tests and Properties of Concrete and Concrete-Making Materials . ASTM Special Technical Publication Num- ber 169-A. US DOE (1986). Design Validation Final Report . DOE-WIPP-86-010. US DOE (1987). Exhaust Shaft Grouting Program "C" . Report for Week Ending June 21, WIPP, Carlsbad, NM. US DOE (1984). Geotechnical Activities in the Waste Handling Shaft . WSTD- TME-038. US DOE (1983). Geotechnical Activities in the Exploratory Shaft . TME 3178. US DOE and State of New Mexico (1981). Agreement for Consultation and Cooperation on WIPP by the State of New Mexico and US DOE, Modified 11/30/84 and 8/4/87, 1981. US Environmental Protection Agency (1985). "Environmental Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, Final Rule," 40CFR191. Federal Reg- ister . Vol 50, p. 38065, September, 1985. Van Sambeek, L. L. and J. C. Stormont (1987). Thermal/Structural Modeling of Test Series A of the Small Scale Seal Performance Tests . SANDS 7- 7037, Sandia National Laboratories, Albuquerque, NM. Van Sambeek, L. L. (1987). Thermal and Thermomechanical Analyses of WIPP Shaft Seals . SAND87-7039, Sandia Na- tional Laboratories, Albuquerque, NM. Wakeley, L. D. and D. M. Walley (1986). Development and Field Placement of an Expansive Sa 1 1 -Sat ur a ted Concrete (ESC) for the Waste Isola- tion Pilot Plant (WIPP) . SL-86-36, Sandia National Laboratories, Albu- querque, NM. Wakeley, L. D. (1987a). Dependence of Expansion of a Salt-Saturated Con- crete on Temperature and Mixing and Handling Procedures . SL-87-20, Sandia National Laboratories, Albu- querque, NM. Wakeley, L. D. (1987b). Durability of a Chloride-Saturated Concrete for Sealing Radioactive Wastes in Bed- ded Rock Salt . American Concrete Institute SP-1011. Wakeley, L. D. Nature of and D. M. Roy (1986). the Interfacial Region Between Cementitious Mixtures and R ocks for the Palo Duro Basin and Other Seal Components . BMI/ONWI- 580, Office of Nuclear Waste Isolation. E-361 Wakeley, L. D., D. M. Walley and A. D. Buck (1986). Development of Fresh- Water Grout Subsequent to the Bell Canyon Tests (BCT) . SL-86-2, Sandia National Laboratories, Albuquerque, NM. Wakeley, L. D. and D. M. Roy (1985). Cementitious Mixtures for Sealing Evaporite and Clastic Rocks in a Radioactive- Waste Repository . SL-85-16, Office of Nuclear Waste Isolation. Wegener, W. (1983). Sinking of a New Shaft and Watertight Lining of an Old Shaft at the Heilbronn Rock Salt Mine, Proceedings of the Sixth International Symposium on Salt . Vol. 1. Yost, F. G. and E. A. Aronson (1987). Crushed Salt Consolidation Kinet- ics . SAND87-0264, Sandia National Laboratories, Albuquerque, NM. Zeuch, D. (1987). Personal communica- tion to J. C. Stormont, October, Sandia National Laboratories, Albu- querque, NM. ZoBell, C. E. and M. A. Molecke (1978). Survey of Microbial Degradation of Asphalts With Notes on Relation- ship to Nuclear Waste Management . SAND78-1371, Sandia National Labora- tories, Albuquerque, NM. i I X < Wheelwright, E. J., F. N. Hodges, L. A. Bray, J. H. Westsik, D. H. Lester, T. L. Nakai, M. E. Spaeth, and R. T. Stula (1981). Development of Backfill Material as an Engineered Barrier in the Waste Package System - Interim Topical Report . PNL-3873, Pacific Northwest Laboratory. ■1- t E-362 REFERENCES FOR APPENDIX E Boms, D.J., and J.C. Stormont, 1988. An Interim Report on Excavation Effect Studies at the Waste Isolation Pilot Plant: The Delineation of the Disturbed Rock Zone . SAND87-1375, Sandia National Laboratories, Albuquerque, New Mexico. Deal, D. E., and J. B. Case, 1987. IT Corporation, Brine Sannplinq and Evaluation Program. Phase I Report . DOE-WIPP-87-008, prepared by the Engineering and Technology Department of the Management and Operating Contractor, Waste Isolation Pilot Plant Project, for the U.S. Department of Energy. Nowak et al. (E. J. Nowak, D. F. McTigue, and R. Beraun), 1988. Brine Inflow to WIPP Disposal Rooms: Data. Modeling and Assessment . SAND88-0112, Sandia National Laboratories, Albuquerque, New Mexico. Peterson et al. (E. W. Peterson, P. L. Lagus, and K. Lie), 1987. WIPP Horizon Free Field Fluid Transport Characteristics . SAND87-71 64, Sandia National Laboratories, Albuquerque, New Mexico. Peterson et al. (E. W. Peterson, P. L Lagus, J. Brown, and K. Lie), 1985. WIPP Horizon In Situ Permeability Measurements . Final Laboratories, Albuquerque, New Mexico Stormont et al. (J. C. Stormont, E. W. Peterson and P. L. Lagus), 1987. Summary of and Observations about WIPP Facility Horizon Flow Measurements Through 1986 . SAND87-0176, Sandia National Laboratories, Albuquerque, New Mexico. In Situ Permeability Measurements . Final Report, SAND85-7166, Sandia National j Saulnier, G. J., and J. D. Avis, 1988. Interpretation of Hydraulic Tests Conducted in the Waste-Handling Shaft at the Waste Isolation Pilot Plant (WIPP) Site . SAND88- 7001, prepared for Sandia National Laboratories by INTERA Technologies, Inc. Stormont, J. C, 1988. Preliminary Seal Design Evaluation for the Waste Isolation Pilot | Plant . SAND 87-3083, Sandia National Laboratories, Albuquerque, New Mexico. j E-363\364 S < s :3 t9 ^^V• -V-. APPENDIX F RADIOLOGICAL RELEASE AND DOSE MODELING FOR PERMANENT DISPOSAL OPERATIONS F-i/ < s u ::3 O y, TABLE OF CONTENTS Section Page F.1 INTRODUCTION F-1 F.1.1 Overview of AIRDOS-EPA F-1 F.1.2 Meteorological Modeling F-1 F.1.3 Stack Effluent Modeling F-2 F.1.4 Dispersion Modeling F-2 F.1.5 Terrestrial Modeling F-2 F.1.6 Dose Modeling F-3 F.2 PLUTONIUM-EQUIVALENT CURIE F-1 7 F.3 DESCRIPTIONS OF ACCIDENT SCENARIOS ANALYZED IN THE SEIS F-19 F.3.1 Accidents involving CH waste F-19 F.3.2 Accidents involving RH waste F-25 F.3.3 Accidents involving flammable or detonable gases F-27 REFERENCES FOR APPENDIX F F-32 F-iii m c u a. % < X o i LIST OF TABLES Table Page F.1 Meteorological data: assessment of routine releases F-5 F.2 Frequency of atmospheric stability classes for each direction F-6 F.3 Frequencies of wind directions and true-average wind speeds F-7 F.4 Frequencies of wind directions and reciprocal-average wind speeds . . F-8 F.5 Stack information F-9 F.6 Terrestrial modeling assumptions F-10 F.7 Bioaccumulation factors F-12 F.8 Dose receptor assumptions F-1 3 F.9 Dose rate conversion factors F-1 4 j F.10 Organ dose correction factors (unitless) F-1 5 F.11 Radionuclide specific parameters 50-year committed dose factors .... F-1 6 F.1 2 PE-Ci weighting factors for selected radionuclides F-1 8 ! F.1 3 Catastrophic hoist accident F-25 F-iv F.1 INTRODUCTION This appendix provides information concerning the radiological dose assessment modeling used to evaluate the risks associated with WIPP operations. A discussion of the AIRDOS-EPA computer model and Its input parameters is provided, the concept of Plutonium equivalent curies is explained, and descriptions of accident scenarios are presented. In response to numerous comments on the draft SEIS accident analysis, variations to the accident scenarios have been postulated in F.3 to consider alternate assumptions which result in more severe but less likely consequences. The credible accident scenario having the highest projected consequences is that of a postulated drum fire in the underground waste disposal area. F.1.1 OVERVIEW OF AIRDOS-EPA AIRDOS-EPA (Moore et al., 1979) estimates the radiation dose to either a maximally exposed individual or to an exposed population from the release of a specified quantity of radionuclides to the atmosphere. The code estimates concentrations of radioactivity in air, deposition buildup on ground surface, and ground surface concentrations based on release information, characteristics of the area surrounding the release site (e.g., agricultural productivity and land use), and specified meteorological conditions. These estimates, combined with intake rates for man, were used to estimate the radiation dose to an exposed adult human from potential exposure pathways for routine and accidental releases. F.1.2 METEOROLOGICAL MODELING The WIPP site area was modeled as a 50-mile-radius circular grid system with the site located at the center. Site-specific meteorological data, typical of annual average conditions, were specified for the assessment of routine annual releases. The annual frequency of wind direction was first determined for each of the 1 6 principal compass directions. The frequency of each Pasquiil stability category, ranging from category A (very unstable) to category G (extremely stable), was then determined for each of the 16 directions. The average wind speed was entered for each wind direction and Pasquiil category. The average depth of the atmospheric mixing layer (lid) for the area was specified to limit the vertical dispersion of the plume after it travels some distance downwind of the source. The lid value used applies to routine and accidental releases. For the assessment of accidental releases from the WIPP, stable meteorological conditions that allow minimal dispersion were assumed: a wind speed of 2 m/s under stability class F (very stable) conditions with wind direction constrained to a single direction for the maximum individual and annual average conditions with wind direction constrained to the direction having the highest consequences for the general population. F-1 F.1.3 STACK EFFLUENT MODELING The waste handling building stack and/or the exhaust shaft are the two possible release points for routine and accidental releases (release points are referred to as "stacks" for modeling purposes). AIRDOS-EPA requires input describing each area or point of release. Because the air will be discharged from the "stacks" at a relatively high velocity, the release will effectively take place at a height above the physical stack. Models for momentum-dominated plumes (Rupp et al., 1948) were used to estimate effective stack heights for releases associated with routine operations and projected accidents. This method employed an effective "stack velocity" in the vertical direction to determine the effective height of the release since the discharge from the stack will be angled. The effective point of release was also offset to account for the angled discharge. For releases associated with postulated accidents, the effective stack heights were estimated using Rupp's equation and reflected actual stack velocity measured during the postulated accidental release. F.I. 4 DISPERSION MODELING t I I The Gaussian plume model of Pasquill (1961), as modified by Gifford (1961), estimates plume dispersion in the downwind direction. The values recommended by Briggs (1 969) for the horizontal and vertical dispersion coefficients were used for dispersion and depletion calculations. The code permits consideration of dry deposition and scavenging for determining deposition of radionuclides on ground surfaces. Dry deposition is the process by which particles are deposited on grass, leaves, and other surfaces by impingement, electrostatic deposition, chemical reactions, or chemical reactions with surface components. The rate of deposition on earth surfaces is proportional to the ground-level concentrations of the radionuclides in the air (Slade, 1968). Scavenging is primarily due to washout of particles from a plume by rain or snow and is, therefore, a function of the precipitation rate. The scavenging coefficient was averaged over an entire year, including periods during which rain or snow would not fall. Scavenging can thus be described as a continuous removal of a fraction of the plume per second over the entire year. The value for the total ground deposition rate used in assessing routine releases was the sum of the dry deposition and the scavenging rates. The code removes the deposited fraction and maintains a mass balance along the plume as the concentration of the plume decreases. For the accidental release assessment, scavenging due to precipitation was conservatively ignored. F.I. 5 TERRESTRIAL MODELING As previously stated, the area surrounding the WIPP site was modeled as a 50-mile radius circular grid system with the WIPP facilities located at the center. Within the grid, 20 distances were specified in each of the 16 compass directions. Each distance F-2 represented the midpoint of a sector. Eleven distances were specified within a 5-mile radius of the WIPP. The remaining nine distances were specified at about 5-mile incremental distances from the center of the site. Within each sector formed by the grid system, WIPP-specific data were used for population, agricultural area, surface- water area, and numbers of beef and dairy cattle. These data are summarized in Section 2.1 of the draft Final Safety Analysis Report (DOE, 1989). Other factors used in modeling terrestrial and food crop transport are provided in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.109 (NRG, 1977). One-half of the anticipated operational life of the facility, 12.5 years, was specified as the period of time allowed for long-term buildup of radioactivity on surface soils. F.I. 6 DOSE MODELING The AIRDOS-EPA computer model estimates radiological intake rates at specified environmental locations. Resultant doses are then calculated through various exposure modes, using the ground-level concentrations in air and ground deposition rates computed from the meteorological input. To estimate the collective population dose, average values in the crosswind direction over each sector were used for the air concentrations and ground deposition rates. The average individual dose was determined by dividing the population dose by the number of individuals in the exposed population. The dose to an individual receiving a maximum dose (maximally exposed) was determined directly by the code. For accident assessments, it was assumed that the maximally-exposed individual was located on the center line of the discharge plume at the point of highest off-site ground- level concentration for the entire duration of the accident. The population dose for accident assessments was calculated using annual average meteorological conditions (wind speed and stability class frequency distribution) with a constant wind in the direction which maximizes the collective population doses. Exposure pathways, primarily the air pathway, are discussed in Subsection 5.2.3.2. The model calculates doses to total body, lungs, red bone marrow, lower large intestine wall, stomach wall, kidneys, liver, endosteal cells, thyroid, testes, and ovaries. The doses calculated are 50-year Committed Effective Dose Equivalents (GEDE) resulting from a one-year exposure for routine releases or one-time exposure for accidental releases. The internal dose conversion factors used in the calculation were those reported in Dunning (1986). The inhalation factors were based on the IGRP Task Group Lung Model, which simulates the behavior of particulate matter in the respiratory tract. The inhalation factors used correspond to a median aerodynamic diameter of 1 micron. The ingestion factors were based on a four-segment catenary model with exponential transfer of radioactivity from one segment to the next. Retention of nuclides in other organs was represented by linear combinations of decaying exponential functions. In the inhalation and ingestion models, cross-irradiation (irradiation of one organ by nuclides contained in another) is included. F-3 c u I X i ''^ r, The Dunning dose factors are based on the ICRP and NCRP models endorsed by the DOE in its August 5, 1985, Vaughan memorandum (DOE, 1985). Further, Dunning calculated dose factor using the same organ uptake fractions for daughter products as for the parent, as recommended in more recent ICRP guidance. Comparison of the Dunning dose factors with those recommended by the Vaughan memorandum (DOE, 1985) indicates that Dunning's approach is slightly more conservative. External dose rate conversion factors developed by Kocher (1981) are used. Where the chemical form and solubility of nuclides in the source term was not known, the solubility class which yielded the highest effective dose commitment was used in the model. For the alpha emitters, a quality factor of 20 was used in the calculation as recommended in ICRP Publication 26 (ICRP, 1977). Input parameters to the AIRDOS-EPA model specific to the WIPP site are documented in Tables F.I through F.11. F-4 TABLE F.1 Meteorological data: assessment of routine releases Parameter Value (units) Lid height Average temperature Average rainfall Frequency of atmospheric stability classes for each direction Frequencies of wind directions and true-average wind speeds Frequencies of wind directions and reciprocal-average wind speeds Pasquill Category Temperature Gradients^ E F G 1 .435 (m) 288.8 (°K) 24.13 (cm/yr) Table F-2 Table F-3 Table F-4 0.0055 (°K/m) 0.0280 (°K/m) 0.0400 (°K/m) Categories A-D are not utilized in the AIRDOS-EPA Code; Categories E-G are AIRDOS- | EPA Code default values. 1 F-5 (3 •i-oor».i--r-r^^i-o>'^cD-r-cDcoio CM"^"^COOC>J^flOC\l"^lOOOO-r^O)^CO^lOCVlT-^i-lOCOOOpjN-tD-r7tOCp h- <3) If) l0Oinh«.T-C0 ""^r^COCDT-OJCMCOOCDOCMCDlOCOlO r^COOCOCOCDCOCJ)h~^CVJ^^COh-lO ir)coc\j-»-cvjcjc\ic\ico^iO' o> CD lO 00 in in in CO CO (O CO in O 0) >. (S c ■D •g o ._ (0 r: o (0 E _ CO © o T3 c m c 3 5 O DC E CD > a> ti LU ■ (D DC c o CO o X CO I o o X CO o X o o X o CO X o o X o o X X CM c\i o o •^ X \n 00 X X Oi lO a> 00 o X X 00 IT) lO 00 X 00 00 o X lO 00 o X CO CO O X CO 00 C\j o X o CT> b X CM IT) b X CO to I o ^ X o CO b X X y- CM I O ^ X CM X o OJ X CM CO M b X (6 I o X CM fO I o X CM in b X CM X CM eo' X CM 00 X •<* CO 'b X CM iri I o X CO m b X X ▼- CM b b X CM CO I o X O) CM b X oi fo o X o> (O' X o CM m b ^-- X CO CO X b X CM X CO CO b X CO in b X 00 CO c^ o X in X CO CM I o X CM X o 00 CO I o X CM cvi X CM CO n b X o 00 b X CM CO b CO b X 00 CO o 3 DC X CM in CM CO X CM CO CO V) O X X CO 1- b CO b X o X X (O in CO o 1- O) % ay CM b X in X CM o X CO o X 00 CO J d) O X CM in en CO CM 3 Q. CO <3> D) c 'c c Q CO c g> ■>» (D CO o ■D c (0 0) c o o £ N 'co (0 O CO (D c ■c S (0 ^ (1) Q. t— CO -J o f?r c> CO o Q> u (0 o o c E ■D o o CD > to a. 7S , 0) E ^ CO o c c o CD o 3^ CO (0 0) 0) x: n o c c sz 3 0) D) o D w c (0 — T3 (1) •5 2 0) CO (D (/) o O (0 o o Q. (0 >- T5 O <0 •*— ^ 0) c CO 0) ^ D 50 ppm continuously for 12 weeks (Nomiyama et al., 1986). Other unspecified treatment-related hepatic effects were also noted by Nomiyama et al. (1986). Renal effects include increased kidney weights and dysfunction in rats exposed to >150 ppm for 12 weeks (Nomiyama et al., 1986) and renal tubular meganucleocytosis in rats exposed to >300 ppm for 7 hr/day, 5 days/week for 104 weeks (Maltoni et al., 1986). Hematological system alterations have been experienced by experimental animals subjected to trichloroethylene. Rats treated for 10 days with >50 ppm had concentration-dependent inhibition of delta-aminolevulinate (ALA) dehydrogenase activity in the liver and bone marrow cells, increased ALA synthetase, decreased heme saturation of tryptophan pyrrolase, and decreased cytochrome p-450 in the liver (Fujita et al., 1984). Nomiyama et al. (1986) exposed rats to >50 ppm for 12 weeks and observed dose-related changes in hemoglobin, hematocrit, and erythroblast count. Myleotoxic anemia has been shown in rabbits (Mazza and Brancaccio, 1967). Teratogenicity data for trichloroethylene in humans are inconclusive. In rat pups, skeletal ossification anomalies were produced by dams subjected to >100 ppm for 4 hr/day on days 8 and 21 of gestation (Healy et al., 1982). Trichloroethylene is weakly mutagenic in some microbial test systems (Clayton and Clayton, 1981). These data are inadequate to assess human carcinogenicity caused by trichloroethylene (ATSDR, 1988b). However, trichloroethylene has been proven to be carcinogenic in rats and mice, causing renal adenomas and carcinomas, lung adenomas, and hepatomas (Maltoni et al., 1986; NTP, 1986b, 1982; Fukuda et al., 1983; Bell et al., 1978). G-17 EFFECTS ON WILDLIFE The availability of data pertaining to the toxic effects of trichloroethylene on wildlife is limited. Chlorophyll-containing algae and plants exposed to trichloroethylene lose their color at 600 mg/L (Verschueren, 1983). LCqq values of approximately 50 mg/L were noted for three freshwater species tested. The EPA has reported lowest effect levels (LECs) for acute exposure to trichloroethylene of 45 mg/L, and 2 mg/L for freshwater and saltwater organisms, respectively. No LECs for chronic exposures have been reported. REGULATIONS. STANDARDS. AND GUIDELINES Human Health OSHA TWA 100 ppm 200 ppm ceiling (29 CFR 1910.1000, Table Z-2) (54 FR 12, 2923-2959, Table Z-1-A) ACGIH TLV-TWA 50 ppm TLV-STEL 200 ppm (ACGIH, 1986) NIOSH TWA 25 ppm (10 hr) (Hazline, 1989) IDLH 5,400 mg/m^ EPA Group B2 probable human carcinogen (IRIS, 1989) lARC Group 3 indefinite animal carcinogen (lARC, 1987) Aquatic Organisms There are inadequate data for establishing ambient water quality criteria for aquatic organisms. G-18 G.8 AIR DISPERSION MODEL The U.S. Environmental Protection Agency's (EPA's) Users Network for Applied Modeling of Air Pollution (UNAMAP) 6 version of the Industrial Source Complex Model (ISC) (EPA, 1 988b) was used to estimate ambient concentrations of materials released from emission sources (stacks) at the WIPP site. Releases resulting from routine operations (long-term) and postulated on-site accident events (short-term), aboveground and underground, were modeled. LONG-TERM MODEL The long-term version of the model (ISCLT) projected the annual average aboveground concentrations, based on the annual average of meteorological data recorded at Carlsbad during the 5-year period 1950 to 1954. Input mixing heights and ambient temperatures were obtained from Holzworth (1972) and National Weather Service records, respectively. The model was run in the "regulatory default" mode. For convenience, the emission rate was assumed to be 10 grams per second (5 gms/sec from each stack). To determine the ambient concentrations resulting from a different emission rate, a ratio of the actual and assumed emission rates was taken and applied to the predicted ambient concentrations. The receptor field consisted of a rectangular grid extending 50,000 meters north, east, south, and west from the originating point of the emission. The physical location of the point of origin was the centerline of the vertical ventilation exhaust duct. Ambient concentrations for underground workers were estimated manually using the following assumptions: • Waste disposal room dimensions are 10 meters by 91 meters by 4 meters • Air velocity is 0.4 m/sec • Air flow is parallel to the long axis of the chamber • Hazardous chemicals are uniformly mixed in the air stream. Using these assumptions, ambient concentration (wg/m^j is the quotient of the release rate (ag/sec) and the ventilation volumetric flow rate (m7sec). SHORT-TERM MODEL Short-term concentrations were estimated by the ISC model (ISCST) running in the short-term mode. Short-term releases were assumed to be discharged from the G-19 emergency filtration system with a flow rate of 60,000 cfm to a single stack. Generic meteorological data (48 combinations of wind speed and stability customarily used in UNAMAP screening models) were used in the ISCST model. For each of the 48 combinations, a 1-hour duration was arbitrarily assigned. In all instances, wind was assumed to be blowing from the south. Hypothetical exposed individuals were located due north of the stack at distances out to 50,000 meters. The assumptions inherent in generic meteorological data are: • The 48 combinations cover the entire spectrum of meteorological conditions that could be obtained. • Each of the 48 combinations can occur at some time or other. Thus, the highest potential short-term exposure can be identified as well as the distance at which this exposure occurs. This is a health-protective approach, since there is a low probability of all the necessary conditions occurring simultaneously. The emission rate was again assumed to be 10 grams per second. To determine the ambient concentrations resulting from a different emission rate, a ratio of the actual and assumed emission rates was taken and applied to the predicted ambient concentrations. Manual calculations were used to estimate ambient air concentrations of hazardous chemicals affecting workers in the waste handling building and underground during postulated on-site accidents. Some accident-specific assumptions are described in Appendix F. For each accident event, it was assumed that the total release was equal to the total mass of volatile organics in the void volume of breached containers. Concentrations in the air in the vicinity of each accident were estimated using these assumptions. For the aboveground accidents, the release was assumed to disperse into a hemisphere which expands at a given rate for a given time based on the air flow into the waste handling building. Concentrations of organics within the volume of the hemisphere were assumed to be uniform. For underground accidents, a similar procedure was followed. However, the underground release was assumed to disperse into an underground mined out area which was 4.0 meters x 3.4 meters x 6.0 meters in dimension. Again, concentration within this volume was assumed to be uniform. Estimations of particulate releases of lead during a single drum fire underground were made based on the vapor pressure of elemental lead. The ISCST model was used to predict the maximum aboveground air concentration on-site. G-20 G.9 EXPOSURE ASSESSMENT Consistent with tlie health-protective approach to risk assessment, potential exposures to releases of hazardous chemicals resulting from routine operations are estimated for hypothetical workers located at the points of maximum on-site concentrations above and below ground identified by the air dispersion modeling. Estimates of potential exposures were also made for a hypothetical resident located at the point of maximum concentration at the WIPP site boundary. EXPOSURE PARAMETERS The potential exposed individual was assumed in each case modeled to weigh 70 kg (about 154 lbs). Adults are used as the model residential receptor since no actual individual exists at the site boundary. In fact, the actual resident nearest to the facility is more than 3 miles from the boundary. The increased sensitivity of the elderly or very young individual from considerations such as body weight is mitigated by the additional dilution of the already very low predicted concentrations at the site boundary (see Section 5.0). The daily respiratory volume was assumed to be 20 cubic meters (m*^) for a 24-hour period (residential exposures) (EPA, 1986) and 12 m^ for an 8-hour period (occupational exposures) (EPA, 1985c). Due to a lack of chemical-specific data for volatile organics, a transfer coefficient of 1 .00 was used to model uptake and absorption via the lungs for these chemicals. The rate of lead deposition in the lungs was assumed to range from approximately 30 to 50 percent of particulates inhaled, while up to 70 percent of deposited lead was assumed to be absorbed within 10 hours of exposure (ATSDR, 1988a). To maintain a health-protective approach, a transfer coefficient of 0.35 (i.e., 70 percent x 50 percent) was used to represent deposition and absorption in the exposure estimates for lead. Potential exposures from the inhalation of hazardous chemicals during routine operations are estimated for occupational and hypothetical residential individuals during above- and belowground operations. The concentrations of hazardous chemicals in air that are predicted at each exposed individual location are evaluated to determine if, based on the postulated scenario, the concentrations will remain constant or increase with time during the exposure period. The aboveground worker and hypothetical residential individual are continually exposed to 42-drum units from the waste handling building during the Test Phase and the Disposal Phase and 6,000-drum units from underground emissions during the Disposal Phase. Similarly, the underground worker is continually exposed to 6,000-drum units during the Test Phase and the Disposal Phase. G-21 The concentration of hazardous chemicals in air from underground operations does not remain constant during the Test Phase because the rooms will not be backfilled and sealed. During the Test Phase, the number of drums increases by 17,600 drums, or 1-drum unit, per year. The concentration of hazardous chemicals in air at the aboveground worker and the residential individual are averaged over the 5-year period by multiplying the predicted air concentration by a weighting factor. A weighting factor (WF) of three was calculated using the following equation: WF= (Ui + Ug + Uj... U^) / n, n = 5 where: Uj = number of drum units present per year, i = 1,...n This method conservatively assumes that the drums will be emitting volatile organic compounds over the entire 5-year period. Based on calculations of the emission period, this is unlikely. For example, the entire mass of methylene chloride would be emitted in 2 years if it continuously diffused through the carbon composite filter at the calculated emission rate provided in Subsection 5.2.4.2, Table 5.35. Therefore, an additional measure of conservatism is added by assuming the organics are emitted over the entire 5-year period. Concentrations available to individuals potentially exposed as a result of accident events were based on the total void volume gas concentrations and short-term modeling employing the specific dispersion characteristics of a given accident area. ESTIMATION OF DAILY INTAKES OF HAZARDOUS CHEMICALS The TLV-based, or IDLH-based estimated intakes (Igj) for the accident scenarios are estimated by the following formula: lai = (Ci)(V)(Ai)(E)(f3) where: L: = TLV- or IDLH-based estimated intake (mg/exposure). 'ai Cj = concentration of constituent in air at the receptor location (mg/m ), V = respiratory volume (m'^/day), A| = transfer coefficient for i**^ chemical. •:>'*i E = seconds or minutes per exposure, fg = conversion factor. G-22 38^S The respiratory volume of 20 m /day and transfer coefficients of 0.35 for lead and 1 .0 for all volatile organic compounds are used in the upper-bound transportation accident to estimate intake of a hypothetical exposed individual located 50 meters from the accident. An exposure of 30 minutes is postulated during the accident. The conversion factor is 1 day per 1 ,440 minutes. The estimated intakes for the accident scenarios postulated to occur during operations at the WIPP are also calculated using the above equation. Because the exposure to a worker is estimated, a respiratory volume of 12 m^workday is used in the calculation of intake. The transfer coefficients of 0.35 for lead and 1.0 for volatile organic compounds were utilized as above. Each exposure period in minutes was then converted, using the factors of 1 hour per 60 minutes and 1 workday per 8 hours. Based on the air modeling, the exposure period for workers during accidents in the waste handling building is 1 minute and in the underground is 15 seconds. A conservative 30-minute exposure period is assumed during the underground fire scenario at the WIPP. For the defined time period of each accident, the concentration of chemicals in air at the location of the worker is assumed to be constant. For routine operation, the annualized averages for each chemical for both the Test Phase and the permanent Disposal Phase were used to estimate the chemical-specific daily intakes for the residential, aboveground occupational, and underground occupational receptors. The daily intake was estimated by where: I, = (C:)(V)(A:) / (f)(W). i = 1 6. (G-2) l^j = estimated daily intake of the i^*^ chemical (mg/kg-day), i = 1,...6, Cj = concentration of the i^^ chemical (jug/m^), V = scenario-specific respiratory volume (m'^/day), Aj = transfer coefficient for the i^^ chemical, i = 1,...6, f = conversion factor (1,000 /^g/mg), W = body weight (kg). G-23 i G.10 RISK ESTIMATION While the estimation of human health risks for this assessment employed a quantitative evaluation of the data available on waste characterization, these estimates are more meaningful when viewed in a relative, and therefore more qualitative sense. The precision of these estimates was limited by the uncertainties associated with the size and quality of the data base. In this assessment, these limitations were partially mitigated by defining a range of extremes. However, overriding uncertainties still persist. An analysis of these uncertainties is given in Section 5.0. LONG-TERM RISK ESTIMATION FOR CARCINOGENS: ROUTINE OPERATIONS For any Class A or B carcinogen (by the classification of the EPA's Carcinogenic Advisory Group) that is projected to average greater than 1 percent by weight of the waste, predicted air pathway exposures that may result from emissions associated with routine facility operations are compared to unit cancer risks (EPA, 1986). Excess incremental lifetime cancer risks resulting from inhalation of vapors are estimated for the exposed individuals associated with each scenario. These estimates are based on guidance provided by the SPHEM and the Air Toxics Assessment Manual (California Air Pollution Control Officers Association [CAPCOA], 1987). Of the representative chemicals for the waste, there are three volatile organics that are Class A or B carcinogens: carbon tetrachloride, methylene chloride (dichloromethane) and trichloroethylene (ICE). The estimated daily intakes for these chemicals were used to estimate the risk of the occurrence of one excess case of cancer as a result of the estimated exposures to these chemicals. This lifetime incremental excess cancer risk is given by Rj = q^ l^jLC , i = 1,...,3, (G-3) where: Rj = excess incremental lifetime cancer risk for the i*"^ chemical, i = 1 3, q.,* = chemical-specific cancer potency factor (mg/kg-day)" , l^j = estimated daily intake of the i*^ chemical for a given individual (mg/kg-day), i =1 3, LC = lifetime correction factor. The cancer potency factors used for carbon tetrachloride, methylene chloride, and trichloroethylene were 1.36 x ^0'\ 1.40 x 10-^, and 1.30 x lO^^ (mg/kg-day)"\ respectively. (IRIS, 1989). G-24 The lifetime correction factor was used to adjust the risk estimates to the specific length of the exposure period. The resulting estimate was interpreted as the lifetime risk of a single excess cancer occurrence based on the specific exposure period. An average lifetime is defined as 70 years (EPA, 1986). For the WIPP, four LCs were required. These are: • Residential: 5/70 and 20/70, because residential exposures are assumed to be for 24 hours per day, 365 days per year for the two exposure periods. • Occupational: (8/24) (240/365) (5/70) and (8/24) (240/365) (20/70), since occupational exposures are assumed to be 8 hours per day, 240 days per year for the entire 5-year and 20-year period. LONG-TERM RISK ESTIMATION FOR NONCARCINOGENS: ROUTINE OPERATIONS Potential risks were estimated for noncarcinogens projected to average greater than 1 percent by weight of the waste (Rockwell, 1988). Estimates of daily intakes for each chemical were compared with acceptable daily levels for chronic intake (AlC) according to procedures for deriving "hazard indices" described in the SPHEM (EPA, 1986). The hazard index (HI) for a given chemical may be defined as the ratio between the daily intake of that chemical and an acceptable reference level. Clearly, an HI less than unity (one) implies that the exposure to the given chemical is acceptable. Hazard indices were calculated for each of these based on the estimated daily intakes. The chemical-specific hazard index was estimated as follows: HIj = l^j / RLj, i = 1 3, (G-4) where: HIj = hazard index for the ith chemical, i = 1 3, l^j = estimated daily intake of the ith chemical for a given individual (mg/kg-day), i = 1 3, RLj = reference level for the i*^ chemical (mg/kg-day), i = 1 3. Here the reference level is the AlC, since exposures for the routine operations scenario are assumed to be over periods of 5 continuous years and 20 continuous years. The AlCs for 1,1,1-trichloroethane and 1,1,2-trichloro-1,1,2- trifluoroethane used in the assessment are 6.3 and 30 mg/kg-day (IRIS, 1989). The oral AlC was used for 1,1,2-trichloro-1,2,2-trifluoroethane because the Inhalation AlC was unavailable. G-25 RISKS ASSOCIATED WITH ACCIDENT SCENARIOS Accident events as defined in Appendix F are short-term events with respect to potential exposures and associated risl■ I • National Association of Governors' Highway Safety Representatives annual meeting in 1989 in Tulsa, Oklahoma. The DOE provided an information booth at this event, which brought together about 400 State highway safety officials. Community Activities. The DOE has held both regularly and specially scheduled community update meetings with community leaders in New Mexico. Updates on the WIPP project have been held in Carlsbad, Artesia, Roswell, Vaughn, and Hobbs. Seminars explaining how to participate in the Federal government procurement system have also been held in these locations for local businesses and contractors. In the informal context of "community days," the DOE has provided the community with opportunities to meet with WIPP staff members and tour its facilities. These events included the following: • WIPP Family Day at the WIPP site in 1987 and 1989. The DOE invited families of WIPP employees to tour the site. These events provided WIPP employees' family members with a general oven/iew of the facility, a demonstration and overview on transportation, an environmental overview, and tours of the Waste Handling Building and the underground areas. • Southeast New Mexico Community Leaders Day in 1988. The WIPP Public Affairs Office organized this event for elected officials and community leaders in southern New Mexico. The event included surface and underground tours and overviews of the WIPP project. H-4 • Southeastern New Mexico Community Days in 1988. Organized by the WIPP Public Affairs Office, this event drew about 1 ,460 persons. The DOE provided overviews and surface and underground tours. • Northern New Mexico Community Day in 1988. The WIPP Public Affairs Office organized this event, which included a general overview, transportation overview and demonstration, environmental overview, and tours of the Waste Handling Building and the underground areas. The event drew about 785 persons. • Water Fair. The DOE assisted the State of New Mexico in gathering water samples from the Carlsbad area by co-sponsoring a Water Fair with the Environmental Improvement Division. More than 70 samples were brought to the fair by residents wishing to receive free water analyses. • Eddy County Fair, 1985 through 1989. The DOE provided an information booth and exhibit at this fair in Carlsbad, New Mexico. About 2,500 people visited the booth. • Lea County Fair, August 1989. The DOE provided an information booth and exhibit at this fair in Lovington, New Mexico; almost 700 people visited the booth. • Eastern New Mexico State Fair in 1986, 1987, 1988, and 1989. The DOE provided an information booth and exhibit at this fair in Roswell, New Mexico. About 2,000 persons visited the booth. • New Mexico State Fair in September 1 988 and 1 989. The DOE sponsored an information booth and exhibit at this fair in Albuquerque, New Mexico. A total of approximately 1 8,000 persons stopped at the booth. • Knowles Frontier Day, July 1989. The WIPP Public Affairs Office provided an information booth and exhibit at this event which is based around fire protection and emergency response; over 1 00 people visited the booth. • Science showcase. In 1987, 1988, and 1989, the DOE participated in the Carlsbad School System's Science Showcase program. The goal of this program is to encourage Carlsbad's young people to view science as a creative discipline that offers a wide range of career opportunities. Each year, more than 1,100 students, teachers, and parents learn about the WIPP at this event. Media. The DOE, through its Office of Intergovernmental and External Affairs and the WIPP Public Affairs Office, is committed to responding to press inquiries with accurate and timely information. In addition to requests for information from southeastern New Mexico, information has been provided to regional media including The Albuquerque Journal and Tribune . Albuquerque television stations, Albuquerque radio stations (KOB and KGGM), the Boise Statesman in Idaho, and the Denver Post and Rocky Mountain News in Colorado. National requests have included inquiries from The Chicaoo H-5 i I Tribune . USA-Todav News. Newsweek and Time magazines, The New York Times . Cable News Network , and The MacNeil/Lehrer Report . Media events sponsored by the DOE were designed to provide the media with in-depth information about key issues of public interest. For example: • The DOE exhibited the TRUPACT-II testing in Albuquerque, New Mexico. Local and national media and public officials were invited to this event. The TRUPACT-II containers were dropped from 30 feet onto an unyielding surface, dropped onto a blunted spike, and burned. • The DOE sponsored a tour to demonstrate the TRUPACT-II full-scale model in Carlsbad, New Mexico; Idaho Falls, Idaho; and Portland, Oregon. The purpose of this tour was to answer questions from interested media about the proposed transportation routes for waste materials and about the proposed contents of the TRUPACT-II containers. Publications. In addition to the public information activities described above, the DOE has prepared numerous publications addressing different WIPP issues. The titles of these publications are: "Waste Isolation Pilot Plant ~ WIPP" "In Situ Testing at the Waste Isolation Pilot Plant" "Visitor Information"* "Certification Requirements" 'Transuranic Waste" "Environmental Protection" "Participants/Lines of Communication" "Why Salt? Why Southeastern New Mexico?"* "Raptor Studies and the WIPP Environment" "Waste Handling Procedures at WIPP" "Commonly Asked Questions"* ■Transportation: A Satellite Tracking System" 'Transportation: TRUPACT-II"* "Safety Throughout the Project" "Waste Handling Building" "Highway Route Selection" "States Training and Education Program" "Public Law 96-164" "Where Will Waste Come From?" "WIPP Project Speakers Bureau Brochure" "Draft Plan for Waste Isolation Pilot Plant Test Phase: Performance Assessment and Operations Demonstration"* "DOE Invites Public Comments on WIPP-SEIS Document." Spanish translations of these publications are being prepared. H-6 H.2 INTERGOVERNMENTAL AFFAIRS An important function related to the WIPP Project Office of Public Affairs is to keep interested government officials informed of key issues and progress related to the WIPP project. In the process, the DOE has worked closely with numerous Federal, State, and local government agencies. In some cases, the DOE has regularly attended meetings of key governmental agencies, and the WIPP project staff members have participated in the ongoing meetings of governmental groups as follows: • The Environmental Evaluation Group (EEG) provides independent oversight of the WIPP project. The group has a professional staff and is responsible to the president of the New Mexico Institute of Mining and Technology. WIPP staff members have conducted 30 quarterly reviews of the WIPP project for the EEG and published 42 reports on their investigation and analyses of the WIPP. • The Radioactive and Hazardous Materials Committee (RHMC) oversees WIPP project activities for the New Mexico legislature. Since 1979, WIPP staff members have attended about 50 meetings of the RHMC. • The Radioactive Waste Consultation Task Force (RWCTF) is an executive task force that oversees the WIPP project for the Governor of New Mexico. In 1985, the DOE was invited to the meetings of the RWCTF and has attended eight meetings since then. • The National Academy of Sciences (NAS) WIPP Panel is composed of 11 prominent scientists and has met approximately 3 times a year since 1979. WIPP project staff members were available for the 30 meetings. • The Pacific States Alliance (PSA) is a four-state committee established to study and recommend measures to transport radioactive material safely through Washington, Oregon, Idaho, and Wyoming. The DOE participated in five meetings in 1 988 and 1 989 with the PSA and attends all PSA meetings to identify concerns, address questions, and provide project updates. • The Western Governors' Association (WGA) is an alliance of governors from 11 western States dedicated to uniformly representing the western governors in intergovernmental affairs. The DOE regularly attends WGA meetings to identify concerns, address questions, and provide project updates. • Congressional support. The WIPP Project Office has responded on numerous occasions to requests for information from different members of Congress and has conducted briefings and tours for interested members who have visited the facility. H-7 In addition to regular involvement with these governmental groups, the WIPP Project Office of Public Affairs has met on request and initiated meetings with other governmental groups interested in the project. These meetings have included the following: • Santa Fe Interested Citizens. Approximately 20 elected and appointed Santa Fe, New Mexico, leaders toured the WIPP site and received briefings. National Congress of American Indians (NCAI). The WIPP Project Office met with NCAI members on four occasions. In December 1987, WIPP staff members met with the leaders of New Mexico Indian Tribes and Pueblos. In February 1988, WIPP staff members met with officials of Indian Tribes and Pueblos from outside New Mexico. In December 1 988, a WIPP representative met with tribal officials at a meeting arranged by the NCAI at a transportation coordinating group meeting. In September 1989, WIPP staff attended and participated in the NCAI-sponsored tribal seminar on nuclear waste. This seminar's purpose was to familiarize Federal officials with tribal cultural and sovereignty rights. All Indian Pueblo Council (AlPC). After AlPC publicly expressed opposition to the WIPP project, the DOE met with the AlPC in 1988 to hear concerns and respond to questions and comments. The AlPC represents New Mexico's 1 9 Indian pueblos on matters for which unity and numbers enhance the pueblos' interests. Interstate Route 84 Task Force. In July 1 988, WIPP staff members conducted a public information tour in Oregon along the route of proposed Interstate Route 84 to provide information on the transport of TRU wastes through Oregon and to identify and address concerns. WIPP project staff members responded to media questions, provided technical expertise, and displayed the full-scale TRUPACT-II model. Hanford Waste Board and Advisory Committee (Oregon). This group sponsored four public information meetings along the proposed Interstate Highway 84 corridor in Oregon. The DOE attended these meetings to provide the public with information on the transport of TRU wastes through Oregon and to identify and address concerns. WIPP project staff members responded to media questions, provided technical expertise, and displayed the full-scale TRUPACT-II model. Western Interstate Energy Board (WIEB). WIPP project staff members attended three meetings held by the WIEB on the WIPP during 1987 and 1988. The WIEB is an interstate compact group representing 16 western States in many environmental and intergovernmental affairs. Southern States Energy Board (SSEB). The SSEB held a meeting on the WIPP in 1987 which WIPP project staff members attended. The SSEB is a non-profit interstate compact serving as the regional representative of 16 southern States in energy and environmental matters. The SSEB also held H-8 a meeting in Carlsbad, New Mexico and toured the WIPP site in September 1988. DOE Field Offices. Personnel associated with or supporting the WIPP Project Office meet with the DOE's Idaho, Oak Ridge, and Savannah River Operations Offices to plan, coordinate, and interface with the States within their regions. Office of Civilian Radioactive Waste Management (OCRWM). The WIPP Project Office met and worked with DOE OCRWM five times in 1987, 1988 and 1989. During these meetings, the DOE attended OCRWM's Transportation Coordination Group meetings to exchange information about transportation policy, hosted the OCRWM Transportation Institutional Support Manager on a visit to the WIPP site, and participated in the OCRWM Institutional Planning for Transportation Activities meeting. Mine Safety and Health Administration (MSHA). Pursuant to a Memorandum of Understanding between MSHA and DOE, the MSHA conducts safety inspections of the underground WIPP facility. Other State of New Mexico Agencies. The DOE met with the State Highway Commission to discuss highway upgrading and with the Radiation Technical Advisory Council to discuss TRU waste transportation and other agenda items. The State Highway Commission has responsibility for maintenance of State roads and shipments of hazardous materials over those roads. The Radiation Technical Advisory Council is responsible for radiation protection in New Mexico. Local government agencies. The DOE met with the Raton, New Mexico City Council in 1 988 to address concerns about waste transportation. After the meeting, the City Council defeated a resolution to restrict the transportation of radioactive waste through city limits. Instead, the council voted to support the New Mexico Municipal League's resolution. The DOE has addressed the Santa Fe City Council on the constituents in and the handling of radioactive mixed wastes and has participated in public forums sponsored by the League of Women Voters, City of Santa Fe, and Santa Fe County. Commercial Vehicle Safety Alliance (CVSA). The DOE met with the CVSA in 1988 to keep informed on CVSA's pilot study for the inspection of radioactive shipments. The CVSA is an alliance of States that is trying to establish uniform inspection procedures for all hazardous materials shipments. Confederated Tribes of the Umatilla Indian Reservation (CTUIR). The DOE attended a CTUIR sponsored workshop on transportation of radioactive materials in 1988. The DOE gave a WIPP update to the CTUIR Board of Trustees in August 1989. The CTUIR is composed of the Umatilla, Cayuse, and Walla Indian Tribes in northeastern Oregon. H-9 Eight Northeast Tribes of Oklahoma. The DOE met with this group in 1 988 to inform the tribes about WIPP issues. This group is a State-chartered forum that represents the Eastern Shawnee, Seneca-Cayuga, Quapaw, Peoria, Wyandot, Miami, Modoc, and Ottawa Indian Tribes on issues of common concern. H-10 H.3 INTERGOVERNMENTAL AFFAIRS AND PUBUC INFORMATION PLAN FOR THE WIPP SEIS In conjunction with the preparation of the WIPP final SEIS, the DOE Albuquerque Operations Office has established an Office of Intergovernmental Affairs and Public Information (lAPI). The objective of the lAPI Office is to ensure that public information and public participation activities for the SEIS are in compliance with the CEO's regulations implementing the NEPA and DOE's NEPA guidelines. To ensure the public has adequate opportunities for involvement in the SEIS, the DOE implemented the following activities: • Intergovernmental Affairs. The DOE has met with 1 ) representatives of the States of New Mexico, Colorado, Utah, Idaho, Washington, Oregon, Wyoming, California, Arizona, Nevada, Kentucky, and Arkansas; 2) the Western Governors' Association; 3) the Southern States Energy Board; 4) the National Congress of American Indians and Council of Energy Resource Tribes; 5) Environmental Protection Agency and the Bureau of Land Man- agement; 6) key environmental groups; 7) the Environmental Evaluation Group; and 8) Congressional representatives from the host and corridor States and from oversight committees such as the House Armed Services Committee. The purposes of these meetings were to discuss the planned content of the SEIS, to receive any input regarding environmental issues, and to review the schedule for completion of the NEPA process. These meetings provided important input into the development of the SEIS, particularly in the focusing of transportation issues and collection of relevant data. The meetings helped the SEIS Office of Intergovernmental Affairs and Public Information identify information needs that government officials and the interested public may have. • Federal Register Notices. A Notice of Preparation of the SEIS appeared in the Federal Register on February 1 7, 1 989. On April 21 , a Notice of Avail- ability for the SEIS was published that also announced the beginning of the public comment period. Subsequently, the DOE published five more Federal Register Notices announcing various changes and additions to the public hearing schedule and extensions of the public comment period (May 26, June 12, June 26, July 7, and July 11, 1989). The total public comment period was 90 days in length. • Toil-Free Request Line. At the beginning of the pubic comment period, the DOE established a toll-free telephone line connected to an answering machine at the SEIS Project Office. This line allowed citizens from around the U.S. to call 24-hours a day, seven days a week to register to speak at the public hearings on the draft SEIS. The line was also available to request copies of the SEIS; to obtain fact sheets, summaries, or other informational H-11 materials on the SEIS; to be placed on the SEIS mailing list; or to receive a return phone call from someone on the SEIS Project Office staff. Mailing List. The DOE developed a comprehensive mailing list for distribution of the SEIS and other materials. The mailing list is a compendium of approximately 2,000 interested citizens; Federal, State, and local agencies; elected officials; tribal officials; public interest groups; and others. Sources for this mailing list consisted of those responding to the February 1 7, 1 989, Federal Register notice, lists from the 1 waste generator or storage facilities, the FEIS distribution list, telephone requests received on the SEIS toll-free telephone line, the DOE Office of Intergovernmental and External Affairs, and others. In response to informational materials prepared by the SEIS Project Office during the early public information efforts on the SEIS, numerous interested parties asked to be added to the mailing list. I Public Hearings. During the 90-day public comment period, the DOE held a total of nine public hearings on the draft SEIS in seven States, including: Atlanta, Georgia Pocatello, Idaho Denver, Colorado Pendleton, Oregon Albuquerque, New Mexico Santa Fe, New Mexico Artesia, New Mexico Odessa, Texas Ogden, Utah May 25, 1989 June 1, 1989 June 6, 1989 June 8, 1989 June 13-14, 1989 June 15-17, 1989 June 22, 1989 June 26, 1989 July 10, 1989 The DOE'S approach for notifying the public of an upcoming public hearing included public service announcements, display ads, press releases, and press conferences. For example, prior to the public hearing in Atlanta on May 25, the DOE sent public service announcements to 27 radio and television stations in Georgia, South Carolina, Tennessee, Kentucky, and Ohio. In the same States, the DOE took out display-type advertisements in 16 newspapers of general circulation. Two days before the hearing, the DOE issued a press release, and on the day before and the day of the hearing, the DOE held press conferences. Similar efforts were undertaken for all of the hearings. As a result of these types of activities, the DOE succeeded in attracting close to a thousand commenters to the nine hearings, in addition to the almost 900 written comments it received. • Others. A variety of press releases and public sen/ice announcements regarding the SEIS have been prepared and distributed to the media and to others on the mailing list. H-12 APPENDIX METHODS AND DATA USED IN LONG-TERM CONSEQUENCE ANALYSES l-i/ii ?<'N< •■>:< "V: TABLE OF CONTENTS Section Page 1.0 INTRODUCTION 1-1 1.1 METHODS 1-4 1.1.1 The NEFTRAN Code 1-4 1.1.2 The Swift II Groundwater Transport Code 1-7 1.1.2.1 Implementation of Brine-Reservoir and Borehole Submodels 1-7 1.1.3 Calculations for Radiation Exposure Pathways 1-13 1.1.3.1 Philosophy of Dose Limitations in ICRP 1-14 1.1.3.2 Release at the Head of the Intruding Well 1-15 1.1.3.3 Geologist Exposure 1-15 1.1.3.4 Doses Received by Indirect Pathways 1-17 1.1.3.5 Exposure from Stock Well Water 1-23 1.1 .4 Calculations for Chemical Exposure Pathways 1-31 1.1.4.1 Lead Solubility in WIPP Composite Brine 1-31 1.1.4.2 Modeling Assumptions for Calculating Lead Solubility in Culebra Groundwater 1-33 1.1.4.3 Lead Solubility in Culebra Groundwaters 1-34 1.1.4.4 Health Effects Associated with Stable Lead from Wind Dispersion 1-34 1.1.4.5 Health Effects from Exposure to Stable Lead in Beef .... 1-42 1.1 .5 Assessing Compliance with the EPA Standards 1-43 1.1.5.1 Performance Assessment 1-45 1.1.5.2 Application to WIPP 1-46 1.2 DATA 1-49 1.2.1 Final Waste Porosity 1-49 1.2.2 Radionuclide Sorption 1-52 1.2.2.1 Rationale for Extrapolation of K^ Values 1-56 1.2.2.2 Rationale for Choices of Recommended K^ Values 1-60 1.2.3 Brine Reservoir Parameters 1-61 1.2.3.1 Initial Reservoir Pressure 1-61 1.2.3.2 Reservoir Thickness 1-64 1.2.3.3 Reservoir Transmissivity 1-64 1.2.3.4 Reservoir Fluid Density 1-65 1.2.3.5 Reservoir Boundary 1-65 1.2.3.6 Reservoir Porosity 1-67 1.2.3.7 Reservoir Compressibility 1-68 l-iii Section Page 1.2.4 Borehole Parameters 1-68 1.2.5 Repository Source-Term Parameters 1-71 1.2.6 Culebra Parameters 1-75 1.2.7 Location of the Stock Well 1-83 REFERENCES FOR APPENDIX I 1-85 -IV LIST OF FIGURES Figure Page 1.1.1 1.1.2 1.2.1 1.2.2 Conceptualization of the high-pressure release scenario 1-8 Hypothetical example of a complementary cumulative distribution function (CCDF) 1-48 Generalized distribution of Castile Formation brine occurrences and approximate extent of the Castile "disturbed zone" in the Northern Delaware Basin 1-62 Representative plan for plugging a borehole through the repository to a Castile Formation brine reservoir 1-70 LIST OF TABLES Table Page 1.1.1 Maximum dose received by a member of the drilling crew (CH TRU waste) 1-16 1.1.2 Maximum dose received by a member of the drilling crew (RH TRU waste) at 1 00 years after site closure 1-16 1.1.3 Radionuclide concentrations in the dry mud pit from CH TRU waste contributions 1-18 1.1.4 Radionuclide concentrations in the dry mut pit from RH TRU waste contributions 1-18 1.1.5 Air concentration and deposition flux values for CH TRU waste 1-19 1.1.6 Air concentration and deposition flux values for RH TRU waste 1-20 1.1.7 Steady-state soil concentrations (CH TRU waste) 1-21 1.1 .8 Steady-state soil concentrations (RH TRU waste) 1-21 1.1.9 Soil-to-plant and forage-to-food-product transfer factors (Case II) .... 1-22 1.1.10 50-year committed dose equivalent (CDE) and committed effective dose equivalent (CEDE) factors (rem///Ci) 1-24 1.1.11 Maximum doses received by a person through indirect pathways for CH TRU waste 1-25 1.1.12 Maximum doses received by a person through indirect pathways for RH TRU waste 1-25 1.1.13 Steps in the calculation of human exposure: from radionuclide concentrations in the stock well water to their concentrations in beef (Cases IIA, MB, IIC, and IID) 1-27 1.1.14 Steps in the calculation of human exposure: from radionuclide concentrations in beef to committed dose to humans (Cases IIA, IIB, IIC, and IID) 1-28 1.1.15 Steps in the calculation of human exposure: from radionuclide concentrations in the stock well water to their concentrations in beef (Cases IIA[rev] and IIC[rev]) 1-29 1.1.16 Steps in the calculation of human exposure: from radionuclide concentrations in beef to commited dose to humans (Cases IIA[rev] and IIC[rev]) 1-30 1.1.17 Element concentrations entered into the EQ3NR code: brine solubility calculations 1-32 1.1.18 Ionic species and total lead solubility in WIPP composite brine 1-33 1.1.19 Element concentrations entered into the EQ3NR code and Pb solubility for the dominant aqueous species: Culebra solubility calculations 1-35 1.1.20 Calculation of the lead ambient air concentration at the exposed individual location, human lead intake via inhalation, and lead hazard index for humans 1-37 l-vi Table Page 1.1.21 Calculation of lead intake by humans, lead concentration in beef, lead intake by humans via beef ingestion, and human hazard index, Case llC(rev) I.39 1.1 .22 Release limits for containment requirements (cumulative releases to the accessible environment for 1 0,000 years after disposal) 1-44 1.2.1 Final void volumes in waste 1-50 1.2.2 Kjj values for radionuclide transport in the matrix of the Culebra dolomite (ml/g) I.53 1.2.3 Kj values for radionuclide transport in the fracture clays of the Culebra dolomite (ml/g) I.53 1.2.4 Kg values for radionuclide transport in the fracture clays of the Culebra dolomite (ml/m^) I.54 1.2.5 Kj and Kg values for radionuclide transport in tunnels, seals, and MB 139 I.54 1.2.6 Sources of K^ data used to estimate values for repository (Case I) and Culebra (Case II) transport (saline water ± organic ligands) I.57 1.2.7 Base-case and range of values of parameters describing the brine reservoir |.63 1.2.8 Specifications for intrusion borehole for Case II simulations 1-69 1.2.9 Specifications for repository parameters used in Case II simulations |.72 1.2.10 Specification of mass inventory of waste radionuclide species and stable lead in the repository for the Case II simulations I-73 1.2.11 Specification of mass inventory of waste radionuclide species and stable lead in the repository for the Case ll(rev) simulations 1-74 1.2.12 Parameter base-case and range values selected for the Culebra dolomite |.76 1.2.13 Free-water diffusion coefficients (cm^/s) for radionuclides and stable lead for the Case II simulations 1-78 1.2.14 Summary of porosities measured in Culebra core samples 1-79 1.2.15 Summary of formation factors and calculated tortuosities from Culebra core samples 1-82 l-vii/viii i 1.0 INTRODUCTION This appendix describes the analytical methods, codes, and exposure calculations used to calculate the impacts from the postulated long-term release scenarios discussed in Subsection 5.4. It also presents the basis for selecting the input data values used in the codes. COMPARISON WITH THE DRAFT SEIS Two principal changes have been made for this final SEIS since the draft SEIS was published in April 1 989. In Case I, a model describing the potential for release from an undisturbed repository, a third scenario has been added, Case IC. This scenario assumes a near-complete failure of tunnel and shaft seals, letting some radionuclide- bearing brine move through those tunnels and shafts to the Culebra aquifers, whence they move to the hypothesized stock well 5 km downstream. In addition, the earlier Cases IIA and IIC have been recalculated as Cases IIA(rev) and llC(rev). These two were chosen for recalculation because they were the extremes of the earlier analyses. Those scenarios were analyzed using a one-dimensjonal, stream- tube, single-point-injection version of the SWIFT-II code. For this final SEIS, these two calculations have been repeated with a more realistic version of that code, one that incorporates two-dimensional transport with lateral diffusion, allows for a time-dependent width of the injection plume, and uses radionuclide-specific diffusivities. The code also had available an improved description of the transmissivity field of the Culebra based on more data (i.e., the results of the H-1 1 multipad tests) than had been available for inclusion in the draft. The more important inputs used in the analyses reported in this final SEIS are compared below with those used in the draft SEIS. Brine reservoir . The description of the brine reservoir under the site is based on measurements made on the WIPP-12 brine reservoir. Somewhat higher initial pressures have since been observed in a brine reservoir at the Beico well to the south, but the brine reservoir description in the revised Case II has not been changed. All the other input parameters for Case IIC are taken at the end of their ranges. Brine reservoir parameters will be varied in the final performance assessment. Borehole properties . The properties of the deteriorated drill hole are already at the extremes of their ranges as given in Subsection 1.2.4. No new data have come to the DOE'S attention to warrant changing these inputs further. Waste properties. A few changes were made in the properties of the waste and the waste disposal panels. The quantities of radionuclides present are larger, because the mass inventory for the whole repository has been scaled up to fill the entire repository to its design volume (Appendix B). 1-1 i Also, the inventory was aged for 175 years instead of 100 years before starting the calculations, this being the sum of the time to the end of the institutional control period (1 00 years) and the time (75 years) until the borehole plug starts to deteriorate. Salado brine inflow . The brine inflow to the panels was increased from 1 .3 m^/yr to 1 .4 m^/yr, as a result of a modification of the Salado lithostatic pressure value (from 14 MPa to 1 4.8 MPa) used in estimating long-term brine inflow rates. Brine properties and inflow into the Culebra. The density of the Castile groundwater was increased from 1 .0 g/cm'* to 1 .24 g/cm^ in the calculations to be consistent with its load of solutes. The net effect has been to decrease the rate at which brine enters the Culebra from the borehole by 30 percent (Table 5.65). For example, in Case IIA(rev), the inflow from the borehole at early times is reduced from 11.2 m^yr to 8.7 m^/yr; and in Case llC(rev) at early times from 99 m^/yr to 74 m^/yr. Groundwater transport . An important difference from the draft SEIS has been to build increased capabilities into the SWIFT II code, allowing it to make more realistic predictions. The original Case II calculations used a one-dimensional stream-tube approach for simulating the transport of contaminants in the Culebra. The revised Case II transport calculations presented in this final SEIS use a two-dimensional system: 1) to provide estimates of breakthrough concentrations for the contaminants at the stock well that more realistically incorporate lateral dispersion and species-specific effects, and 2) to provide quantitative estimates of the cumulative release of radionuclides at distances from the waste panel coincident with the present land- withdrawal boundary and with the stock well location. The added capability for calculations in two dimensions permits an explicit time-dependent size of the initial Injection disturbance shown in Figure 5.7. Species-specific diffusivities . Separate diffusivities have been included for each radionuclide as opposed to one figure for all. Thus in Case IIA(rev), the former figure of 1 X 10"^ cm^/s now ranges from 1.0 to 3.8 x 10"^ cm^/s; and in Case llC(rev) the former diffusivity figure of 5 x 10'^ cm^/s now ranges from 5 x 10"^ cm /s to 2.0 X 10"^ cm^/s (Tables 1.2.12 and 1.2.13). The net effect is to increase the diffusion into the matrix on either side of the fractures. Culebra transmissivitv distribution . The Case II calculations reported in the draft SEIS used a Culebra groundwater flow model calibrated to data collected approximately through October 1987 (LaVenue et al., 1988). An additional modeling effort has been completed that includes an expanded area covered and an expanded and revised data base of transmissivities and fluid heads. The new model differs from the previous one in that it is calibrated to all significant transient events (shaft construction, and the H- 3 and H-11 multipad tests) near the off-site transport pathway between the waste disposal panel and the stock well. (See Subsection 4.3.3.3.) Stock well location . Transmissivity data imply more fracturing south of the site. This results in a flow path that flows first to the east, then south, rather than almost straight south. As a result, the hypothetical stock well has moved about 540 m to the southeast. The distance along the flow path to the site boundary is now 3,610 m instead of 2,860 m, and to the stock well the distance is now 5,960 m instead of 4,840 1-2 ^s^SSSKS •s« m. The straight line distance from the center of the southwest panel to the stock well Is 5.04 km. Integrated releases . A principal purpose for including a two-dimensional flow model instead of a one-dimensional one was to be able to make realistic evaluations of the Integrated releases of contaminants past the site boundary and past the stock well. These results are presented in Subsection 5.4.2.8. 1-3 .1 METHODS 1.1.1 THE NEFTRAN CODE The NEFTRAN code (Network Flow and Transport) (Longsine et al., 1987) is used to calculate radionuclide releases from an undisturbed repository in Cases lA, IB, and IC. It is a groundwater flow and radionuclide transport code developed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Codes that preceded NEFTRAN are NWFT (Campbell et al., 1980) and NWFT/DVM (Campbell et al., 1981). It was designed with the assumption that all significant flow and radionuclide transport progresses along discrete one-dimensional legs or paths. A flow field is represented by the assemblage of these legs forming a network. The solution of the flow equations in NEFTRAN requires pressure boundary conditions and it is required that these conditions be specified as part of the input data. NEFTRAN first solves the flow equations for the network using Darcy's Law. From this, the average interstitial fluid and radionuclide velocities for each leg are calculated. The code then uses a Distributed Velocity Method (DVM) applied over the entire length of the migration path using an average velocity for each isotope calculated from the isotopic velocities in all legs. The DVM technique treats convective-dispersive transport by simulating the movement of an ensemble of representative particles. Dispersion is treated by assigning a velocity distribution to these particle ensembles (Campbell et al., 1986). The user can set up and input any network in the generalized network scheme through a specification of the number of legs, the number of junctions, the junctions bounding each leg, and the junctions where boundary conditions are specified. The hydraulic head gradient provides the driving force for fluid flow through the leg. Conservation of mass at each junction is the assumption that allows the flow network to represent a flow system. This conservation law is given by 2 M, = J (1-1) where j is the index of summation over all legs that are connected at the given junction, and M: is the mass flow rate for the j^" leg in units of mass per unit time. For the case when the j^'^ leg is bounded by junctions j1 and j2, the mass flow rate in the leg is represented by the equation (1-2) 1-4 i S:sj where A, is the cross-sectional area, Kj is the hydraulic conductivity, Em is the elevation of the i* junction, g is the acceleration due to gravity, P,, is the pressure at the i^^ junction, Z| is the length of the leg, and p^ is the fluid density. To account for the effects of brine concentration on the flow, the hydraulic conductivity is weighted as K. = k: : 3 (1-3) Kj is the fresh-water hydraulic conductivity for the j**^ leg, /z^ and p^ are the respective leg. viscosity and density of fresh water at approximately 20 degrees C, /x, and p, are the respective actual viscosity and density in the j*^ '- A matrix equation is developed by applying Equation (1-1) to a boundary junction, substituting Equation (1-2) for M: with Gj = Aj Kj /Zj g, and repeating this procedure for each junction in the network. The resulting matrix equation is G p = e (1-4) where is a matrix of coefficients containing functions of Q = A- KJZ- g, p is a vector of unl t' can be determined by introducing a velocity distribution P(v). The equation describing the density of particle at point x is obtained by summing over all possible donor points in the following manner Po(x, t) dvP(v)p(x - vAt, t - At) (1-7) where At = t - t' I The propagation of the initial conditions from time t' to time t is given by Equation (1-7). An integration over "injection" time must be performed in addition to that over velocity, if a source S(x,t) is included. Sources could result from either transport of wastes from the repository or decay of a radioactive parent. The propagation of the density function from time t' to t (Equation 1-7) is implemented numerically in DVM by discretizing time and space. Also, the velocity-space domain is discretized by dividing the velocity dimension into a few intervals based on equal probability. The propagation of particles is then implemented by simulating the migration of particles in each velocity interval. For the latter, the location of the source is time dependent. NEFTRAN provides for every species to have a different retardation factor in each leg of the migration path. The average species velocity for each leg is treated separately. The mean species velocity caused by dispersion in the leg for the k**^ species in the f^ leg is given by Vkj = Vfj/Rkj (1-8) NEFTRAN maintains a mean velocity for each species while calculating distributed velocities about the mean in each leg. When particles begin a time step as a parent species and end the time step as a daughter, NEFTRAN calculates the average velocity by weighting species velocities with the average time spent as each species Vm(i.. P) = (At) j^l TSi (1-9) 1-6 Mi The output of NEFTRAN consists of the following: 1 . Pressure at each junction of the flow network 2. Volumetric flow rate at each leg of the flow network 3. Discharge rate (in curies/day) of each radionuclide as a function of time at the end of the transport path specified by the user. In the calculation of Cases lA, IB, and IC, the arrival times of radionuclides at the top of shaft or any other point of interest were determined by the times at which the discharge rates rose to 10"''° Ci/day. The threshold used for the arrival of stable lead was 8 X 10"^ mg/L. 1.1.2 THE SWIFT II GROUNDWATER TRANSPORT CODE The SWIFT II (Sandia Waste Isolation Flow and Transport) Code is used to calculate releases from a disturbed repository (Cases HA through IID, including Cases IIA[rev] and IIC[rev]). This code requires specification of the time-varying flow out of a brine reservoir and up the borehole to the Culebra. This flow rate is calculated by analytical models described in this subsection. SWIFT II is a fully transient, three-dimensional code that has been under development and maintenance since 1 975. The program has been comprehensively documented and extensively tested. Calculational comparisons to experimental data have resulted in a program that is both accurate and versatile. SWIFT II solves the coupled equations for transport in geologic media. This code considers the following processes: • fluid flow • heat transport • dominant-species miscible displacement (brine) • trace-species miscible displacement (radionuclide chains). The first three processes indicated above are coupled by means of the fluid density and viscosity. This coupling results in a determination of the velocity field that is needed for a calculation of the third and fourth processes. •.1.2.1 Implementation of Brine-Reservoir and Borehole Submodels Figure 1.1.1 is a drawing of a brine-reservoir breach. It represents a borehole that passes through the repository and connects a brine reservoir to the Culebra. LaVenue et al. (1 988) have detailed the most recent model of the Culebra, having calibrated the steady-state flow field to the field data using SWIFT II. The analyses for cases IIA(rev) and llC(rev) use the transmissivity distribution Culebra model of LaVenue et al. (1988), updated as described in Subsections 4.3.3.2 and 5.4.2.6, with the pressurized brine reservoir specified analytically as a source term. 1-7 '4mmm zed Brine Reservoir (Castile) NOT TO SCALE REF: LAPPIN et al., 1989. LEGEND k 1 - HIGH PERMEABILITY AREA kg - LOWER PERMEABILITY AREA k- - VERY LOW PERMEABILITY AREA o Ty^ - WELL RADIUS r 2 - RADIUS OF HIGH PERMEABILITY AREA r 3 - RADIUS OF LOWER PERMEABILITY AREA r-*co - RADIUS OF VERY LOW PERMEABILITY AREA FIGURE 1.1.1 CONCEPTUALIZATION OF THE HIGH-PRESSURE RELEASE SCENARIO 1-8 In terms of its initial and hydraulic properties, sented by the form the brine-reservoir submodel is repre- Q = Aq + Bq(5p (1-10) where dp is the change in pressure within the Culebra source block m (i.e., the block where the breach will penetrate the Culebra Dolomite) during time-step 61 Quantity Q is the volumetric rate of water injection into block m during time-step ^t. Q, as well as the flow-rate parameters Aq and Bq, are assumed constant during d\. Aq and Bq are defined by equations 1-34 and 1-35, respectively. Q varies as a function of time step to reflect depletion of the brine reservoir. The brine-reservoir submodel is discussed in the following three subsections. The first subsection describes the influence functions P, and W, used to characterize pressure and flow rate, respectively, at the borehole-reservoir interface. The second subsection specifies brine-reservoir response in terms of P, and its time derivative P',. The third and final subsection couples the Culebra and the reservoir to determine a Culebra source term of the form specified in Equation (1-10). Influence Functions . Van Everdingen and Hurst (1949) consider two basic influence functions useful in determining pressure drawdown and flow rate at the borehole- reservoir interface. W, represents a constant-pressure condition at r= r^ (Figure 1.1.1). This term is called the terminal-pressure influence function. The second influence function P, represents a constant-rate condition at r = r^. This term is the terminal- rate influence function. These functions provide basic functions that, through superposition, result in a general solution. P| and W| are derived from a dimensionless flow equation assumed to have cylindrically symmetric form D ^"^D D D 5Ap 3r D aAp dt~ (1-11) where Ap is the pressure drawdown. For well radius r^, porosity (p, total compressibility c, viscosity jli, and reference permeability kQ, the dimensionless quantities in Equation (1-11) are defined as follows: Tq = r/r^. to = t/t^, t^ = 0cr^/i^/ko. ^"d ^a = ^K (1-12) The reference permeability k^ is set equal to k for an homogeneous system. The result is kj = 1 , which is the form of the flow equation given in Van Everdingen and Hurst (1949). Initial conditions assuming a state of equilibrium in the borehole and reservoir result in the equation Ap(rD,tD=0) = (1-13) 1-9 j The boundary condition at the wellbore-reservoir interface distinguishes two influence I functions. For P,, ^^P (r = 1 t ) = -1 (1-14) For W|, Ap(rD=1.tD) = 1 (1-15) j The constant-rate influence function, P,, is obtained as a solution of Equation (1-11) evaluated at the wellbore interface P, = Ap(rD=1,tD) (1-16) I The dimensionless flow rate at the wellbore interface, W,, is given by ^^pQ/^rQ{TQ=^ ,\q). Integration over dimensionless time yields the constant-pressure influence function W, = tr aAp arn dtr (1-17) ''D = -' Van Everdingen and Hurst (1949) assumed homogeneity and derived analytic expressions for P, and W,. Frick and Taylor (1962) tabulated these functions. Observations indicate that brine reservoirs at the WIPP site have heterogeneous hydraulic properties. The brine reservoir properties are based on WIPP-12 data. These data indicate that a relatively high-permeability region k^ located near the well serves as a collection area for a larger region having a lower permeability kg (Figure 1.1.1). Lappin et al. (1989, Section 3.4.3) present interpretations of the WIPP-12 brine-reservoir test data that result in two permeability regions k^ and kg surrounding the borehole. The assumption is made that yet a third low-permeability zone kg provides an effectively infinite source of pressurized brine. Its distance r > r^ is sufficiently great, however, and its permeability kg (equal to the permeability of the intact rock) is so small that it does not participate within the time scale of observations from the WIPP-12 field testing. For the three-zone characterization of the brine resen/oir, the dimensionless permeability function assumes the form kol^o) = ki/ko kg/kg ka/ko 1 < ro < r^g "■02 < I'd - ''D3 I'd > ''D3 (1-18) I The radii r^, r^, and r^ are specified in Figure 1.1.1. For this heterogeneous system, the reference permeability has been arbitrarily set to k^ = k^. Assuming heterogeneous properties makes an analytic solution difficult. As a result, the study uses the numerical MO algorithms of the GTFM model (Pickens et a!., 1987) to generate the desired influence functions. A tabulation of these functions provides input for SWIFT II. Generalized Brine-Reservoir Response . The influence function W, represents the total flow that occurs in response to a pressure drop of unity. If the pressure drop at the wellbore Ap^ = Ap^{rQ=^) is constant, but differs from unity, then the flow rate is Ap^W|. If Ap^ varies as a function of time, then the principle of superposition (Carslaw and Jaeger, 1 959) yields the cumulative fluid flow ''d'^d' 'D Ap' (A)W rt --A)dA (1-19) where Ap,^ denotes the pressure drop at the wellbore-reservoir interface and the prime denotes differentiation with respect to the argument. Carter and Tracy (1960) approximate Equation (1-19) with a form more suitable for numerical computations by assuming a linear variation within a given time step tp" ^ t^ < \q^'*'^ Wd"^' = Wd" + Qd (to - to") (1-20) where a superscript denotes the time level and Qq represents an average rate of flow during the time step. Carter and Tracy (1 960) evaluate the flow rate Qq by equating the right-hand sides of Equations (1-19) and (1-20). Through the use of a step-function Laplace transforms with respect to t^ the equation becomes sAp^ W, = [(Wd" - Qq tD")/s] + [Qd/s2] (1-21) where s is the Laplace-transform variable, and the bars denote transformed quantities. The analysis of Carter and Tracy becomes approximate with Equation (1-21). The identity 1/s^ = sP,W| (VanEverdingen and Hurst, 1949, p. 316) allows one to solve for Ap^. Performing an inverse Laplace transform and solving the resulting equation for Qq gives Qd= (Ap/+1 - Wd" P',"+1)/(P,"+i - to" P',"+^) (1-22) This equation gives the flow rate as a function of the pressure drop Ap^ at the wellbore. The injection volume W can be accumulated numerically as a function of time, and P, and P', can be evaluated from tables. However, Equation (1-22) applies only to the brine reservoir. The hydraulic coupling to the Culebra is presented below. Reservoir-Borehole-Aquifer Couplino . The following equations characterize the pressure response of the brine reservoir. Q = A| -I- B|Ap bw (1-23) 1-11 where the subscript j, is used to distinguish brine-reservoir quantities, and A, = -{QvyWJW"P,'"+V(P,"^^ - to^P,'"-^^) and B, = QJiPr' - to" P|"^' ) (1-24) (1-25) In order to characterize the borehole, the analysis assumes a finite transmissibility T^ in the plugs and rubble. The borehole flow is governed by the equilibrium condition Q=Tw(Pbw-Pw-^sg^'^) (1-26) Saturated brine of density p^ is assumed to occupy the wellbore with a vertical distance Ah separating the centroids of the Culebra and the brine reservoir. The static pressure difference APo = Pbo " Ps ^^^ ' Po ^^" ^® substituted into Equation (1-26), giving the equation Q = T^ (APw - APbw - APo) (1-27) where Ap^^ and Ap^ represent pressure drops of the brine reservoir and the aquifer, respectively. For the pressurized release considered here, Ap^ is inherently negative and APbw inherently positive. Hydraulic coupling to the Culebra focuses on the grid-block m that was penetrated by the wellbore. The pressure p of this grid block, as determined by the finite-difference method, represents an average over the pore volume V of the block. This pressure is influenced by several factors. These include the pore value of the block, its transmissive connections to neighboring blocks, and the hydraulic connection between the wellbore and the grid block. To characterize the latter, the following relation between the borehole flow and pressure differences is assumed Q = M (p^-p) (1-28) which indicates a proportionality between flow rate and pressure drop between the wellbore and the grid-block center. M, the mobility, is given by M = 2jt(iiJn){KIp^ g)Az/ln(ri/rJ (1-29) where K is the hydraulic conductivity of the grid block, Az is the thickness of the Culebra, and p^ and Mq a""® reference values of density and viscosity, respectively. These parameters are used to convert hydraulic conductivity to permeability. The quantities p and m vary as functions of the average salinity of the fluid in the grid block. 1-12 The distance r^ of Equation (1-29) refers to the Culebra Dolomite and should not be confused with the radius (cf. Equation [1-18]) used to characterize the permeability distribution of the brine reservoir. After defining Ar as a pseudo-grid-block radius, Ar = (Ax Ay/jr)^, and after determining the average pressure of the cone of influence in the Culebra Dolomite over the range r^-13 (u/Uq)^ resuspension rate = 10 wind velocity = 3.73 m/s density of dry drilling mud = 1 .4 g/cm^ mud pit surface area = 46.45 m^ depth available for resuspension = 1 .0 cm deposition rate = 1.68 x lO'""^ Ci/m^-s particle size. plume vertical standard deviation = o^ = 40.92 m plume lateral standard deviation = a = 57.68 m (uq = 1 m/s) The source area is approximated by choosing a vertical standard deviation and lateral width of the assumed Gaussian distribution and identifying a virtual point source 20.6 m (22.5 yd) upwind of the leeward side of the pit. Steady-state soil I concentrations at 100 years (within 2 percent of steady state) appear in Table 1.1.7 for I CH TRU waste. RH TRU waste steady-state soil concentrations appear in Table 1.1.8. Transfer factors used in the dose calculations are given in Table 1.1.9. Data on human food consumption per capita are required for the four pathways. Data for the United States were taken from Till and Meyer (1983, Table 6.8). They are 508 j g/day for milk, 86 g/day for meat products, 103 g/day for below-surface crops, and I 202 g/day for above-surface crops. Each steer eats 1 5 kg of fresh forage per day. 1-20 TABLE 1.1.7 Steady-state soil concentrations (CM TRU waste) Nuclide Americium-241 Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Uranium-233 Uranium-235 Concentration (Ci/kg(soil)) 8.62 X 10-"''^ 8.85 X IQ-""^ 4.31 X lO"""^ 4.70 X 10'"''* 1.17 X 10'^^ 7.84 X lO-''^ 3.92 X 10-^ Cf. Corrected from Lappin et al., 1989, Table 7-5. TABLE 1.1.8 Steady-state soil concentrations (RH TRU waste) Nuclide Strontium-90 Cesium-137 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Americium-241 Concentration (Ci/kg(soil)) 1.18 X 10-"''* 1.13 X lO-""* 6.76 X lO'""^ 1.78 X lO-""^ 5.70 X lO-""^ 2.54 X lO*""^ 1.38 X 10 r^5 -21 TABLE 1.1.9 Soil-to-plant and forage-to-food-product transfer factors (Case II) Nuclide Soil-to-Plant (kg-soil/kg-plant) Forage-to-Food Product (day/kg-food or day/liter-milk) Beef: Americium-241 4.2 X 10"; 9.2 X 10"; 1.4 X 10"; 1.4 X 10"; 1.4 X 10"; 1.4 X 10"; 1.7 X 10"; 1.7 X lO"'^ Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Uranium-233 Uranium-235 Strontium-90 1.25 4.8 X 10 ^ Cesium-137 Milk: Americium-241 4.2 X 10"^ 9.2 X 10"; 1.4 X 10"; 1.4 X 10"; 1.4 X 10"; 1.4 X 10"; 1.7 X 10"; 1.7 X 10'^ Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Uranium-233 Uranium-235 Strontium-90 1.25 4.8 X lO'^ Cesium-137 Dried edible below surface crops: Americium-241 6.4 X 10"^ Neptunium-237 Plutonium-238 1.4 X W'l 1.4 X 10"? 1.4 X 10"^ 1.4 X 10"^ 9.0 X io;t 9.0 X 10"; Plutonium-239 Plutonium-240 Plutonium-241 Uranium-233 Uranium-235 Strontium-90 4.7 X 10"; 3.2 X 10"^ Cesium-137 Dried edible above surface crops: Americium-241 2.8 X 10"^ 1.5 X wi 1.7 X 10"; 1.7 X lo;; 1.7 X 10"; 1.7 X 10"! 1.0 X 10"^ 1.0 X 10""^ Neptunium-237 Plutonium-238 Plutonium-239 Plutonium-240 Plutonium-241 Uranium-233 Uranium-235 Strontium-90 2.2 2.2 X 10"^ Cesium-137 3.6 5.0 1.0 1.0 1.0 1.0 3.4 3.4 8.1 2.0 2.0 5.0 2.7 2.7 2.7 2.7 6.1 6.1 1.4 7.1 10^ lo;^ 10^ 10 10 10^ 10" 10- 10 10 -3 10^ 10^ 10^ 10^ 10"^ 10" 10- 10^ 10' Cf. Lappin et al., 1989, Table 7-6. Note . All data are from Till and Meyer (1983), Tables 5.17, 5.18, 5.36, and 5.37. Transfer factors were selected assuming that vegetables would be washed before being eaten. 1-22 The analysis used various computer codes to tabulate the committed effective dose equivalent for various body organs per unit activity inhaled or ingested. The organs included in these tabulations are those explicitly considered by the ICRP to be at risk. The committed dose equivalent is the total dose equivalent that an organ or tissue of the body is expected to receive over the 50-year period following exposure. It is recognized that in most environmental applications, more rigorous evaluation requires information on the time variation in the dose equivalent rates for the various tissues at risk. This information provides the time dependence of environmental conditions, and therefore, that of the intake could be assessed with consideration of the years of remaining life. It is also recognized that overestimates by factors of 2 to 3 in the risk are possible by not using the time-dependent nature of the organ dose equivalent rates and the years of life remaining. Committed dose equivalent (CDE) and committed effective dose equivalent (CEDE) factors used in the analysis are shown in Table 1.1.10. Tables 1.1.11 and 1.1.12 list the maximum doses received by a person through indirect pathways for each nuclide of importance. These pathways include ingestion of foods provided by animals feeding on the land, as well as crops grown below and aboveground (root and leafy vegetables). The inhalation pathway assumes a breathing rate of 2.7 x 10"^ m^/s. The tables summarize the exposure calculated 'or a person living on the hypothetical farm described in the subsection below for a 50-year committed effective dose equivalent. 1.1.3.5 Exposure from Stock Well Water In addition to radiation exposure at the top of the intrusion borehole at the WIPP site itself in Case II, there is a possible exposure pathway through a stock well that taps the Culebra aquifers; a stock well that is at the closest point downstream for the salinity of its water to be low enough for cattle to drink (Subsection 1.2.7 below). There is no radionuclide or stable lead release to the stock well until after 200,000 years, and hence no human exposure. The starting point for all six variants of Case II is the concentrations of radionuclides at the stock well (Table 5.68). Discharge rates and concentrations at 10,000 years are used because they are still rising at that time, which is the end of the calculation. The human exposure calculated is the exposure of a person who eats beef from those cattle. The calculation assumes that eight cattle graze in the square mile (2.6 km^) around the well. Each animal requires 13 gal/day (49 l_/day) of water to drink. Therefore, allowing for rainfall at the rate of 20 cm/yr and evaporation at the rate of 200 cm/yr and a stock pond whose area is 139 ft^ (0.0013 hectares), this well is pumped at the rate of 120 gal/day (460 L/day). The result is an evaporation-caused increase in radionuclide concentrations by a factor of 1.1635. The maximally exposed individual is assumed to eat beef from the cattle at the rate of 86 g/day (NCRP, 1984, Table 5.3). 1-23 TABLE 1.1.10 50-year committed dose equivalent (CDE) and committed effective dose equivalent (CEDE) factors (rem// R). The effects of the uncertainties will be incorporated into a single CCDF for each disposal system. If this CCDF meets the requirements above, then that disposal system is deemed to comply with Part B of the EPA standards. 1.1.5.1 Performance Assessment A performance assessment consists of four parts: • Scenario development and screening, • Consequence assessment, • Sensitivity and uncertainty analysis, and • Regulatory compliance assessment. Scenario development and screening examine possible future events or processes that might affect a repository, assign probabilities to them, and determine which possibilities merit detailed consideration. Consequence assessments estimate the releases that might arise from the scenarios of interest. Sensitivity and uncertainty analyses identify important processes and parameters and illuminate the sources and extent of 1-45 uncertainties in tine consequence assessment, thus enabling the regulator to evaluate the confidence that can be placed in the results. Finally, a regulatory compliance assessment combines the results of the scenario analyses, consequence assessments, and sensitivity and uncertainty analyses and determines whether the repository is in compliance with the requirements of the EPA standards. In a Monte Carlo simulation using deterministic models for predicting consequences, the following approach can be used to generate a CCDF. (If a stochastic or some other model is used, another technique would be used to generate a CCDF.) This process is described in greater detail in Hunter et al., 1986. Assume that K scenarios have been identified as important. For WIPP, K might be as large as 10. These scenarios are analyzed by choosing appropriate ranges and distributions for the model's input parameters and then statistically sampling from these ranges to obtain sets of input values for the scenarios. This sampling must be done by some means such as Latin Hypercube sampling, so that all samples have the same probability of occurring. (The same set of input parameters is used for all scenarios In order to ensure that any variation observed between scenarios is due to scenario differences and not to differences in sampling.) For WIPP, the number of sets of input parameters, N, might be 100. Thus there will be NK sets of consequences, R^^, calculated in the performance assessment. For WIPP this may amount to as many as 1,000 calculations. For each scenario, a probability will also be estimated. The sum of these probabilities P,^ cannot, statistically speaking, be greater than one. Each probability is therefore normalized by dividing them by the sum of probabilities SP,^ and by N. Similarly, the consequences are normalized by dividing them by the release limits given In Table 1.1.22. There result NK pairs of normalized consequences and associated probabilities. These pairs are ordered by the magnitude of their consequences, with the largest consequence first. Then the CCDF [P(Release > R,^,^)] is the sum of all normalized probabilities P|^/(NzP,^) down to that point on the list. A CCDF generated in this manner is actually a step function consisting of NK steps (Figure 1.1.2). The EPA containment requirements are indicated as the forbidden area in the upper right hand area in this figure. If the CCDF remains outside this forbidden area, the standard is met. This particular hypothetical example indicates a region of possible violation. 1.1.5.2 Application to WIPP Case II will probably be one of the scenarios entering into the CCDF for the WIPP. Its probability of occurrence is high. Using EPA's figure of 30 holes per square kilometer of repository area, 51 holes may be drilled into the WIPP in 10,000 years (i.e., 1 every 200 years on the average). Inasmuch as the waste disposal panels subtend about 7 percent of the repository area, the probability of a hole intersecting a waste panel is [1 - (1 - .07)^^] = 1 - .025 « .98 1-46 Then, because about half the WIPP area appears to be underlain by a brine reservoir (Earth Technology Corporation, 1987), this probability should be multiplied by 0.5. Finally, superimposed on this should be the probability that the drill hole will actually go as aeep as the Castile brine reservoir-it could be being drilled for potash evaluation. Taking this probability arbitrarily as another 0.5, the net probability of Case II is 0.98 X 0.5 X 0.5 = .25. (Not knowing what the other scenarios might be, this probability cannot be normalized by dividing it by eP,^.) If Case llC(rev) should be one of the 100 or so sets of input parameters with which Case II is analyzed its probability would be the overall Case II probability divided by 100, or 2.5 X 10"^. However, this is not likely, as Case llC(rev) analyzes the consequences of an extreme case in which all the input parameters (except the initial pressure in the brine reservoir) are taken at the extremes of their ranges. The calculation of integrated release for Case llC(rev) in Subsection 5.4.2.8 of this SEIS would therefore appear as one of the last steps in the lower right hand corner of Figure 1.1.2. Thus this calculation alone, although with an integrated release of 3.2, does not per se indicate noncompliance with the regulations. 1-47 4 /% 1.0 -L_ 1 1 ^ EPA CONTAINMENT L ,^ REQUIREMENT LU 10-1 " ^ ^-, CO < POSSIBLE LU _J ^ 10-2 /VIOLATION _ m CC / Li. O o § > o 10""3 IJ z I \J ^^ CQ "" CO A k o CC a. 10-4 — ^ 10"5 , , ,^ 1 O-"" 10 + 101 102 103 SUMMED NORMALIZED RELEASES, R REF: HUNTER, et al., 1986. FIGURE i.1.2. HYPOTHETICAL EXAMPLE OF COMPLEMENTARY CUMULATIVE DISTRIBUTION FUNCTION (CCDF) 1-48 1.2 DATA 1.2.1 FINAL WASTE POROSITY If it is assumed, for purposes of calculation, that structural changes do not take place after mechanical compaction, the void volume remaining within a room after waste compaction determines the maximum amount of brine that may eventually enter the room. This value is difficult to estimate, however, because the mechanical and physical properties of the waste are highly variable and poorly characterized. The compressive stress exerted by the surrounding salt is not sufficient to completely eliminate all voids in the waste. As the waste is compacted, its resistance to additional densification increases, and it becomes rigid enough to prevent further void reduction. A near-term limiting void volume within the repository, associated with purely mechanical densification and expected to be attained in 60 to 200 years, is used for this analysis and is assumed to represent a "steady-state," Even after this time, the state of the repository will continue to change, as biological decomposition and chemical corrosion alter the chemical and structural nature of the waste. This longer-term evolution of the physical state of the repository is expected to be complex, to occur over a long period of time, and to include interactions between compaction processes and possible repository expansion as a result of gas generation. Its quantitative characterization may never be possible. At least for metal wastes, densification may continue beyond that produced by early room closure, and conse- quently the near-term limiting void volume is considered the greatest void volume that will exist within the waste. The final room porosity enters the calculations in this report in three ways. First, the estimated porosity is used to estimate the final permeability of the repository. This value is used in the Case I calculations, but does not enter directly Into the Case II calculations. Permeability is used there to determine whether Castile brine-reservoir fluids effectively mix with the waste in the repository. Second, the final porosity estimate is used to estimate the volumes available within the repository for gas storage or saturation with brine. Finally, the porosity estimate is used to determine the volume of brine available to dissolve radionuclides. Dissolution is limited either by the mass required to reach the solubility limit of individual radionuclides or by the total mass of the radionuclides present, whichever is less. The final void volume used here is based on the distribution of waste types in storage (Table 1.2.1) (DOE, 1988a). A total of 6,804 drums are assumed to be stored in seven- pack configurations within a disposal room, each with an internal volume of 0.21 m'^. In assigning final porosities to each component, combustible waste (low-strength plastics, paper, and rags) is assumed to have such low strength that the near-term interconnected void porosity will be 0.1 or less after compaction to lithostatic pressure (approximately 14 MPa). Because combustible waste will collapse to a dense, interlocking structure, its hydraulic response is considered to be similar to that of silt, with a hydraulic conductivity of 10"^ m/s. (The porosity is n = vyv, where V is the 1-49 TABLE 1.2.1 Final void volumes in waste Waste Form Combustible Sludge Metal/Glass Other Total Emplaced Percent by weight of total waste in storage 30 17 33 20 100 Initial volume in disposal room (m"^) 429 243 472 286 1430 Percent of solids per drum 24.8 66.5 21.9 Final Solids volume (m^) 106 162 103 93 464 Void volume (m'^) 12 18 68 25 123 Waste volume (m^) 587 i I Cf. Lappin et al., 1989, Table 4-5. I Sources . DOE, 1988a; Clements and Kudera, 1985. V = V^ + V3, where Vg is the solid compacted volume, and V^ is the void volume volume that the waste would occupy if no voids were present. Later, the void ratio, e, will be used, which is defined as e = V^/Vg, or e = n/1 - n.) The mechanical properties of sludge are not well defined, but this category of waste represents only 1 7 percent of the total waste inventory. Sludge is much more difficult to compact than combustible waste, and therefore its total void content after compaction is likely to be greater. The same interconnected porosity, 0.1, is assumed for it in the compacted state, however, because many sludges may have a high cement content and are expected to form hydration products that decrease void interconnec- tivity. In the absence of any data about the hydraulic conductivity of sludge, a value two orders of magnitude greater than for grout has been assumed. The hydraulic conductivity of grout is 1 x lO""*^ m/s (Coons et al., 1987), implying a final-state conductivity of 1 x 1 0'^ m/s for sludges. 1-50 The strengths of metallic and glass wastes make them much less compactible than combustible and sludge wastes. Most of the waste is metallic in content. The final porosity assumed for metal and glass waste is 0.4, based on powder-metallurgy literature (Hausner and Kumar, 1982) and on data on supercompaction, which suggests that compaction of metal waste to much greater than 0.6 of theoretical solid density is not likely. A lower final porosity for the metal waste can be expected, however, if the crushed-salt and bentonite backfill intrudes into the open spaces between the pieces of metal, a process that could reduce porosity by as much as 50 percent. Thus, a lower bound to metal-waste porosity is taken to be 0.20. The properties of the waste category referred to as "other" remain undefined. In the absence of further information about the composition of this waste, its compacted porosity is assumed to be the average porosity of the combined combustible, sludge, and metal and glass waste categories, weighted according to the portion of the inventory that each represents. Final void volumes for combustible, sludge, and metal and glass waste categories are given in Table 1.2.1. The volume of solid waste per drum is computed using the average initial void fraction of each waste category (Clements and Kudera, 1985). Adding in the void volume of the unspecified "other" cateaory of waste (20 percent of the inventory), the total void volume per room is 123 m , corresponding to a solids volume of 464 m^. This 123 m^ volume, divided by the total volume, 587 m^, yields a porosity of 1 23/587 = 0.21 0. If the void volume in the metal waste is assumed to be reduced 50 percent by salt intrusion, the net void volume per room is approximately 123-34 = 89 m^. This corresponds to a porosity of 89/587 = 0.152. The "expected" final void volume for the consolidated waste is the average of the estimated void volumes, or 106 m^ per room, corresponding to an interconnected void porosity of 106/587 = 0.182. To be conservative, the release scenarios in Subsection 5.4 use a saturated void volume of 123 m^. The estimates above apply only to the waste and do not include any final-state porosity of backfill in the room, because the compacted salt-bentonite backfill is expected to be relatively impermeable. The void volume calculations take no credit for the fact that the metal and glass waste may contain minor amounts of easily compacted materials such as combustibles or sorbents. The only study that has quantitatively inventoried the contents of TRU waste in detail (Clements and Kudera, 1 985) showed that metals represent only about 80 percent by weight of the INEL metal waste. The remainder of the metal category contents is combustible material (12 percent) and cement (5 percent), which would reduce its compacted porosity. A major uncertainty in this analysis is introduced by the absence of any information about the compactibility of the various waste types, although tests to determine compactibility are in progress. An estimate was made of how rapidly the limiting void volume within a disposal room is approached (Figure 5.3). The calculated rate of closure of an empty disposal room (Munson et al., 1989) was used to determine the void volume at a given time. The void volume was obtained by subtracting the volumes of the solids in the waste and backfill and the volume of brine flowing into the room as a function of time (Nowak et al., 1988) from its current volume. Figure 5.3 is not completely consistent with values -51 listed in Table 1.2.1 because, in the absence of experimental results, equal rates of consolidation of waste and backfill were assumed. An assumption in using closure data for an empty room for this estimate is that any backstress by the room contents is insufficient to retard void reduction. This appears to be warranted for room porosity greater than approximately 0.3: finite-element calculations show that backstress is significant only during the latest stages of closure. The no-backstress assumption is also consistent with the current model for compaction of the waste, which assumes that the final void volume depends only on the stress applied to the waste, and not on the stress history; that is, the only effect of backstress Is to prolong the time required to achieve the final compacted state. This assumption, however, is another source of uncertainty. Estimates using these assumptions show that the limiting void volume could be achieved in 40 to 60 years; 60 to 200 years is assumed in Subsection 5.4.2.4, Brine Inflow. The amount of brine flowing into the room during 60 years is estimated to be between 6 and 37 m^, a factor of 4 less than would be required to saturate the 1 23 m^ of void volume at final-state. In fact, all this brine can be sorbed by the bentonite in the backfill (Subsection 5.4.2.4). In addition, the pressure of decomposition gases within the room, even assuming none leak out, would not reach lithostatic pressure in 60 years (Lappin et al., 1989, Subsection 4.10.2). 1.2.2 RADIONUCLIDE SORPTION The Kjj values used in the SEIS analyses are summarized in Tables 1.2.2 through 1.2.5. Table 1.2.2 contains K^ values that are used to calculate radionuclide retardation in the matrix of the Culebra dolomite. Tables 1.2.3 and 1.2.4 contain sorption ratios for the clays that line the fractures in the aquifer. Table 1.2.5 contains K^ values for use in radionuclide transport in the tunnels, seals, and Marker Bed 139 at the repository level. If the volume of the clays within the fractures is known, then the K^jS in Table 1,2.3 can be used to calculate retardation within the fractures using the following expression Rf = 1 + />cKdc(V^). (1-38) where K^^ is the distribution coefficient for the clay given in Table 1.2.3; p^ is the density of the clay (2.5 g/cm^); 3^ is the thickness of the clay coating the fracture; and 3 is the fracture width (Neretnieks and Rasmuson, 1984). Surface area-based distribution coefficients Kg (ml/m^) for the clay are listed in Table 1.2.4. These were calculated from the K^s assuming a surface area of 50 m^/g. This is similar to the surface area of 32 m^/g measured by Nowak (1980) on a reference montmorillonite used in europium sorption studies and within the range of 15 to 88 m^/g measured by Soudek (1984) on montmorillonite used in ion exchange studies. A retardation factor for use in a transport equation for fracture-dominated flow where sorption occurs on the surface of the fracture fill clay can be calculated as R = 1 -H aKJ

is the porosity. Similarly a retardation factor for use in transport through the porosity-controlled flow in the tunnels and seals can be calculated as R = 1 + /9Kd (1 - (1-40) where K^j is the distribution coefficient given in Table 1.2.5, p is the grain density, and

tr) o o o ^ X X X X r^ -- '- "* ^ A A 'h 'h % X X in in A A *"^ S, o ^ X o t: X ^t ■.- <0 '■^ CM CM in O O 2$ X o o o CO c (D o >. >> Si "^ (0 (0 ■Jo <2 O O c o < CQ 3 a> (D S O >*-• C (0 % oc Z x: o c O 00 o o> a C CQ 00 O) U 0) O Q CQ •c o Q. 0} ir E O E < o 00 CM O o CVJ ,- a 3 Q. (0 o 'c CO O) o o cc 0) IS CVJ o o Ai (0 OC Q O Q 0} m O) m CD m (D (0 O C s 2 (D 0} c "5 0) cc CO 5? (S o ° X X •^ CO 8 o in CO o o o (S o ° X ^^ ^ S^ ;l 2o o 8S o "^ CVJ p> o X f^ r^ 1- iJ^ CO ^ « jC C < to CD o 0) c o Vi O m CO 0) c ffl m 2 c c ±i c c *w (0 w CQ CQ CO JO (D C o z (D ■Jo C o (0 O A ^ CO b o CO n o * * c 00 >> ^ _l T" e8 ^ (D CO c § r^ CO 05 Q. a O 1-58 a> c a T3 9 T3 _2 o c o O (O CNJ tu —I 03 < 3 LU Q. z E o E < U O CC 0) to Q g ■D ffl o A w o CM O ■^ CO CO O 2 X X «*> CM X X A A U5 ■"t O o CO (O T- O o (O o o (A CM 8 CO o ^ S 00 (O ^ ^ o X "^ ^ « CO cm' CM X w CM O _ O O"^ TJ- T- T- O X X T- 00 r«* o O - i5 OJ n 2 In ^ Vy ® to w D C D 3 3 3 CC o < CC o CC o a> c o t5 O o (D 2 15 c to c o c o Vi o to ^ (0 o O O o c g o 00 (0 O o I o o o o c c c cF CT c o o o o ■^ 3 3 ctf m. s o o o CM CM o CO CO CO X X CO sz p o K ^ 8 O o> 1 ■(3 c ""^ (0 S c •c S" o o ^ a o c5 Q ^C CO 9. 9. X 2 X V (D c co'^ ^o w ^ •s; Q ^ Americium . K^^ values are decreased by factors of 3 to 1 ,000 from values listed in the table to account for the potential effects of organic complexation. For example, Swanson (1 986) found that EDTA significantly decreased Am sorption onto kaolinite and montmorillonite. The magnitude of this effect was a function of the pH and concentrations of EDTA, Ca, Mg, and Fe in solution. Curium . K^ values were decreased by factors of 3 to 1 00 from the values listed in Table 1.2.6 based on the assumption of similar behavior to Am and Eu. Uranium and Neptunium . Generally, low K^^s have been measured in waters relevant to the WIPP. The K^^ of uranium is very dependent on the pH and the extent of complexation by carbonate and organic ligands. A low value (K^j = 1) has been assumed in the SEIS to account for the possible effects of complexation and competition. Theoretical calculations (Siegel et al., 1989) and arguments based on similarities in speciation, ionic radii, and valence (Chapman and Smellie, 1 986) suggest that the behavior of neptunium will be similar to that of uranium. Thorium . There are few data for thorium under conditions relevant to the WIPP. Thorium K^ values were estimated from data for plutonium, a reasonable homolog element (Krauskopf, 1986). Data describing sorption of Th onto kaolinite (Riese, 1982) suggest that a high concentration of Ca and Mg will prevent significant amounts of sorption onto clays in the repository. Stability constants for organo-thorium complexes suggest that organic complexation could be important in the repository and may inhibit sorption (Langmuir and Herman, 1980). Radium and Lead . There are very few sorption data for radium under conditions relevant to the WIPP. K^ values in Table 1.2.6 were estimated from assumption of homologous Ra-Pb behavior in Tien et al. (1983). Data from Riese (1982) suggest that Ra will sorb onto clays but that high concentrations of Ca and Mg will inhibit sorption. Langmuir and Riese (1985) present theoretical empirical arguments that suggest that Ra will coprecipitate in calcite and gypsum/anhydrite in solutions close to saturation with respect to these minerals. 1-60 1.2.3 BRINE RESERVOIR PARAMETERS In order to model the hydraulic connection between the Culebra member and a brine reservoir in the underlying Castile Formation, it is necessary to realistically define the hydrologic parameters that govern the transient hydraulic response of the brine reservoir. These parameters include Pj = initial reservoir pressure b = reservoir thickness T = reservoir transmissivity p = reservoir fluid density r^, = effective distance to the reservoir boundary

>..•••• ERDA-9 DOE-1* • /Belco^^J • • . + \. / DISTURBED ZONE «" Cabin Baby • 1 o • • T21S T22S T23S R30E R31E R32E 6 mi R33E LEGEND REF: LAPPIN, 1988. A BOREHOLE IN WHICH CASTILE FORMATION BRINE WAS ENCOUNTERED • BOREHOLE THAT PENETRATED THE CASTILE FORMATION BUT DID NOT ENCOUNTER BRINE (ERDA-9 IS INCLUDED FOR REFERENCE ONLY) FIGURE 1.2.1 GENERALIZED DISTRIBUTION OF CASTILE FORMATION BRINE OCCURRENCES AND APPROXIMATE EXTENT OF THE CASTILE "DISTURBED ZONE" IN THE NORTHERN DELAWARE BASIN -62 TABLE 1.2.7 Base-case and range of values of parameters describing the brine reservoir Parameter Symbol Base case Range Units Initial pressure P| Effective thickness b Transmissivity of Tj inner zone Distance to intermediate rg zone contact Transmissivity of Tq intermediate zone Distance to outer r3 zone contact 12.7 7.0 to 17.4 MPa Transmissivity of outer zone m Fluid density p^ Porosity ^ Compressibility a 7.0 7.0 to 24.0 m 7x10-^ 7x10"^ to 7x10"^ m^/s 300 100 to 900 m 7x10-^ 7x10"^ to 7x10"^ m^/s 2,000 30 to 8,600 m IxlO'"*^ Constant m^/s 1240 Constant kg/m^ 0.005 0.001 -0.01 1x10"^ 1x10-''° to 1x10"^ 1/Pa Cf. Lappin et al., 1989, Table E-4. For this modeling study, the base-case reservoir pressure is taken from the highest pressure monitored during the testing of the brine reservoir at WIPP-12, which is equivalent to a pressure of 12.7 MPa at the reservoir center. Of the 13 wells in the northern Delaware Basin that have encountered brine reservoirs, only 4, including WIPP-12, have been tested adequately enough to estimate the formation pressure. These pressures range from 12.6 to 14.3 MPa at formation depth (Popielak et al., 1983). Minimum pressures for nine other wells have been estimated from the minimum pressure needed to allow flow at the surface. From these nine estimates, minimum formation pressures range from 7.0 to 17.4 MPa. The range of initial reservoir pressures for this study is therefore taken to be 7.0 to 17.4 MPa. The base-case value is representative of the WIPP-12 reservoir (for which the best data are available) and is 12.7 MPa at reservoir depth. 1-63 1.2.3.2 Reservoir Thickness In most cases, the brine reservoirs encountered in the Castile Formation are in the lower portion of the uppermost anhydrite unit present at that location. The uppermost anhydrite unit at WIPP-12 is Anhydrite III, which is locally 96.6 m thick (Popielak et al., 1983). The WIPP-12 brine reservoir was at the base of Anhydrite III and appears to have been limited to a small fractured zone. Anhydrite III at WIPP-12 was mapped by coring, caliper logs, acoustic televiewer logs, neutron logs, and spinner logs (Popielak et al., 1983). A review of these observations identified seven megascopic fractures in Anhydrite III. All these fractures were high- angle fractures with dips ranging from 70 degrees to vertical. Only two showed any evidence of brine production, as identified by the spinner log conducted by the USGS (D'Appolonia Consulting Engineers, 1982). The uppermost brine-producing fracture (fracture C) extended from 916.2 to 917.1 m; the lowermost (fracture D) extended from 919.0 to 921,1 m BGS. These depths were taken from the acoustic televiewer log. The spinner log defined the inten/al from which nearly 100 percent of the flow was coming as that between 916.2 to 921.4 m BGS, which correlates well with both the caliper log and the acoustic televiewer log (Popielak et al., 1983; D'Appolonia Consulting Engineers, 1982). Because the resen/oir is heterogeneous and composed of high-angle fractures, its thickness is difficult to define from borehole reconnaissance at a single location. The base-case effective thickness of the reservoir is estimated to be 7 m and to occur between 915.3 and 922.3 m BGS. The center of the reservoir is taken to be at a depth of 91 8.8 m BGS, which is the center of the interval that produced nearly all of the inflow at WIPP-12 (D'Appolonia Consulting Engineers, 1982). All downhole pressures are referenced from a depth of 918.8 m BGS. The base-case effective thickness of 7 m shown in Table 1.2.7 can be considered a minimum thickness. From the center of the reservoir to the base of Anhydrite 111 is approximately 12.0 m. The maximum effective thickness will be considered 24 m centered at 918.8 m BGS. Because the product of hydraulic conductivity and thickness (transmissivity) cannot be determined in the reservoir characterization analyses, sensitivity calculations will be performed upon transmissivity. The variation in transmissivity caused by thickness uncertainty will be less than the variation in transmissivity caused by uncertainty in hydraulic conductivity. As a result, the total variation in formation transmissivity will be driven largely by hydraulic conductivity variation, as described in the following subsection. 1.2.3.3 Reservoir Transmissivity For modeling, the WIPP-12 reservoir is conceptualized as being composed of two separate, concentric, fractured media with different transmissivities. Because this modeling study will allow very long flow periods in the brine reservoir, the far-field matrix is also expected to contribute to the reservoir response. This matrix is modeled by attaching an infinite low-transmissivity zone to the outside edge of the intermediate zone through the application of a Carter-Tracy boundary condition (Carter and Tracy, 1960; Reeves et al., 1986). This outermost zone represents the intact Castile Anhydrite III. Popielak et al. (1983) determined that the intact formation matrix had a permeability -64 of less than 2 x 10"^^ m^. Assuming a thickness of 7 m, the transmissivity of the outer zone is equal to approximately 1x10'^^ m^/s. The transmissivity of the outer zone is at least six to eight orders of magnitude smaller than the base-case transmissivity of the inner and intermediate zones. For this modeling study, the outer zone represented by the Carter-Tracy boundary condition is assigned a constant transmissivity of 1 x 10"^^ m^/s. From hydraulic interpretations, It was determined that the inner region of the reservoir can be modeled as a cylindrical zone having a transmissivity of 7 x 10"^ m /s and extending out from the well to an effective radius of 300 m. The remainder of the reservoir was interpreted as having a smaller mean transmissivity. This intermediate zone is assigned a lower transmissivity equal to 7 x 10"^ m^/s out to a radius of 2,000 m. These values are interpreted from WIPP-12 testing and are considered base- case values listed in Table 1.2.7 for the hypothetical brine reservoir. The transmissivities of these two zones are estimated to range, somewhat arbitrarily, by two orders of magnitude from the base-case values. The only Castile brine reservoir transmissivity data available for comparison to these base-case values and ranges are presented by Popielak et al. (1983), who determined transmissivities from as low as 1.6 x 10"^ m^/s at ERDA-6 to as high as 8 x 10^ m^/s at WIPP-12. 1.2.3.4 Reservoir Fluid Density The brine from the WIPP-12 brine reservoir has an average level of total dissolved solids of 328,000 mg/L, as determined from laboratory analyses of 1 3 water samples (Popielak et al., 1983). The average specific gravity, based on 59 field analyses, is 1.215. In addition to these traditional analyses, four borehole-pressure-gradient surveys were performed in 1982 and 1983 at WIPP-12 as part of the hydraulic testing program. These surveys showed pressure gradients ranging from 0.0121 to 0.0123 MPa/m, with an average of 0.0122 MPa/m. This average gradient corresponds to an average fluid density of 1 240.6 kg/m^. For this study, the base-case brine-reservoir fluid density is taken to be 1241 kg/m^. This parameter will not be varied, and a representative range Is not defined. 1.2.3.5 Reservoir Boundary Because of the isolated distribution of brine resen/oir encounters in the Castile Formation, the reservoirs must be considered limited, with some outer boundary beyond which hydraulic communication is minimal. Methods used to infer the limits of brine reservoirs in the Castile Formation are varied. One method is to look at a map of wells penetrating the Castile Formation and identify which wells did, and which did not, encounter a brine reservoir. When a well that encountered a brine reservoir is surrounded by wells that did not, the distance of the latter wells from the brine reservoir well represents a maximum radius for the boundary of that reservoir. For example, WIPP-12 is surrounded by four nearby wells that did not encounter brine In the Castile Formation. These four wells range in distance from 2 to 3 km from WIPP- 12. Therefore, if it is conservatively assumed that WIPP-12 is located at the center of the reservoir and the reservoir is circular, the WIPP-12 reservoir has at most a 2,000 m radius. Most brine reservoirs in the Castile in the northern Delaware Basin are found 1-65 to have radii varying from approximately 800 to 3,200 m. Other shapes than circular are possible, but they have not been included in the analysis. A recent investigation of a different kind that may be used to delineate the extent of the WIPP-12 brine reservoir is a time-domain electromagnetic survey (IDEM) performed at land surface over the waste emplacement panels (Earth Technology Corporation, 1988). This study suggests that there is a low-resistivity body, interpreted as a brine reservoir, within the Castile Formation under portions of the waste emplacement panels. If one assumes that this brine is connected to the WIPP-12 brine reservoir, then one reservoir boundary is at least 1,600 to 2,000 m from WIPP-12. Another method of inferring the reservoir extent at WIPP-12 is to estimate the total bulk volume of the reservoir using the concept of the storage coefficient of an elastic aquifer. The storage coefficient is defined as the volume of water removed from a vertical column of aquifer of height m and unit basal area when the head declines by one unit (Domenico, 1972). The equation for the storage coefficient can be written as S = bpg{a+4>^) (1-41) where b is the aquifer thickness, p is fluid density, g is the acceleration due to gravity, a is the compressibility of solids, (p is the porosity, and y? is the compressibility of the fluid. Domenico (1 972) showed that the amount of water released from storage for a given head decline over an area A is equal to AV = S A A h where h is the given head decline. Equation (1-42) can be expanded to Ay = p g {a + T-f, 238p, 8.77 X 10^ 1.71 X 10^ 3.06 X 10^ ob . 226R3 _ 210pb 234u 2.44 X 10^ 6.26 X 10'^ 3.01 X 10^ 23^h 7.70 X 10"^ 2.02 X 10"2 0^ 226Ra 1.60 X 10^ 9.89 X 10"^ O'* 210pb 2.23 X 10^ 7.64 X 10^ 0^ 241pu . 24lAm . 237np 241pu 1.44 X 10^ 1.03 X 10^ 4.56 X 10"* o" ^233u ^229^^, 24lAm 4.32 X 10^ 3.43 X 10° 2.25 X 10^ 2.06 X 10^ 237np 2.14 X 10^ 7.05 X 10""* 1.53 X 10"* 7.93 X 10"* 233u 1.59 X 10^ 9.65 X 10"^ 9.82 X 10^ 9.81 X 10^ 229Th 7.43 X 10^ 2.10 X 10"^ 0^ Stable Pb n.a. n.a. n.a. 1.33 X 10^ 1.33 X 10^ Cf. Lappin et al., 1989, Table E-5. Note , n.a. means not applicable. * Initial inventory in Ci is presented in Table B.2.13. ® The transport calculations start 175 years after the beginning of institutional control. ^ Because ^^^Pu and ^^^Pu have short half-lives and large retardation factors, their migration from the source is minimal. Therefore, the conservative approach converts all^^Pu and ^'^''Pu to daughter products at simulation beginning. ^ Because of large retardation factors relative to their parents, ^^h and ^^h migration is controlled by their parents. Because of this fact and the fact that both nuclides have very little mass in place at 1 75 years, they are not considered initially present at 1 75 years. ^ These nuclides are not present in quantities large enough at 175 years to warrant source inclusion. 1-74 Source Term in the Repository . The concentration of the waste species in these fluids is constrained by their solubilities. For the radionuclides, the solubilities were set equal to 10"^ molar for Cases IIA, IIA (rev), and IID and to 10"^ molar for Cases IIB, IIC, and lie (rev). The solubility for stable Pb was set at 116 mg/L in the repository fluids. All fluids entering the borehole from the waste panel had concentrations at these values except as modified by radioactive decay and the total mass available in one panel. The solubility of stable lead in the Culebra groundwaters was specified at 54 mg/L. 1.2.6 CULEBRA PARAMETERS A fractured, porous medium is assumed to exist along the travel path between the breach borehole and the stock well. The definition of the flow path, the stock-well location, and the solute-transport properties within the Culebra are discussed below. Additional discussion on fracturing in the Culebra and its effect on hydraulic and tracer tests is presented in Reeves et al. (1987). The base case and range of values for the Culebra parameters are summarized in Table 1.2.12. The range of values is presented for discussion purposes only. They are not used in the Case IIA and IIA (rev) simu- lations. For Cases IIB, IIC, IIC (rev), and IID, lower or higher end values of the range were selected, whichever would result in more rapid or longer distance solute transport. A double-porosity flow is assumed along the travel path. The double-porosity data base is limited; base case and ranges of parameter values are documented using available data, but must be considered as uncertain. Regional Flow Field . A review of the hydrologic modeling for the Culebra in the vicinity of the WIPP site is discussed in Lappin et al. (1989, Section 3.3.5). The Culebra groundwater flow model by LaVenue et al. (1 988) was used in Cases IIA, IIB, IIC, and IID for estimating the Darcy velocity distribution in the regional flow field and for determining the travel path from the borehole to the stock well. Calibration of the model included hydrologic data available up to about October 1 987. The model was calibrated to undisturbed head conditions only and did not include data from the large- scale multipad pumping tests that have been performed at the WIPP site. For Cases IIA (rev) and IIC (rev), this flow field description was updated to include all data collected through June 16, 1989. (See Subsection 4.3.3.2.) As discussed above in Subsection 1.2.4, the borehole is assumed to be drilled through the center of the southwestern waste panel. A particle-tracking code was used to determine the flow path for transport from this release location to a hypothetical stock well. The location of the stock well was based on two constraints: the well is assumed to lie on a flow path from the breach borehole, and the well must be located in an area where the water is potentially fresh enough to support stock. 1-75 TABLE 1.2.12 Parameter base-case and range values selected for the Culebra dolomite Parameter Symbol Base Case Range Fracture porosity Longitudinal dispersivity Matrix distribution coefficient Case IIA: Plutonium Americium Uranium Neptunium Thorium Radium Lead Cases IIB, IIC, IID Plutonium Americium Uranium Neptunium Thorium Radium Lead 0' Kd Kd Kc Kc Kd Kd Kd Kd Kc Kc 1 .5x1 0'^ 1 .5x1 0"^ to 1 .5x1 0'^ 100 50 to 300 50 200 1 1 50 0.1 0.1 25 100 1 1 25 0.05 0.05 Units Free-water diffusivity Radionuclides: Case IIA Cases IIB, IIC, IID D' D' D' 5x1 o;? 1x10"^ 5x10"'^ 5x10"^ to 9x10"^ n.a. n.a. cm?/s cmx/s cm^/s Cases IIA (rev] IIC (rev) Stable Pb: Case IIA Cases IIB, IIC, 1 and IID D' D' See Table L2.13 4x1 o;5 1x10"^ n.a. n.a. cm^s crrr/s Matrix tortuosity Case IIA, IIA (rev) Cases IIB, IIC, IIC (rev), IID 0.15 0.15 0.03 0.03-0.5 n.a. n.a. Fracture spacing Cases IIA, IIA (rev) Cases IIB, IIC, IIC (rev), IID 2L' 2L' 2L' 2.0 2.0 7.0 0.25-7.0 n.a. n.a. m m m Porosity 4>' 0.16 0.07-0.30 Cases IIA, IIA (rev) Cases IIB. IIC, IIC (rev), IID 0' 0.16 0.07 n.a. n.a. m ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g ml/g Cf. Lappin et al., 1989, Table E-6. Note- The Culebra groundwater flow model presented in LaVenue et al. (1988) was used for calculating fluxes and determining flow paths. The transient fracture flux along the flow path from the release point in the Culebra aquifer to the off-site stock well is calculated through hydraulic coupling of the brine reservoir, borehole region, and Culebra aquifer. 1-76 Free-Water Diffusivitv . Base-case and range values for free-water diffusion coefficients for the radionuclides of interest and stable lead are presented in Lappin et al. (1 989; Subsection E.2.4.2). For the calculations reported in the draft SEIS, a single value was necessary for all members of a decay chain because of the numerical formulation of the SWIFT II model. For Case IIA, values of 1x10"^ cm^/s and 4x10"^ cm^/s are selected for the radionuclides and stable lead, respectively. Values a factor of two smaller were used for the Case MB, IIC, and IID simulations. For Cases IIA (rev) and lie (rev), improvements in the SWIFT II codes permitted species-specific diffusion coefficients (Table 1.2.13). The base-case and range of values selected for this study (Tables 1.2.12 and 1.2.13) are substantially lower than those in Reeves et al. (1987) for two reasons: 1) the previous study did not specifically address the radioactive decay-chain members identified in the present study, and 2) the much higher salinities that are a result of flow from the Salado and Castile can cause a reduction In the free-water diffusivity by as much as a factor of two. Matrix Porosity . Porosities have been measured in the laboratory for 82 core samples of Culebra dolomite from 15 borehole or hydropad locations at and surrounding the WIPP site. The results are summarized in Table 1.2.14. Porosities were measured by the Boyle's Law technique using helium or air on all samples and by the water- resaturation technique on 30 samples. An excellent correlation was obtained between porosity values from the two techniques. From the 82 samples with porosity measurements using the Boyle's Law technique, an average porosity of 15.2 percent was obtained with a range from about 3 to 30 percent. For comparison, core samples from the H-3 and H-1 1 hydropads, which are the two hydropads closest to the off-site pathway, had average porosities of 19.8 percent (6 samples) and 16.2 percent (10 samples), respectively. Porosities ranged from about 17 to 24 percent for the H-3 hydropad and about 10 to 30 percent for the H-11 hydropad. Matrix porosities of Culebra dolomite measured by Sandia National Laboratories using the ^Na diffusion technique range from 1.1 to 8.7 percent. Corresponding tortuosities range from 0.03 to 0.09. The porosities calculated from the diffusion experiments are termed diffusion-porosity values and are lower than those measured by Boyle's Law or mercury-porosimetry techniques. These values lie at the lower end of the range of values shown in Table 1.2.14. Possible explanations for the differences between values measured by these different techniques include sample heterogeneity, incomplete resaturation of previously dried samples, and deviations of actual pore geometry from the Idealized model assumed in simple versions of Pick's First Law of Diffusion for solute migration in a porous rock. In general, the samples used in the diffusion measurements are fine-grain dolomites free from large cracks and are chosen for mineral homogeneity and structural competence. No claim has been made that these samples are representative of the Culebra dolomite in general or that these results are transferable to field-scale transport. For transport along the off-site pathway in the Culebra, a base-case matrix porosity of 16 percent is chosen for the Cases IIA and IIA (rev) simulations. For Cases IIB, IIC, IIC (rev), and IID, a matrix porosity of 7 percent is selected as a lower end value. 1-77 Table 1.2.13 Free-water diffusion coefficients (cm^/s) for radionuclides and stable lead for the Case II simulations I I ! Element Pu Am U Np Ra Pb Th Case IIA (rev) Case IIC (rev) Range of Values in Literature® 1.7 X 10"^ 1.8 X 10"^ 2.7 X 10-^ 1.8 X 10-^ 3.8 X 10-^ 4.0 X 10"^ 1.0 X 10-^ 8.5 9.0 1.4 9.0 1.9 2.0 5.0 10- 10" 10^ 10 10^ 10^ 10- -7 4.8 X 10-^- (3x 10"^) 5.3 X ~ 1.1 X 5.2 X ■■^- (3x 10-^i 10-^- 4.3 X 10-^ 10 10 -7 - (3x 10^ 7.5 X 10^ 8x 10-^ 5x 10'^ - 1.53 X 10-^ ® Data from values compiled by Brush (1988) (indicated by parentheses); values calculated from the Nernst expression by Li and Gregory (1974) (underlined); and measurements by Torstenfelt et al. (1982) (all others). Temperature dependence has not been considered for the recommended values. Literature values are further discussed in Lappin et al. (1989), Section E.2.4.2. Cf. Lappin et al., 1990. 1-78 Table 1.2.14 Summary of porosities measured in Culebra core samples Porosity Determination (%) Sample Identification Helium Water Well Number or Air Resaturatlon H-2a -1 11.6 11.3 H-2a -2 12.2 H-2b 1-1 14.1 H-2b 2-1, 3-1 15.4 H-2b 1-2 11.8 H-2b 2-2, 3-2 10.3 H-2b1 -IF 10.5 H-2b1 -1 8.2 8.8 H-2b1 -2 14.2 H-2b1 -3 15.3 15.8 H-3b2 1-3 18.8 h-3b2 1-4 16.8 H-3b3 2-3, 3-3 18.0 H-3b3 2-4, 3-4V 20.2 H-3b3 1-6 24.4 H-3b3 2-5, 3-5 20.5 H-4b 1-9 29.7 H-4b 2-6, 3-6V 20.8 H-5b -1 12.5 H-5b1 -1A 13.0 H-5b1 -IB 15.6 H-5b1 -2 22.8 23.7 H-5b1 -2F 24.8 H-5b1 -3 13.3 12.8 H-6b 2-7 10.8 H-6b 2-8 11.6 H-6b 1-7 10.7 H-6b 1-8 25.5 H-7b1 -1 17.7 18.1 H-7b1 -1F 14.9 H-7b1 -2A 20.6 H-7b1 -2B 27.8 H-7b2 -1 15.9 14.8 H-7b2 -2 11.8 12.9 H-7C -1A 12.5 12.9 H-7C -IB 16.5 H-7C -1C 13.4 H-7C -IF 13.8 H-10b -1 8.9 H-10b -2 11.5 11.7 H-10b -2F 6.6 H-10b -3 11.2 10.6 1-79 ::><■: TABLE 1.2.14 Concluded Porosity Determination (%) Sample Identification Helium Water Well Number or Air Resaturation H-11 -1 15.5 15.3 H-11 -2 10.5 11.3 H-11 -2F 10.4 H-11b3 -1 30.3 27.5 H-11b3 -1F 22.3 H-11b3 -2 9.9 10.3 H-11b3 -2F 12.3 H-11b3 -3 13.0 12.6 H-11b3 -4 15.2 H-11b3 -4F 22.4 W-12 -1A 2.8 W-12 -IB 11.4 W-12 -2 11.6 11.9 W-12 -2B 12.6 W-12 -2F 13.5 W-12 -3 13.4 13.0 W-13 -1 14.3 15.2 W-13 -2 21.9 22.6 W-13 -2F 26.0 W-13 -3A 17.9 W-13 -3B 9.7 W-25 -1 11.5 12.0 W-26 -1 12.4 12.2 W-26 -1F 11.2 W-26 -2 12.6 12.6 W-26 -3 12.7 W-28 -1A 14.2 W-28 -1B 13.0 W-28 -2 18.7 18.8 W-28 -3 17.0 16.9 W-28 -3F 17.9 W-30 -1 12.8 12.4 W-30 -2 15.0 15.2 W-30 -3A 17.6 W-30 -3B 14.9 W-30 -3F 14.9 W-30 -4 23.9 AEC-8 -1 7.9 8.6 AEC-8 -IF 12.2 AEC-8 -2 10.9 10.6 Number of Samples = 82 Average = 15.2% Standard Deviation = 5.3% Range = 2.8 to 30.3% j Cf. Lappin et al., 1989, Table E-8. -80 ^fSSEiSS^Hipft Although lower values have been measured or derived, an average lower value of 7 percent along the flow path is considered most representative. Matrix Tortuosity . Tortuosity values for dolomite are not available, although a review of the literature does permit an estimation of a potential range. Bear (1972), in his review of unconsolidated media, presents values ranging from 0.3 to 0.7. De Marsily (1986) reports tortuosities varying from 0.1 for clay to 0.7 for sand. Barker and Foster (1981) report diffusion coefficients for CI' in chalk samples that indicate tortuosities of 0.02 to 0.17. Katsube et al. (1986) calculate tortuosity values from 0.02 to 0.19 from diffusion experiments on crystalline-rock samples. As noted earlier, diffusion experiments performed by Sandia National Laboratories on a limited number of core samples have yielded tortuosities in the range of 0.03 to 0.09. Matrix tortuosity estimates for the Culebra were calculated based on formation-factor and matrix-porosity determinations on 15 core samples. The values, ranging from 0.03 to 0.33 with an average value of 0.14, are summarized in Table 1.2.15. For the Case IIA and IIA (rev) simulations, a base-case matrix tortuosity of 0.15 was selected as representative. This value is the same as that used in the regional-scale transport simulations presented in Reeves et al. (1987). A lower-end estimate of 0.03 for matrix tortuosity was selected for the Case IIB, IIC, IIC (rev), and IID simulations. Rock Density . Rock-density determinations were performed on 73 Culebra core samples from 1 5 borehole or hydropad locations. The values range from 2.78 to 2.84 g/cm^ with an average and standard deviation of 2.82 and 0.02, respectively. A value of 2.82 g/cm^ was chosen as the base-case value for all simulations. Fracture Porosity. Estimates of the fracture porosity can be obtained by interpreting tracer tests conducted at sites exhibiting double-porosity transport behavior. Tracer tests have been performed at five locations (H-2, H-3, H-4, H-6, and H-1 1 hydropads) at the WIPP site. Of these, the tests conducted at the H-3, H-6, and H-1 1 hydropads appear to demonstrate fracture-transport behavior as evidenced by the very rapid tracer breakthrough between wells on at least one flow path at each hydropad site. Detailed test interpretations have been reported for only the H-3 hydropad (Kelley and Pickens, 1986). A first estimate of the fracture porosity can be calculated from the convergent-flow tracer tests by the relation ^f = Q tp/ TT r^t b (1-47) where i^:' m- LIST OF FIGURES Figure Page L.2.1 Cross section of TRUPACT-II L-3 L.3.1 Cross section of NuPac 72B cask L-19 LIST OF TABLES Table Page L.2.1 Regulatory testing requirements and the actual TRUPACT-II certification testing program L-1 1 L-iv L.1 INTRODUCTION This appendix was prepared in response to comments on the draft SEIS. It provides information that supplements Subsection 3.1.1.3, which discusses the shipping containers and casks to be used for transporting TRU waste to the WIPP. It discusses both the TRUPACT-II container, which will be used to transport contact-handled TRU waste, and the NuPac 72B cask, which will be used to transport remotely handled TRU waste. The discussions include descriptions of the TRUPACT-II and the NuPac 72B designs, but they are mainly directed at the certification of these designs by the U.S. Nuclear Regulatory Commission (NRC) and the analysis and tests necessary to obtain the certification. The design of the TRUPACT-II was certified by the NRC on August 30, 1989. This appendix presents a detailed discussion of the NRC requirements for the designs to be certified. It further describes how compliance has been demonstrated for the TRUPACT-II container and how It will be demonstrated for the NuPac 72B cask. Also discussed are the NRC's requirements for the fabrication, operation, and maintenance of the shipping containers or casks, including restrictions on the waste to be transported. The last section describes quality assurance for the TRUPACT-II and NuPac 72B programs. The initial Certificate of Compliance for the TRUPACT-II by the NRC limits shipments to only certain waste forms (see Annex 1 to this appendix). In the future, the DOE will apply to the NRC to amend the Certificate of Compliance to include other TRU waste forms known to exist. Most of the information in this appendix was obtained from the Safety Analysis Report for the TRUPACT-II container (DOE, 1989a), the TRUPACT-II Operation and Maintenance Manual (DOE, 1989b), and the Quality Assurance Plan for the Transportation and Receipt of Transuranic (TRU) Waste (DOE, 1989c). L-1 L.2 THE TRUPACT-II SHIPPING CONTAINER The TRUPACT-II container will be used for shipping contact-handled (CH) TRU waste. It has been designed and constructed to meet the regulations issued by the NRC for 'Type B packaging"^ In 1 CFR Part 71 . A Type B packaging with double containment is the type of container that must be used for the transport of TRU waste containing more than 20 curies of plutonium per package. A certificate stating that the TRUPACT-II complies with the NRC regulations was issued by the NRC on August 30, 1989. The NRC certificate is reproduced in this appendix as Annex 1 . The TRUPACT-II shipping container has been designed to be rugged and lightweight, because these characteristics enhance the safety of transportation. The use of rugged, yet deformable, packaging features provides capabilities which prevent the release of contents if it were subjected to extreme abuse in an accident. A lightweight design allows the transport of a larger payload per shipment while meeting highway weight limits, thereby reducing the number of waste shipments. Before proceeding with the fabrication of the TRUPACT-II containers, four full-scale containers were built and tested. One of these served as the engineering prototype; the other three were full-scale containers that were tested in accordance with the NRC's requirements for certification. In addition, a thorough analysis of the CH TRU waste was performed to establish payload-control procedures that meet NRC criteria for transport. These controls have been approved by the NRC as acceptable methods for complying with the applicable regulations for payloads. L.2.1 DESCRIPTION OF THE TRUPACT-II SHIPPING CONTAINER As shown in Figure L.2.1, the TRUPACT-II container is a cylinder with a flat bottom and a domed top; it is transported in an upright position. The overall dimensions of the TRUPACT-II are approximately 8 ft in diameter by 10 ft in height; the inner containment vessel is approximately 6 ft in diameter by 8 ft in height. To provide double containment for the TRU waste, it consists of an inner containment vessel and an outer containment vessel; the latter is part of the outer containment assembly. NRC regulations require the two separate levels of containment to be used for shipments of plutonium in excess of 20 curies per container. ^ In the NRC regulations governing the transportation of radioactive materials (10 CFR Part 71), the term "packaging" is used to mean the shipping container or cask and the term "package" is used to mean the shipping container together with Its radioactive contents. L-2 STAINLESS STEEL SKIN (PROTECTIVE STRUCTURE AND IMPACT LIMITER) HONEYCOMB DUNNAGE AND IMPACT LIMITER LYTHERM INSULATION INNER CONTAINMENT VESSEL OUTER CONTAINMENT VESSEL FOAM HONEYCOMB DUNNAGE AND IMPACT LIMITER 10" FORKLIFT POCKETS' NOT TO SCALE FIGURE L.2.1 CROSS SECTION OF TRUPACT-II L-3 The inner and the outer containment vessels have removable lids that are held in place by banded lockrings and retaining tabs. The containment vessels are nonvented and are designed for a maximum normal operating pressure of 50 pounds per square inch. The capacity of each TRUPACT-II shipping container is 7,265 lb of payload, including pallets, slip sheets, and waste, packed In either 55-gal drums or two 67-cubic-ft standard waste boxes. The maximum gross shipping weight of a loaded TRUPACT-II container is 1 9,250 lb. The weight of the payload is restricted to meet highway weight limits. Up to three TRUPACT-II containers may be transported in each truck shipment. They will be hauled on a custom-designed semitrailer pulled by a conventional tractor. L.2.1.1 Inner Containment Vessel The inner containment vessel is a stainless-steel pressure vessel that contains the waste payload. The payload is protected by spacers that are made of aluminum honeycomb and are located in each of the two domed heads of the inner vessel (Figure L2.1). The lower body of the inner containment vessel has a closure ring with two grooves, each containing an 0-ring seal. The upper lid of the vessel has a mating flat surface that seals against the two 0-rings once the lid and the body are assembled. Compression of the 0-rings between the lid and the body form a bore-type seal. As the lid is lowered onto the body, retaining tabs on a lockring slide through recesses in the mating tabs on the body. When the lid is fully engaged, the lockring can be rotated to the closed position; the lockring cannot be rotated unless the lid is correctly mated to the body. The locking mechanism secures the lid to the body, and this maintains leaktight seals under both normal and accident conditions. L2.1 .2 Outer Containment Assembly The outer containment assembly is made of stainless steel and polyurethane foam. It consists of an exterior stainless-steel shell and a stainless-steel pressure vessel, the outer containment vessel (Figure L.2.1). Between these steel shells there is a layer of fire-retardant polyurethane foam approximately 10 inches thick. The steel walls surrounding the foam layers are lined with a heat-resistant ceramic-fiber paper, which enhances the resistance of the polyurethane foam to fire damage. On the outside of this foam and ceramic fiber, the exterior stainless-steel shell acts as a protective structure and an impact limiter. This multilayered design increases the overall strength of the container and provides the ability to withstand potential accidents associated with transport. Like the inner containment vessel, the lower body of the outer containment vessel has a seal flange ring with two grooves, each containing an 0-ring seal. The upper lid of the vessel seals against the two 0-ring seals of the body when assembled. The lockring secures the lid in place and maintains leaktight seals under both normal and accident conditions, providing the same containment capability as the inner vessel (double containment). L-4 L.2.2 NRC CERTIFICATION ^ The DOE agreed to have the NRC certify the designs of the shipping containers or casks used for the transport of contact-handled or remotely handled TRU waste, respectively. This agreement was stated in the second modification (August 4, 1 987) to the consultation and cooperation agreement between the DOE and the State of New Mexico (see Subsection 10.2.5). The NRC requirements for the certification of shipping containers and casks are included in 10 CFR Part 71, "Packaging and Transportation of Radioactive Materials." There are two basic types of packagings for radioactive materials: Type A and Type B; the latter is the type that the NRC requires for the transport of the type of waste that will be sent to the WIPP. Type A packages must withstand normal conditions of transport without loss or dispersal of their radioactive contents as demonstrated through tests outlined in regulations issued by the Department of Transportation (49 CFR Part 173). Type B packaging must withstand both normal and accident transport conditions without releasing its radioactive contents. In order to transport TRU waste containing more than 20 curies of plutonium per package, the Type B packaging must have a double containment. L2.2.1 Procedure for NRC Certification L.2.2.1.1 General Procedure In order for the design of a packaging to be certified, the applicant (usually the developer of the packaging) must submit to the NRC a description of the package; an evaluation of the package; and a description of the quality assurance program for the design, fabrication, assembly, testing, maintenance, repairs, modification, and use of the proposed package. The description of the package must be In sufficient detail to identify it accurately and provide a sufficient basis for evaluation. For the packaging, this description must include a number of specified items, such as the containment system, materials of construction, weights and dimensions, methods of fabrication, and lifting and tiedown devices. In addition, the description must include information about the payload. For example, it must identify the radioactive constituents of the payload and their quantity, Identify fissile constituents, describe the chemical and physical form, and state the maximum heat generated by the radioactive payload. The evaluation of the package is to consist of a demonstration that the packaging complies with the standards specified in 10 CFR Part 71. The standards in Subpart E include general design requirements (e.g., fastening devices for containment vessels, 2 To be consistent with the NRC regulations, the terms "packaging" and "package" are used in this section to mean the shipping container and the shipping container loaded with radioactive waste, respectively. L-5 maximum surface temperatures), requirements for lifting and tiedown devices, external radiation limits, and special requirements for packages containing fissile materials or Plutonium in excess of 20 curies. Subpart F specifies the evaluations that must be performed to demonstrate that the package can withstand normal and accident conditions without loss of integrity. The evaluations of response to normal transportation conditions are to include the following: exposure to high and low temperatures, reduced and increased external pressure, vibration, and a water spray simulating a heavy rainfall; a free drop for a specified distance (referred to as a handling drop); and an impact by a vertical steel cylinder, 1-1/4 inches in diameter, dropped from a height of 40 inches onto the most vulnerable surface of the package. It is also necessary to determine and demonstrate the response of the package to accident conditions. The requirements for this evaluation are discussed in detail in the next subsection. For the quality assurance program, the applicant must identify any established codes and standards proposed for use in the design, fabrication, assembly, testing, maintenance, and use of the package. After the application is submitted, the NRC may at any time request additional information. The application is reviewed by the NRC's technical staff, who prepare a safety evaluation report for the particular package design. If the staff determines that all pertinent requirements are met, the NRC issues a certificate of compliance. As already mentioned, the NRC certificate of compliance for the TRUPACT-II design was issued on August 30, 1989. This certificate is reproduced in full in Annex 1 to this appendix. The certificate of compliance specifies procedures for the fabrication, operation, and maintenance of the packaging and defines the payload that may be transported. The certificate is valid for a period of 5 years. At the end of this period, it must be renewed by submitting an application for renewal. L.2.2.1.2 Demonstration of Abilitv to Withstand Accident Conditions To be certified by the NRC as Type B (10 CFR 71.73), a candidate packaging must demonstrate resistance to the worst conditions that can be expected in a transportation accident. To simulate these hypothetical accident conditions, the NRC has specified a series of impact, thermal, and immersion tests that must be performed in a specified sequence. Acceptable packaging performance can be demonstrated by analysis, by testing, or a combination of both. In either case, the most damaging orientation for the packaging must be considered for each accident condition. In other words, the tests must be directed at the weakest part of the package. The hypothetical accident conditions and the sequence in which the tests are to be performed are as follows: 1 ) Free drop. A drop from a height of 30 ft onto a flat, unyielding surface In a position for which maximum damage is expected. 2) Puncture. A drop from a height of 40 inches onto a metal bar that is 6 inches in diameter and no less than 8 inches long and is mounted on an L-6 unyielding surface. This test is also to be performed in a position for which maximum damage is expected. (The DOE conducted the tests with a puncture bar that was 24 to 48 inches long, depending on the orientation of the TRUPACT-II.) 3) Heat. Exposure to a surrounding heat flux with a minimum temperature of 1475°F for 30 minutes. (The TRUPACT-II test units were exposed to a fully engulfing fire to meet and exceed these requirements.) 4) Immersion. Exposure to an external pressure equivalent to immersion under at least 50 ft of water for no less than 8 hours. On completion of these tests, the packaging must maintain Its containment integrity by passing a leakage-rate test (NRG, 1975). The Order of the Tests . The order of the tests is reasoned to be the order of events threatening the packaging In a real transportation accident: impact and puncture followed by exposure to fire. The test sequence, therefore, starts with mechanical impacts and then continues with the fire test; this sequence is designed to inflict maximum heat damage. The mechanical and heat tests are applied to the same specimen. The immersion test may be conducted on a separate specimen, because immersion in water is not likely to occur together with an impact accident (IAEA, 1987). The Free-Drop Test Target . The free-drop test requires the package to strike an unyielding flat target after a free drop from a height of 30 ft, striking the target in a position for which maximum damage is expected. With an unyielding target all of the deformation produced by the test Is transferred to the packaging. An actual accident would usually involve a target that yields somewhat, allowing much of the impact energy to be absorbed by the deformation of the target. Thus, an unyielding target forces the packaging to sustain more damage In a given set of test conditions than would a yielding target. Unyielding targets are specially constructed to have a mass at least 10 times the mass of the package being tested. They are usually made of concrete and steel, and the concrete is often tied to bedrock through a system of steel columns, making the target very stiff or essentially immovable. The surface of the unyielding target is a steel plate that is in intimate contact with the surface of the concrete. Tests have shown that the damage created by realistic hard targets, such as rock outcroppings or bridge abutments, would require velocities on the order of 80 miles per hour (mph) in order to be equivalent to the 30-ft drop (30 mph) on the unyielding target. For softer targets, such as other vehicles, concrete pavements, retaining walls, and earth embankments, the velocity required to produce equivalent damage exceeds 200 mph (Jefferson, 1983). The difference between a yielding and an unyielding target can be seen in the results of two drop tests conducted for the DOE in a previous testing program. Two packagings of the same design were tested at Sandia National Laboratories. One packaging was dropped from a height of 30 ft onto an unyielding target. The second L-7 packaging was subjected to a test not required by the NRC regulations: it was dropped from a helicopter from a height of approximately 2,000 ft onto hard desert soil. This 6,700-lb package reached a terminal velocity of approximately 246 mph and was embedded in a crater approximately 8 ft deep in the desert soil. The packaging suffered no permanent deformation. The 30-ft drop onto an unyielding target caused more damage to the packaging than the 2,000-ft drop onto hard desert soil (McClure et al., 1987). (The packagings in these tests were not TRUPACT-II containers.) The Puncture Test . Puncture loads can be expected in accidents because the surfaces that may be hit by a packaging are not always flat. The puncture tests are conducted to demonstrate the integrity of the containment even when weak points (e.g., container seals) are struck. Puncture loads can also produce a loss of the thermal insulation that protects against fires by tearing a hole in the wall of the packaging. In the puncture test, the packaging is dropped from a height of 40 inches in a position for which maximum damage is expected. The target is the upper end of a vertical steel cylinder that is 6 inches in diameter and of a length that would cause maximum damage to the packaging. This puncture bar must be mounted on an essentially unyielding horizontal surface. The areas exposed to the puncture bar tests are subsequently exposed to the fire test (IAEA, 1987). The Fullv Enqulfino Fire Test . The effects of fire on a shipping container depend on the time, the temperature, and the surface exposed. The NRC regulations require exposure to a temperature of 1 475 ° F for 30 minutes over the entire surface of the packaging. In order to have the entire surface exposed to the fire, the packaging must be suspended approximately 4 ft above the fire surface (i.e., a burning fuel pool). The orientation of the packaging above the fuel pool is designed to provide exposure to the highest temperature. Elevating the packaging ensures that the flames are well developed at the location of the packaging, with adequate space for the lateral in-flow of air. This total surface exposure requirement encompasses such events as burning with a torch that is directed at one portion of the task. Since under most accident conditions the heavy packaging would end up on the bottom of the debris, the actual accident conditions would not duplicate the total surface exposure of the regulatory fire test (IAEA, 1987; Jefferson, 1983). Some fires experienced in actual accident conditions burn longer than 30 minutes, but they either burn at lower temperatures (consuming slower burning materials like wood) or are concentrated over small areas, thus being insufficiently large to envelop the entire packaging. An accident that would produce a heat environment exceeding that called for in the regulations is extremely unlikely (Jefferson, 1983). The Immersion Test . As a result of a potential for transportation accidents near or on a body of water, a packaging could be subjected to an external pressure from submersion under water. To simulate the equivalent damage from this low-probability event, the NRC regulations require that a packaging be able to withstand the external pressures resulting from submersion at reasonable depths. Engineering estimates indicate that water depths near most bridges, roadways, or harbors would be less than 50 ft. Consequently, 50 ft was selected as the immersion depth. While immersion at depths greater than 50 ft is possible, this value was selected to envelop the equivalent L-8 damage from most transportation accidents. In addition, the potential consequences of a significant release of radioactive material would be greatest near a coast or in a shallow body of water. The time of exposure was set at 8 hours, which is time enough to allow the package to come to a steady state from the rate-dependent effects of immersion (IAEA, 1987). Since the main purpose of the immersion test Is to demonstrate that a packaging can maintain Its structural integrity when subjected to an external pressure, a pressure test or calculation may be substituted for the actual immersion. The Leakage-Rate Test . After these accident condition tests, a very stringent leakage- rate specification must be met by the packaging. In order to demonstrate that there will be no release of contents under normal accident conditions, both containment vessels must remain leaktight, in accordance with standard ANSI 14.5-1987 of the American National Standards Institute. The stringency of the postaccident-leaktightness standard requires the packaging design to be so robust that it would have to be subjected to an accident much more severe than those simulated in the certification tests before a release of its contents could occur. L2.3 COMPLIANCE OF THE TRUPACT-II PACKAGE WITH NRC REGULATIONS On March 3, 1989, the developer of the TRUPACT-II shipping container submitted to the NRC, on behalf of the DOE, the documentation required for an application for certification. This documentation consisted of a comprehensive safety analysis report for the TRUPACT-II shipping container (DOE, 1989a, Rev. 2) and a document desalbing the codes used in the preparation and characterization of CH TRU waste. Four revisions to the Safety Analysis Report were made to supplement the document with additional information requested by the NRC and the final results of TRUPACT-II tests. The Safety Analysis Report provides a detailed description of the TRUPACT-II design, operation, maintenance, the payload (CH TRU waste) and quality assurance programs. In addition, the report documents the performance of the TRUPACT-II container in the regulatory tests described above. The manner in which the tests were conducted and the results are discussed below. Compliance with the evaluation requirements of 1 CFR Part 71 was demonstrated by a combination of analyses and testing of the TRUPACT-ll package. The certificate of compliance was issued by the NRC on August 30, 1989. It is reproduced in full in Annex 1 to this appendix. L.2.3.1 Evaluation of Performance As reported in Section 2.6 of the Safety Analysis Report for the TRUPACT-II Shipping Package (DOE, 1989a), the container meets the performance requirements of Subpart E of 10 CFR Part 71 for normal transportation conditions. The compliance was demonstrated through analysis and by performing the required free-drop test from a height of 3 ft. The analyses covered the response of TRUPACT-II components to heat and cold, reduced and increased external pressures, and vibration. Exposures to a L-9 water spray simulating a heavy rainfall and impact by a steel cylinder 1-1/4 inches in diameter (penetration test) were judged to be of negligible consequence because of the TRUPACT-II construction. For the hypothetical accident conditions specified in Subpart F of 10 CFR Part 71, tests with full-scale TRUPACT-II units were conducted. The only exception was the Immersion criterion, for which compliance was demonstrated by analysis, as allowed by the NRC. The tests were first conducted with an engineering prototype container. The results from these tests were used to develop design enhancements for the container. For example, a thin ceramic-fiber paper was added as a liner to the polyurethane foam cavity of the outer containment assembly to provide additional protection from fire. Subsequently, three full-scale certification units were tested during the period from December 1988 to April 1989. The testing was performed at Sandia National Laboratories, Albuquerque, New Mexico. Before being tested, all four full-scale TRUPACT-II containers were loaded with 7,265 lb (maximum allowable payload weight) of concrete in 1 4 drums. The full-scale tests consisted of free drops from a height of 30 ft followed by free drops of 40 inches onto a 6-inch-diameter puncture bar. After undergoing multiple free drops and puncture-bar impacts, the prototype and two certification packages were suspended over a pool containing approximately 8,000 gal of jet fuel, which burned for more than 30 minutes. The external skin temperature exceeded 1475°F during the fire. Because of the excellent thermal properties of the package, the maximum 0-ring seal temperature (on either the inner or the outer containment vessel) reached only 260 'F, well below allowable temperatures for the seal materials used. Also, it was found that at least 5 inches of the original 1 0-inch-thick polyurethane foam in the outer containment assembly remained unaffected after the fire test, further demonstrating the safety margins that have been built into the TRUPACT-II shipping container. As shown in Table L.2.1, the number of drop and puncture tests performed on each test unit exceeded the regulatory requirements in many cases; this was done to confirm that the package could sustain impacts in a variety of "worst-case" orientations and remain leaktight. For example, each of the 30-ft drops on test units 1 and 2 were performed with different sections of the TRUPACT-II container package striking the unyielding target (i.e., tiedown locations on the bottom, top knuckle of the head, etc.). The full-scale testing of the test units under the hypothetical accident conditions was conducted with the first certification test unit at the ambient temperature of Albuquerque, New Mexico, in December 1988 (40 to 70° F). The second and third certification test units were chilled to -20° F before the first drops and again before the final leakage- rate tests to prove the ability of the 0-rings to function properly at low temperatures. The leakage rate of the containment seals was tested before, during, and after the test sequence on each test unit. On the first and the third test units, both the inner and the outer containment vessels were demonstrated to be leaktight. On the second test unit, the outer vessel met the criteria for leaktightness as stated in ANSI 14.5-1987 but the inner vessel did not meet this criteria, because debris resulting from the tests interfered with the upper seal of the inner vessel. A wiper 0-ring was added to the inner L-10 TABLE L.2.1 Regulatory testing requirements and the actual TRUPACT-II certification testing program Number of tests performed Test Required number of tests® Unit 1 Unit 2 Unit 3 30-ft drop 1 3 3 3 40-inch puncture drop 1 5 6 5 Fire test 1 1 1 Immersion 1 By analysis'' By analysis'' By analysis'^ ® From 1 CFR 71 .73; requirements can be met by test or analysis. ^ Same analysis was applicable to all three test units. containment vessel on the third test unit, and its effectiveness was demonstrated by repeating the drop-test sequence. It is important to mention that had the payload been TRU waste during the testing of these three test units, no release of contents to the outside environment would have occurred because all of the test units remained leaktight to the outside. L2.3.2 Fabrication Controls Each step in the fabrication of the TRUPACT-II containers is controlled to ensure that the containers are built to the standards and specifications of the test units used for certifying the design of the package. For example, the stainless steel that Is used for the pressure vessels is traceable to the mill, including the pouring and rolling of the steel. This traceability includes test reports on the chemical and physical properties of the steel. When the steel is received at the TRUPACT Assembly Facility in Carlsbad, New Mexico, It is inspected, and each piece of steel is assigned a unique identification that stays with that piece of steel through machining, welding, and final assembly. This means that the components of any TRUPACT-II can be traced back to their origins. Every machining operation is inspected to verify that the part Is made to the drawing requirements from which it was designed. Welding during fabrication is done in accordance with the applicable standards of the American Society of Mechanical Engineers. Welds are nondestructively examined to ensure that there are no defects. L-11 Containment boundary welds are examined by x-ray. Welding procedures and welder qualifications (welders must be certified) will be available for audit or review. After welding and machining, the finished pressure vessel is proof-tested at 150 percent of its design pressure (50 lb per square inch) and then examined once again, using a liquid-dye penetrant. (A liquid-dye penetrant is used to detect cracks that cannot be seen with the naked eye.) Finally, each pressure vessel is tested to the "leaktight" criteria. The leaktightness of the containment boundary is tested on each unit before delivery. In addition to possible failures of the 0-ring seals, this procedure inspects for leaks in the weld zones and cracks in the vessels. L.2.3.3 Operating Procedures L.2.3.3.1 Pavload Controls and Restrictions . The initial certificate of compliance issued by the NRC (Annex 1 to this appendix) defines the allowable pay load (waste materials) that can be transported. Certification of the TRUPACT-II package requires that the payload be controlled to ensure safe transportation. Each waste container to be transported in the TRUPACT-II shipping container must comply with specific transportation requirements for physical form, the composition and radioactivity of the waste, the chemical compatibility of the waste, and the like. Unique identification codes for each waste container provide a system for tracking the process and packaging history of the waste. This information (along with process controls on waste generation procedures) provides the basis for evaluating the qualification of the waste as payload for the TRUPACT-II. The payload restrictions are described below. Strict controls will be used at the waste generation and storage facilities to determine the compliance of a given waste package with the transportation requirements. If a package does not meet any of the limits, it cannot be a part of the payload. The Safety Analysis Report for the TRUPACT-II Shipping Package (DOE, 1989a) and supporting documents describe in detail the basis for evaluating the safety of the payload. The Waste Acceptance Criteria Certification Committee (WACCC) has been identified to the NRC as the DOE's verification organization. The WACCC will ensure payload compliance with the TRUPACT-II certificate of compliance. To verify payload compliance, the WACCC intends to use a process similar to that used for verifying compliance with the WIPP Waste Acceptance Criteria. Therefore, each shipping facility will be required to submit a TRUPACT-II payload compliance plan and an associated quality assurance plan to the WACCC for review and approval. Detailed compliance procedures will be developed and implemented, and their implementation will be audited by the WACCC. The individual responsible for every TRUPACT-II shipment from a given facility is the Site Certification Official. This person will ensure that the waste containers in a TRUPACT-II shipping container and the total payload are in compliance with all certification and transportation requirements. (See Appendix A for a description of the Waste Acceptance Criteria and their relationship to transportation requirements.) L-12 Physical Form . The physical form of the TRUPACT-II payload is restricted to solid or solidified material. Examples of solid materials are paper, glass, and metals. Examples of solidified materials are cemented sludges. Liquid waste is prohibited in the payload containers except for residual amounts. Sharp objects that might affect the integrity of the payload containers are prohibited unless they are adequately packaged to prevent damage to the payload containers. Sealed containers are prohibited from being included as a part of the waste, except in volumes of 1 gal or less. These restrictions on the physical form of the waste are met during the generation of the waste. Verification procedures like visual examination, x-ray examination, and sampling of previously packaged containers are routinely used as some of the additional controls. Chemical Form and Chemical Properties . The following classes of materials are prohibited from the TRUPACT-II payload unless they have been destroyed, neutralized, or otherwise rendered safe: • Compressed gases • Explosive materials • Nonradioactive pyrophorics • Corrosive materials In addition, there are restrictions on specific chemicals and materials that can be present within each waste form. These restrictions on the chemical constituents of the waste are needed in order to limit the amount of gases (flammable as well as nonflammable) that might be generated from materials in the waste on exposure to radiation. Compliance with these requirements will be achieved through process controls at the waste generator and disposal facilities, including procurement and inventory controls. For example, in the course of being generated, waste will be subjected to neutralization and solidification to remove any corrosives that may be present in the waste. Process-flow analyses yield information on the chemical constituents of each waste form. Chemical Compatibility . The composition of the waste must preclude adverse chemical processes during transport that might pose a threat to the payload. Specifically, it is necessary to establish the following: 1) The chemical compatibility of the waste form within each individual container of waste. 2) Chemical compatibility between waste containers under hypothetical accident conditions. In analyzing the consequences of hypothetical accidents, no credit is taken for the structural integrity of the individual waste containers. All the waste containers are assumed to be breached, and the contents from all the individual waste containers are assumed to mix together. The contents of a waste container (drum or standard waste L-13 box) must be compatible, and the contents of different waste containers in the TRUPACT-II must also be compatible. 3) Chemical compatibility of the waste forms with the inner containment vessel of the TRUPACT-II. 4) Chemical compatibility of the waste forms with the 0-ring seals of the TRUPACT-II. Each waste form to be transported in the TRUPACT-II shipping container is analyzed for the above compatibility criteria, using a method proposed by the U.S. Environmental Protection Agency (Hatayama et a!., 1980). Only compatible waste forms will be part of the TRUPACT-II payload. This will ensure that chemicals that might affect the performance of the inner containment vessel or the 0-ring seals are not released in any significant amounts into the inner containment vessel during transport. In addition, this will ensure that no adverse chemical reactions will take place within the waste containers or between the waste containers under accident conditions. Sampling programs conducted at the waste generating or disposal facilities provide additional verification for the chemical compatibility analyses. Operating Pressure and Gas Generation . The acceptable maximum operating pressure in the TRUPACT-II cavity is 50 lb per square inch (gauge). The payload is limited in order not to exceed this design pressure. In addition, the generation of gas from the waste (which could occur primarily through the exposure of certain materials to radiation) is controlled to prevent the occurrence of potentially flammable concentrations of gases in the payload or the shipping containers. Gas generation is controlled by limiting the radioactivity of the waste and by restricting the constituents 'n the waste that may release gases on exposure to radiation. Decav Heat and Fissile Materials . Decay-heat limits are imposed on each waste container, as well as on the total TRUPACT-II payload, to keep the potential quantity of gases generated below safe limits. In addition, the quantities of fissile materials in the waste containers and the total payload are restricted, so as to remain below the limits established by the NRC to prevent nuclear criticality under all conditions. Waste Containers . Two types of waste containers can be shipped in the TRUPACT-II shipping containers: 55-gal drums and standard waste boxes. The latter are large steel vessels that are designed to fit in the TRUPACT-II cavity (see Appendix D). A payload consists of either 1 4 drums or 2 boxes. The containers must be provided with vents equipped with high-efficiency carbon composite filters that allow gases to be released from the containers while retaining particulates. The main purpose of restrictions on the waste containers is to prevent the buildup of gases within the waste containers. Verification of compliance with these requirements Includes controls on waste generation procedures, visual inspection, records and data bases, and sampling programs. L-14 Weight . Weight limits apply to individual waste containers and to the total payload and are as follows: Container Weight limit (lb) Drum Standard waste box TRUPACT-II shipping container 1,000 4,000 7,265 Radiation-Dose Rates . The radiation-dose rates on the external surfaces of individual waste containers and the three loaded TRUPACT-II containers to be transported on a trailer will be 200 millirem per hr or less at the surface and 1 millirem per hr or less at a distance of 2 meters from the surface, In accordance with 1 CFR 71 .47. L.2.3.3.2 Procedures for Loading and Assembling TRUPACT-II Shipping Containers . Assembling a TRUPACT-II shipment will involve three steps: 1) preparing each of the waste containers (14 drums or 2 standard waste boxes) in accordance with the specifications in the payload-control procedures (Subsection L.2.3.3.1), 2) loading the waste container into the TRUPACT-II cavity, and 3) testing the leaktightness of the seals on the outer and inner containment vessels of the TRUPACT-II shipping containers. Specific instructions for operating the TRUPACT-II container will be given to each facility to ensure that the shipping container is loaded and sealed properly. Once the lids of the outer and the inner containment vessels are removed, the payload is lifted into the cavity of the inner vessel. Specially designed lifting devices will be provided to prevent damage to the inner vessel or the outer containment assembly during loading. Before the lid of the inner vessel is installed, the seals and other components must be visually inspected for damage that could impair their function. If function- impairing damage is present, the damaged components are replaced before further use. Once these steps are completed, the inner vessel is ready to be assembled. This is done by positioning the lid above the body and lowering it into position. The lid is then drawn downward to its fully engaged position. Once the lid is fully engaged, the lockring is rotated, thus engaging the locking lugs and locking the lid in place. Lock bolts are then installed to prevent rotation of the lockring. An assembly-verification leaktightness test is then performed to ensure that the 0-ring seals were properly installed and not damaged during assembly. This assembly procedure ensures containment integrity for the following reasons: 1) The mating surfaces between the body and head of both the inner and outer containment vessels are designed like a double tongue-and-groove joint. The head and body are connected by rotating a lockring, attached to the head, that has tabs that mate with corresponding tabs on the body. If the head and the body are not assembled correctly, it will be impossible to rotate the lockring. Ability to rotate the lockring is one verification that the head-to-body connection is properly assembled. L-15 2) The containment boundary seal is made by an elastomer 0-ring that is located at the head-to-body Interface and is part of the tongue-and-groove joint. There is also a test 0-ring and a wiper 0-ring (on the inner vessel only). When properly assembled, the 0-rings are captured between the head and body. Each time the head is installed on the body, it is necessary to perform a leak test to verify that the 0-rings are in place and that they were not damaged during assembly. Once the lid of the inner containment vessel is properly installed, the outer vessel can be assembled. This is done in the sequence used for the inner vessel, the only difference being that the lockring is rotated and held in position by means of a mechanical actuator ring. In the locked position, lock bolts hold the actuator ring in position, which, in turn, holds the lockring in position. As in the case of the inner vessel, an assembly-verification leaktightness test is required. L.2.3.3.3 TRUPACT-II Transport Trailer . The TRUPACT-II transport trailer is of a gooseneck, dropped bed design which is commonly used in commercial fleet operations. The design has been adapted for the transportation of up to three fully loaded TRUPACT-II shipping packages. The TRUPACT-II transport trailer is 42.2 ft in length, the load bearing bed is 40 inches aboveground and when loaded with TRUPACT-lls, the overall height is 161.5 inches. Each trailer is provided with 12 each, special tiedown devices used for securing the TRUPACT-II packagings in a vertical position to the trailer. The tiedowns are cam operated, adjustable length U-bolts that interface with, and clamp down on corresponding brackets on the TRUPACT-II packaging. The tiedown restraint applied to the TRUPACT-II packages has been designed to satisfy the tiedown requirements of the DOT, 49 CFR 393.102, and the NRC requirement, 10 CFR 71.45. The Safety Analysis Report for the TRUPACT-II Shipping Package given to the NRC in March 1989 provides the necessary analyses for showing how the TRUPACT-II tiedown system meets these requirements. The trailer has been through a series of tests which demonstrated it can be safely used without restrictions on the nation's highways. L.2.3.4 Maintenance A detailed maintenance program has been established by the DOE and approved by the NRC for the TRUPACT-II containers. Maintenance procedures include scheduled inspections and replacement of components, structural and pressure tests, and leaktightness tests for maintenance verification (0-ring seals, vent-port plug seals, etc.). The maintenance procedures are described briefly below. Structural and Pressure Tests . A structural pressure test must be performed on the inner and the outer containment vessel once every 5 years. This involves pressure testing each vessel to 1 50 percent of the maximum normal operating pressure. Leaktightness Tests . Maintenance-verification leaktightness tests must be performed for the main 0-ring seals and for each vent-port plug seal annually or on seal replacement. L-16 Maintenance of Components . Maintenance is specified for certain components, such as fasteners, lockrings, and seal areas and grooves. The threaded parts of fasteners are to be annually inspected for deformed or stripped threads. Visual inspections are required before every use for the lockring bolts (inner containment vessel and outer containment assembly), the vent-port plugs, and the seal-test port. Any damaged parts must be replaced before further use. The lockring of the inner vessel and the locking actuator of the outer containment assembly are to be inspected before every use for any motion-impairing components. Corrective actions are to be taken whenever necessary. Before each use, and at the time of seal replacement, sealing surfaces and 0-ring seal grooves are to be visually inspected for any damage. An annual inspection of the dimensions and surface finishes of the 0-ring seal area is also required. The required measurements include groove widths, tab widths, axial play, and the surface finish of sealing areas. Maintenance, repairs performed, or components replaced will be documented on the TRUPACT-II Maintenance Record Form WP-1709 (DOE/WIPP 88-026). All records of maintenance activities performed on the TRUPACT-II container will be maintained by WIPP Operations for retention and distribution. The records will be designated as quality assurance records and will be maintained as permanent records. All replacement components procured by user facilities will be verified for compliance with applicable material requirements. The DOE shipping and receiving facilities that perform maintenance on TRUPACT-II containers will have in place a quality assurance program that meets the applicable requirements of the DOE (see Section L.4). L-17 L3 THE NUPAC 72B CASK PROGRAM L3.1 BACKGROUND To transport remotely handled (RH) TRU waste, the DOE will use the NuPac 72B shipping cask. The NuPac 72B cask is being designed to meet NRC requirements for Type B packages, and the DOE will apply to the NRC for a certificate of compliance before transporting any waste in the 72B cask. The 72B cask is a scaled-down version of the NuPac 1 25B cask, whose design has been certified by the NRC as a Type B packaging. The 125B cask is being used to transport debris from the core of the damaged Three Mile Island reactor. L3.2 DESCRIPTION OF THE NUPAC 728 SHIPPING CASK The NuPac 72B cask is a cylindrical cask consisting of a separate inner vessel within an outer cask protected by impact limiters at each end. A schematic is shown in Figure L3.1. The outer cask provides the primary containment boundary for the payload, while the inner vessel provides a secondary containment boundary. Neither containment vessel (the outer cask nor the inner vessel) is vented, and each is capable of withstanding an internal pressure of 150 lb per square inch (gauge). The capacity of each cask is 8,000 lb of payload. The payload consists of RH TRU waste in 30- or 55-gal drums contained in a canister. The 72B cask is designed to transport a single canister per shipment. A single 72B cask will be loaded onto a custom- designed semitrailer pulled by a conventional tractor. The inner containment vessel is made of stainless steel and provides a cavity for the payload canister that is approximately 26.5 inches in diameter and 123 inches long. The lid is secured to the body of the vessel by means of eight closure bolts. Internal spacers are provided at the top, bottom, and at two locations near the middle of the inner vessel to center the canister and facilitate the insertion and removal of the canister. The outer cask is a stainless-steel vessel constructed of two concentric shells enclosing a cast-lead shield. The shield is for gamma radiation and is approximately 1 .9 inches thick. The outer cask is approximately 1 42 inches long and has an outer diameter of 42 inches. It is protected at each end by energy-absorbing impact limiters, which are stainless-steel shells filled with polyurethane foam. The impact limiters also act as thermal insulators to protect seal areas from fire during an accident. The payload canister, or RH waste canister, is a DOT 7A Type A carbon steel single shell container measuring approximately 26 inches in diameter with an overall length of 121 Inches. The canister is vented using a carbon composite HEPA filter and is capable of transporting three 55-gallon waste drums. The allowable gross weight of the canister and contents is 8,000 pounds. L-18 O I (M CO J UJ cc O CO < o (N N o < 0. D Z LL O z o o UJ en (/} (/) o o L-19 L3.3 COMPLIANCE WITH NRC REQUIREMENTS In order for the design of the NuPac 72B cask to be certified by the NRC, it will be necessary to demonstrate compliance with the NRC requirements in 10 CFR Part 71 for Type B packages (see Subsection L2.2). Compliance with these requirements may be demonstrated by analysis or by a conhbination of analysis and testing. Since the 728 cask is a scaled-down version of the 1 258 cask, whose design has been certified by the NRC, analysis will be the primary method of demonstrating compliance with the NRC regulations. L3.4 OPERATING PROCEDURES L3.4.1 Pavload Controls and Restrictions As in the case of the TRUPACT-II shipping container, the NRC's certificate of compliance for the 728 cask will specify the allowable payload. The restrictions on the payload will be similar to those discussed in Subsection L.2.3.3.1 for CH TRU waste. Physical and Chemical Form . The restrictions on the physical and chemical form of the payload to be carried by the 728 cask and the necessary payload controls are expected to be similar to those specified for the CH TRU waste in the TRUPACT-II payload. These restrictions are described in Subsection L.2.3.3.1 of this appendix. Chemical Compatibility . The payload for the 728 cask will be evaluated to ensure chemical compatibility within itself and with the cask. The criteria for evaluating and ensuring chemical compatibility are discussed in Subsection L.2.3.3.1. Operating Pressure and Gas Generation . The pressure in both containment levels of the cask is 150 lb per square inch (gauge). The payload is restricted in order to not exceed this design pressure. The generation of gas from the waste is controlled to prevent the occurrence of potentially flammable concentrations of gases. Weight . The maximum weight of the loaded canister in the 728 cask is limited to 8,000 lb. The cask may carry no more than one canister of RH TRU waste. Decay Heat . The thermal design rating of the package is 300 watts internal decay heat maximum. Radiation-Dose Rates . The radiation-dose rates on the external surface of the 728 cask will be below the levels specified in 1 CFR 71 .47 and must comply with 49 CFR 173.441. L.3.4.2 Procedures for Loading the NuPac 72B Cask Loading a 728 cask for transport will consist of the following steps: 1) determining that the payload (the canisters of RH TRU waste) has been verified to meet the payload restrictions specified in the certificate of compliance, 2) loading the prepared L-20 payload canister into the 72B cask, 3) testing the leaktightness of the seals on the containment vessels of the cask, and 4) securing external impact limiters on the cask. Specific procedures for operating the 72B cask will be provided to each waste generating or storage facility to ensure that the cask is loaded and sealed properly. The loading procedures include removing the lids from the containment vessels, loading the waste canister into the vessel, installing the lids, and performing the leaktightness tests. L.3.5 MAINTENANCE OF THE NUPAC 72B CASK As in the case of the TRUPACT-II shipping container, a strict maintenance program will be developed and implemented for the 72B cask. The procedures will be submitted to the NRC as part of the Safety Analysis Report (SAR). The NRG must approve these procedures before the design of the NuPac 728 cask is certified and the cask can be used to transport waste. The maintenance program will include periodic inspections and replacement of components, structural and pressure tests, leaktightness tests, and routine maintenance of all necessary parts of the cask. A comprehensive quality assurance program will also be developed, as discussed in Section L.4. L-21 L4 QUALITY ASSURANCE PROGRAM The NRC regulations in 10 CFR Part 71 include requirements for implementing a quality assurance program that is used in the design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of those components of the TRUPACT-II container and NuPac 72B cask that are important to safety. The quality assurance requirements are not optional; they are mandatory. The quality assurance program provides a systematic approach to ensuring that a design, and the resulting product or service, are safe and satisfactory for the intended use. The program is aimed at preventing problems, not only at detecting and solving them. The quality assurance program is developed and implemented by specially trained full- time employees. They report to the highest level of management in their organizations in order to maintain their independence from concerns about costs or schedules. Their primary function is to make sure that the quality assurance program meets the requirements of the NRC and is effective In producing a product that meets required standards and that will maintain its integrity during operation. This requires ascertaining that all workers are trained and qualified to perform their assigned tasks, all workers are trained to understand the program, and work is properly controlled. Design Control . Quality assurance begins with the design of an item or the description of a service. Large safety margins are established for each item (i.e., if a TRUPACT-II will be operating at a pressure of 50 lb per square inch, it is designed to be strong enough for a pressure of 75 lb per square inch). All of the mathematical calculations and analyses used in making design decisions are reviewed and verified by independent qualified personnel. Procurement Control . Quality assurance requires that the materials used in constructing a shipping container or cask be tested, both chemically and physically, to make sure that they have the properties needed for the TRUPACT-II or NuPac 72B design. Further, the suppliers who manufacture the materials are evaluated to ensure that they have an acceptable program for ensuring that the materials they are furnishing are properly analyzed, chemically and physically, and that the analysis reports match the material shipped. Marking and Control of Materials . Once the material arrives, it is inspected by a quality inspector and stored properly for use. The material is placed in an environment that will not damage it and marked or tagged so that its identity is not lost. The materials used in the TRUPACT-II container or the NuPac 72B cask must be traceable from the production unit in which it is used, back to the purchase order used to buy it and the material test report verifying that the material is suitable. Thus, if a problem L-22 arises in a particular batch of material, the company must identify every production unit in which the material was used. Instructions. Procedures, and Inspection . Work on the TRUPACT-II or the NuPac 72B units is performed in accordance with formal instructions, procedures, or drawings that have been reviewed by engineering and quality assurance personnel. Part of this formal system for controlling the work includes setting points during the fabrication for inspection. If one of these predetermined points is ignored and the inspection cannot be performed at a later time, the unit faces rework. These inspection points are a part of every work plan and ensure that the final unit is acceptable. Those same similar instructions, procedures, and drawings are later used to perform preventive maintenance during the operation of the TRUPACT-II container or the NuPac 72B cask. Control of Processes . Some types of processes require more control than others because special techniques like x-ray examination are needed to determine that they were performed properly. An example of such a process Is welding. The quality assurance program makes special provisions for such processes and for ensuring that the special inspection techniques required for these processes are used successfully. These special provisions include testing the skills of the personnel performing the processes, qualifying the procedure being used, and verifying that the materials and equipment for the process are appropriate. In addition, quality assurance personnel perform in-process inspections to make sure that the controls are being used during the actual work. Records of these activities are kept. Test Control . Any type of testing requires very tight control and careful monitoring by quality assurance personnel. For example, pressure and leaktightness tests on the containment vessels of the TRUPACT-II container are performed in accordance with formal procedures that have been reviewed by both engineering and quality assurance personnel. Tests are witnessed by quality assurance personnel, and test results are formally documented and reviewed for adequacy. Any reworking on the containment boundary of a TRUPACT-II unit requires previous tests to be performed again. Control of Measuring and Test Equipment . Results from Inspections and tests are only as good as the equipment used to measure the results. The quality assurance program requires that the equipment used to measure or test a TRUPACT-II shipping container be calibrated. This means that all measuring and test equipment has to be checked against a national standard for the particular measurement being taken and has to be accurate within a given range. Not only does the equipment have to be checked and adjusted if necessary, it also has to be rechecked periodically. If a piece of equipment is found not to agree with the national standard, the manufacturer has to evaluate each item that was inspected or tested with that piece of equipment. Acceptability of Components. The acceptability of parts of a TRUPACT-II container or a NuPac 72B cask must be apparent at all stages of fabrication. The quality assurance program provides a method of doing this by using inspection hold points, tagging, etc. If an item is found to be unacceptable, the quality assurance personnel document the problem on what is called a nonconformance report. The item is then marked or tagged and segregated from the rest of production until a decision Is reached on what to do with the item. This decision is made by engineering and L-23 quality assurance personnel. Sometimes an item can be reworked and made acceptable; sometimes an item must be scrapped. The provisions of the quality assurance program, however, prevent unacceptable items from being unintentionally used in the production process and provide a method for deciding how to handle unacceptable items. Surveillance . In addition to inspections, quality assurance personnel perform scheduled and unscheduled surveillance of various activities to make sure that employees are operating to the same rules and are performing their jobs well. The activities selected for surveillance are those in progress that are most important to the operation at the time. Corrective Action . The quality assurance program specifies a method for identifying recurring problems and serious problems that might affect the performance of the product. A formal report, called a "corrective action report," is issued by quality assurance personnel when such problems surface. This report must be answered by production or engineering personnel and must include an explanation of what is causing the problem, a description of what is being done to correct the problem, and a description of what is being done to keep it from happening again. Quality assurance then makes sure that the proper actions have been completed and that they are, in fact, solving the problem. These reports are reviewed by the highest level of management, who make sure that all departments respond quickly. Document Control . The different parts of the quality assurance program are formally documented to make sure that personnel understand the rules and controls that are necessary to produce a good product. These documents are themselves controlled to make sure that all personnel are working to the same guidelines and that only the latest documents are in use. If a document is changed, the old document must be returned or destroyed and personnel must be trained to ensure that they understand the new rules. This is true of every document that affects work, including work plans, procedures and drawings, and inspection plans. Quality Assurance Records . The final step before releasing a TRUPACT-II or NuPac 72B unit for use is the review of related quality records. These records tell the production story of a unit. They start with the pedigree of the materials used and proceed through fabrication, inspection, and testing to final acceptance. This final review by quality assurance ensures that the records are complete, inspections have been performed, and the requirements have been met. This same record package, which is several hundred pages, is then retained in duplicate in protected storage for the life of the TRUPACT-II or NuPac 72B unit. Audits . An important mechanism for ascertaining that the quality assurance program is correctly implemented is the audit. Quality assurance personnel audit their facility and operations to see whether all the established rules and regulations are complied with. If deficiencies are found, they are documented, corrected, and verified as effective. The quality assurance personnel who perform these audits are specially qualified through classroom and on-the-job training to spot problems in the system and get them fixed. Auditors from outside the organization also perform this function. For example, the Westinghouse Electric Corporation (the operating contractor for the L-24 WIPP) audits Nuclear Packaging (the manufacturer of the TRUPACT-II container), and the DOE audits Westinghouse, as well as the waste generator and storage facilities. The NRC has also audited Nuclear Packaging as part of the certification process for the TRUPACT-II design and has the prerogative to audit any activities associated with the use of a TRUPACT-II container. Summary . As overlapping as all of the described quality assurance controls may seem, the checks and balances built into the program are necessary to provide the highest assurance possible that the TRUPACT-II container and the NuPac 72B cask will safely perform its intended function. This program will remain in effect as long as TRUPACT-II or NuPac 72B units are being used. L-25 REFERENCES FOR APPENDIX L DOE (U.S. Department of Energy), 1989a. Safety Analysis Report for the TRUPACT-II Shipping Package . Rey. 2, Doclyk'>>: .►• "'Co, UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C 20555 -51989 AUG 3 1989 SGTB:EPE 71-9218 Department of Energy ATTN: Mr. Edward McCallum DP-4 Washington. DC 20545 Gentlemen: Enclosed is Certificate of Compliance No. 9218, Revision 0, for the Model No. TRUPACT-II shipping container. The Departnxjnt of Energy has been registered as a user of this package under the general license provisions of 49 CFR §173.471. This approval constitutes authority to use this package for shipment of radioactive material and for the package to be shipped in accordance with the provisions of 49 CFR §173.471. Sincerely, Charles E. MacDonald, Chief Transportation Branch Division of Safeguards and Transportation, NMSS Enclosures: 1. Certificate of Compliance No. 9218, Rev. 2. Safety Evaluation Report cc w/encl: Mr. Michael E. Wangler DepartJTient of Transportation Mr. 6. J. Quinn Nuclear Packaging, Inc. Mr. J. Tollison Department of Energy ^^/Hr. T. Halverson Mestinghouse L-29 airtt ! »MM »»» iewiiPi f mMMii«»M»!iiMjjji^m NRC FORM tia 10CFW71 CERTIFICATE OF COMPLIANCE FOR RADIOACTIVE MATERIALS PACKAGES U.S. NUCLEAA REQUIATORY COMMISSION 1 tCEP "m^ NUUeEA REVIStON NUMBEA C PACKAGE lOCMTIFICATtON NUMBER USA/9218/B{ij)F d^ PAGE (4UMB£fl 1 • TOTAL NUMB0I PAGES 2 PREAMBLE ■ Thi* cartlficate « issued lo certify that tti« packaging and contents described In Item 5 below, meets the appltcable ufety standards set forth in Title 10, Code of Federal Regulations. Part 71. "Packaging and Transportation of Radioactive Material." b This ceniticale does not relieve the consignor Irom compliance with any requirement of the regulations of the US. Department of Transporlalion Of other applicable regulatory agerKies. including the government of any country through or into which the package will be trmnaported. 3. THIS CERTIFICATE tS ISSUED ON TME BASIS Of A SAFETT ANALYSIS REPORT OF THE PACKAGE DESKIN OR APFTtCATION a ISSUED TO (Mmw tnd Adctmal b Tm.E AND IDENTIFICATK3N Of REPORT OR APPLICATION. Department of Energy Albuquerque Operations Office P.O. Box 5400 Albuquerque, NM 87115 Nuclear Packaging Inc. application dated March 3. 1989, as supplemented, r- r-v oo«ET,^fc^; f7j-9218 4. CONDITIONS ^ , S.' / This certificate is conditional upon fulfilling the reqtiirements o( 10 CFR Part 71. as applicable, and ihe cbnditions specified below (a) *■• Packaging (1) ModetUo.: ^TROPACT-II ^ (2) V\ Description A stainless steel and polyurethane foaVlnsulated shipping container designed to provide double, cental nment'ijor shipment of^«)ntact-handled transuranic waste. The.packaging fonsists of an.unverrtfid, 1/4-inch thick stainless 3teel inner containment ^esseV (I CV), positioned within an outer containment assembly (OCA) consisting T)f an unvented i/4-inch thick stain- less steel outer containment ;V£5seV(0CV), a >10- inch -thick layer of poly- urethane foam «nd a 1/4 to a/S-fnch thick outer stainless steel shell. The package is a right -circular .xyl.inder with outside dimensions of approximately "94 inches diameter and 122 inches h^l^ht. The package weighs approximately 19,250 pounds vhen , loaded with the maximum allowable contents- of 7,265 pounds. ^ - ^•. The OCA has i domed lid which is secured to the OCA body with a locking ring. The OCV containment seal is. provided by a butyl rubber 0-ring (bore seal). The OCV is equijpped^th ^ seal test port and a vent port. The ICV is dimensions inches hei ring. The (bore seal Aluminum s during shi of approxi j a right circular cylinder with domed ends. The outside of the lev are approximately 73 inches diameter and 98 ght. The ICV lid is secured to the ICV body with a locking ICV containment seal is provided by a butyl rubber 0-ring ). The ICV is equipped with a seal test port and vent port, pacers are placed in the top and bottom domed ends of the ICV pping. The cavity available for the contents is a cylinder mately 73 inches diameter and 75 inches height. L-30 COHD[J}OHS fcontinu9d) Page 2 - Certificate No. 9218 - Revision No. - Docket No. 71-9218 5. (a) Packaging (continued) (3) Drawings i i The packaging is constructed in accordance with Nuclear Packaging Inc. Drawing No. 2077-500 SNP, Sheets 1 through 11, Rev. D. The contents are positioned within the packaging in accordance with Nuclear Packaging Inc. Drawing Nos. 2077-007 SNP, Rev. C. and 2077-008 SNP ""--^- ' --^ " " ^ ^ ^ (b) Contents , Sheets 1 and 2, Rev.rC. D CT /^ (1) Type and form of material ^ Dewatereir;"3olid or solidified transuranic wastes.*^ Wastes must be packaged in "SS-gallon drums, standard waste boxes (SWB), 55-gallon drums Within standard waste boxes, or bin^-^itbin standard waste boxes. Wastes must be restricted to prohibit explosives, corrosives, non- radioactive phrophorics jnd pressurized containers. Within a drum, bin or SWB, radioactive pyrophorics niustjnot exceed 1 percent by weight and free liquids must not exceed.! percent by voluue. Flammable organics are limited lO-fOO ppm in the headspace of any drum, bin or SHB. , - (2) Maximum quantity of material per package Fourteen (14) 55-gallon drums or two (2) SWB and not to exceed 7,265 pounds including shoring and secondary containers with no more than 1000 pounds per 55-gallon drum and 4,000 pounds per SWB. Fissile material not to exceed 325 grams Pu-239 equivalent with no more than 200 grams Pu-239 equivalent per 55-gallon drum or 325 grams Pu-239 equivalent per SWB. Pu-239 equivalent must be determined in accordance with Appendix 1.3.7 of the application. Decay heat not to exceed the values given in Tables 6.1 through 6.3 "TRUPACT-II Content Codes", (TRUCON), OOE/WIPP 89-004, Rev. 3. (c) Fissile Class I Physical form, chemical properties, chemical compatibility, configuration of waste containers and contents, isotopic inventory, fissle content, decay heat, weight and center of gravity, radiation dose rate must be determined and limited in accordance with Appendix 1.3.7 of the application, "TRUPACT-II Authorized Methods for Paylaod Control", (TRAMPAC). i i L-31 llllll»»«KiLML)lLMJ il W WW — I Page 3 - Certificate No. 9218 - Revision No. - Docket No. 71-9218 LJlMMllMMMJBkMMmMJiU l tJ OOmXTtOHi (contwumd) US NUCLEAP REGULATORY COMMISSION 7. 8. 9. 10. Each drum, bin or SWB must be assigned to a shipping category in accordance with Table 5. "TRUPACT-II Content Codes", (TRUCON). DOE/WIPP 89-004. Rev. 3 or must be tested for gas generation and meet the acceptance criteria in accordance with Attachment 2.0, to Appendix 1.3.7 of the application. Each drum, bin or SWB must be Tabled to indicate its shipping category. All drums, bins or SWB's within a package must be of the same shipping category. Each drum, bin or SWB must be equipped with filtered vents prior to shipment in accordance with Appendix 1.3.7 of tj» apCjciJij^S ^'""'"S which were not equipped with filtered vents during s1grw.w*t Etf Cpfoetfed before shipment. The mini- mum aspiration time mustj)fc^(fit«mnined from TableM/ through 9.3 in "TRUPACT-II Content Codes", ( TRUCON Q^flt/W I PP 89-004, Rev. 3. ^-<7 . In addition to the t;^irements of Subpart G of 10 CFR Part/Tlj; (a) Each package/iusty-be ^prepared for shipment and operated Ipaccordance with the procedj/r?s desctibtti in Chapter 7.0, "OperjLt^hg^Procedj^s", of the applicatioiL' v V'^'^x /jT?^ ; ^^-'. • \ „^-^ / f^ ■ i-* " "i package must be tested and nialntatned irirccordance witt'the procedures :ribed:t1n Chapter S.O^jfAcceptance les^TiSa Maintenance ^ri)gram", of the lication. ^:l^\ ■.. r->-- ^ L^^. (b) Each package must be 1:e^ed and liialntafned' ifSccordance witt'the procedures descrf'---'-'^- ru-_*-_ *» « w. ^.'_-- ^-_^ v...... . a ..... The contents of ^ach package 'iwistvbeiiii7accor£ftC€ -with Ap "Payload Control Procedures", of-the ^--''---"^'— ^ . . ' pplica.t^iJ^i^. 14. Prior to each shipment, the lid jiiid veri port^als.on the inner«^- -^^ 15. Expiration date: August 31, 1994. L-32 us NUCLEAO PiECULATORY COMMISSION M; J APPENDIX M SUMMARY OF THE MANAGEMENT PLAN FOR THE TRUCKING CONTRACTOR M-i/ii i TABLE OF CONTENTS Section Page M.1 INTRODUCTION M-1 M.2 SAFETY M-2 M.2.1 Policy M-2 M.2.2 Requirements for Protecting Health and Safety M-2 M.2.3 Safety Program M-3 M.2.4 Occupational Safety M-3 M.3 EQUIPMENT M-4 M.4 EQUIPMENT MAINTENANCE M-7 M.4,1 Maintenance Facility M-7 M.4.2 Maintenance Personnel and Equipment M-7 M.4.3 Maintenance Schedule M-8 M.4.4 Quality Assurance and Control M-1 3 M.4.5 Records M-1 4 M.4.6 Security M-14 M.5 DRIVERS M-16 M.5.1 Driver Qualifications M-16 M.5.2 Driver Training Program M-1 7 M.6 PROCEDURES USED IN WASTE TRANSPORTATION M-1 9 M.6.1 Responsibility for Daily Operations M-1 9 M.6.2 Number of Drivers M-1 9 M.6.3 Security M-19 M.6.4 Procedures to be Followed Before the Start of the Trip M-19 M.6.5 Procedures to be Followed at the WIPP Site M-20 M.6.6 General Procedures to be Followed During the Trip M-22 M.6.7 Constant Surveillance M-22 M.6.8 Inspections During the Trip M-22 M.6.9 Procedures at the Waste Site M-23 M.6.1 Problems During the Trip M-23 M.6.1 1 Delivery of Waste at the WIPP Site M-24 M.6.1 2 After-trip Report M-24 M.6.1 3 Penalties for Drivers M-24 M-lli Section Page M.7 PROCEDURES FOR ACCIDENTS AND INCIDENTS M-25 M.8 SHIPMENT TRACKING AND COMMUNICATIONS M-27 M.8.1 Shipment Tracking M-27 M.8.2 Backup Communications M-27 LIST OF FIGURES Figure M.4.1 M.4.2 M.4.3 M.4.4 M.5.1 M.6.1 Page Example of monthly tractor inspection form M-9 Example of monthly trailer inspection form M-1 Example of annual and semiannual trailer inspection form M-1 1 Example of Dawn Trucking shop ticket M-1 5 Driver qualification form M-1 8 Example of driver's vehicle inspection form M-21 Table M.3.1 M.3.2 M.4.1 LIST OF TABLES Page Specifications for the tractors to be used in hauling TRU waste to the WIPP M-5 Overall dimensions of the tractor-trailer unit M-6 Maintenance schedule for tractors and trailers to be used to transport TRU waste to the WIPP M-8 M-iv M.1 INTRODUCTION This appendix has been prepared in response to comments on the draft SEIS. Representative comments include concerns about the trucking contractor's experience and safety programs, drivers' rights and training, tractor-trailer requirements, and general safety issues. This appendix addresses these concerns by describing the provisions that will be made and the procedures that will be followed to ensure that the transportation of waste to the WIPP is conducted safely. This appendix summarizes the management plan developed by the contractor selected by the U.S. Department of Energy (DOE) for transporting transuranic (TRU) waste to the WIPP. The selected contractor is the Dawn Trucking Company of Farmington, New Mexico. The transportation operations will be conducted by truck, using a fleet of tractors provided by the contractor and trailers and shipping containers provided by the DOE. The contractor will conduct the transportation operations from a facility to be developed at Hobbs, New Mexico. The transportation project will be both managed and coordinated from the Hobbs facility, but management and support personnel at the contractor's offices in Farmington will be available to assist if needed. As described in this appendix, the trucking contractor has developed detailed procedures related to safety, equipment maintenance, quality assurance, driver qualification and training, the duties and responsibilities of drivers, dispatching, the reporting of incidents and accidents, and communications procedures associated with shipment tracking. Many of these procedures are based on the regulations issued by the Department of Transportation (DOT) for the transport of hazardous materials, RCRA (40 CFR Part 263) requirements for the transport of mixed waste, and on the experience of the Federal Government in transporting radioactive materials for several decades, particularly the experience of the DOE in transporting weapons. In reviewing the WIPP program activities, the National Academy of Sciences (NAS) concluded that the "system proposed for transportation of TRU waste to the WIPP is safer than that employed for any other hazardous material in the United States today and will reduce risk to very low levels." The DOE and the trucking contractor have tried in this plan to reduce as much as possible the potential for human error or mechanical failure. Extensive driver-training requirements, dry-run readiness experience (see Appendix D.2.3.2), emphasis on safety, inspections that exceed many DOT regulatory requirements, and the use of tractor- trailers equipped with governors that limit speed are a few examples of ways in which transportation risk has been minimized. In addition, this plan will be evaluated for improvements at least annually. M-1 M.2 SAFETY M.2.1 POLICY Safety is of primary importance in planning and conducting all activities related to the transportation of the TRU waste. The objective is to protect the safety of the public and to protect the employees of the trucking contractor from occupational injuries and illnesses. In order to achieve this objective, the trucking contractor will rely on a variety of mechanisms and measures, including the following: • Compliance with all applicable health and safety requirements of the Federal Government, States, and local jurisdictions • Provision of vehicles and equipment with the best available mechanical safeguards, including governors that limit speed, and personal protective equipment • Provision of a facility for equipment maintenance and inspection • Implementation of a safety program, including personnel training in safe work practices • Stringent driver training program and penalty provisions • Accident and emergency training • Provision of a constant-surveillance service for all loaded shipments • Provision of communications equipment and services.. M.2.2 REQUIREMENTS FOR PROTECTING HEALTH AND SAFETY All activities related to the transportation of TRU waste will be conducted in accordance with the applicable health and safety requirements of the Federal Government, States, and local jurisdictions, including the requirements promulgated by the U.S. Department of Transportation in Title 49 of the Code of Federal Regulations (49 CFR). The maintenance facility (see Section M.4) will meet all applicable requirements of the U.S. Occupational Safety and Health Administration and the State of New Mexico. All trucks and drivers will meet the applicable requirements of the U.S. Department of Transportation. To ensure that these requirements are met, the trucking contractor will implement a maintenance and inspection program that will be regularly and continually monitored by contractor and DOE management. Another mechanism for ensuring regulatory compliance will be a safety program, which is discussed in the next subsection. When waste shipments are under way, all applicable regulations pertaining to the shipment of hazardous waste will be followed. M-2 As described in Section M.6, constant-surveillance service will be provided for all loaded shipments. In addition, a satellite-based tracking system will be used to determine the location and progress of all shipments. Such a tracking system is not a Federal requirement but is a voluntary DOE program decision for WIPP shipments. M.2.3 SAFETY PROGRAM The transportation contractor will establish and maintain a safety program that will consist of both a safety orientation for new employees and a continuing education program for all employees. To ensure that the safety program is successful, each employee will be made aware of his or her responsibilities in the program. All employees will be required, as a condition of employment, to observe established safety regulations and practices and to use the safety equipment provided. Every new employee will receive safety instructions, a personnel safety handbook, and any protective equipment deemed necessary. The orientation program for new employees will consist of verbal and written information on job safety, accident- prevention measures, and the responsibilities of the new employee in the safety program. In addition, each driver will be given special training as described in Section M.S. The continuing education program will Include training in applicable safety requirements and regulations, the use of equipment, and safe operating procedures. In addition, safety meetings will be held each week to train and inform employees. All employees will be required to participate in these meetings and to sign an attendance list. The immediate supen/isor will be responsible for conducting the meeting. A brief report on the subjects to be discussed will be prepared for each meeting. M.2.4 OCCUPATIONAL SAFET/ As a matter of policy, no employees will work in surroundings that are unsanitary, hazardous, or dangerous to their health or safety. All employees will be required to maintain their project or work areas. Adequate medical and first aid supplies will be available at all work locations. When needed, the employer will furnish tools, vehicles, and equipment with the best available mechanical safeguards and personal protective equipment. Employees using tools, vehicles, and equipment will be responsible for inspecting them before use to determine that they are in a safe, operable condition. Each member of the management team will be responsible for not only protecting the safety and health of all employees who report to or are assigned to him or her but also for the safe work conduct of those employees. M-3 M.3 EQUIPMENT The tractors used for hauling TRU waste to the WIPP will be provided by the trucking contractor. The trailers and the shipping containers (TRUPACTs) will be provided by the DOE. It is estimated that the tractor fleet will consist of 10 units domiciled in Hobbs, New Mexico. All vehicles will be 1 989 and later models, and will be replaced as needed throughout the program. All equipment used by the trucking contractor to transport TRU waste will conform to applicable Federal regulations (e.g., the requirements for placarding in 49 CFR Part 172); will meet the needs of the DOE; will meet all functional requirements for TRU waste shipments, such as being equipped with special tiedowns for the TRUPACT-II containers; and will have special equipment related to safety. For example, to prevent speed limits from being exceeded, the vehicles will be equipped with governors that will limit the speed to 65 miles per hour. In addition, the tractors will have a Tripmaster, which will automatically record all the speeds the vehicle reached in traveling. The tractors will also be equipped with radiation detection instruments for use by drivers who will be properly trained in their use, in the event of an accident. The specifications for the tractors are given in Table M.3.1. These specifications are based in part on the DOE's experience over the last 12 years in the transport of nuclear materials. The dimensions and weights of the tractors and trailers are given in Table M.3.2. These dimensions and weights are in compliance with applicable Federal and State safety requirements. M-4 Table M.3.1 Specifications for the tractors to be used in hauling TRU waste to the WiPP Make and model: Wheel-base length: Weight (dry): Engine: Power steering: Brakes steering axle: driving axles: emergency brakes: Engine brake: Transmission: Axles steering axle: driving axles: Tires steering: driving: Tire chains: Fenders steering wheels: rear wheels: Fifth wheel: Air-ride suspension: Mobile telephone: Citizens band radio: Other specifications: FLD-12064ST Freightliner 219 inches 15,915 pounds NCT 444 Cummins B/C4 @2100 rpm Ross TAS-65 by TRW, Inc. 15x2 CAM centrifuge drums 16-1/2 X 7 CAM centrifuge drums MGM dual brakes Cummins Brake Retarder Road Ranger 1 8-speed transmission 12000# FF 921 3800# SQ 1 00 A Michelin PXZA-1 Michelin XDHT Laclede Molded fenders Aluminum full fenders 18-inch Holland FW-2535 Freightliner air-ride suspension, 40,000 pounds Motorola Dynatac 6000x 40-channel COBRA 29-H Front leaf springs, 64 inch Aluminum wheels, frame, and fuel tanks Radiation detection meters alpha-beta-gamma meter beta-gamma meter Rockwell tripmaster Heated rear-view mirrors Heated and air-conditioned cab and sleeper Spray guards and mud flaps for the rear and front wheels Locking fuel caps Externally mounted fire extinguisher Tamper-proof fifth wheel locking device M-5 Table M.3.2 Overall dimensions of the tractor-trailer unit Length Tractor, total length: 26 feet 6 inches Trailer, total length: 42 feet 2 inches Total length: 62 feet 10 inches (with overlap of 5 feet 10 Inches) Width Trailer: 8 feet 6 inches Tractor: 8 feet 1 1 inches (includes side mirrors) Height Tractor: 12 feet Trailer with load (maximum): 13 feet 5 Inches Weight Tractor Weight dry Fuel Tire chains Drivers and equipment Spare tire Weight (pounds) 15,915 1,100 91 500 190 Tractor weight Trailer (includes tools and spare tire) Three loaded TRUPACT-II containers (maximum allowable) (Maximum loaded shipping weight of any single TRUPACT-II is 19,250 lbs) Total weight 17,796 8,500 53,299 79,595 M-6 M.4 EQUIPMENT MAINTENANCE M.4.1 MAINTENANCE FACILITY A facility for the maintenance, storage, and dispatching of tractors and trailers will be provided when required by the DOE. Until such time as a facility is required by the DOE, the tractors and trailers will be stored at the WIPP site. The proposed maintenance facility, to be located at a 6-acre site in Hobbs, New Mexico, will be designed to provide most of the facilities needed for fleet maintenance and operation as a truck terminal. It will contain a three-bay maintenance shop with an area of 6,500 square feet and an office building with an area of 1 ,550 square feet. If the proposed site is unavailable when the WIPP opens, an equivalent facility will be used. M.4.2 MAINTENANCE PERSONNEL AND EQUIPMENT Initially, the maintenance facility will be staffed by one mechanic, a shop helper, and security guards (see Subsection M.4.6). A second mechanic will be added when needed. All mechanics will have a minimum of 5 years of qualified experience related to diesel engines, air pressure, brake systems, electrical systems, and arc and gas welding. Certification of training in a 2-year technical school specializing in diesels and heavy equipment will be required. The mechanics will receive special training from the manufacturers of the tractors. The equipment and tools to be provided in the maintenance facility include the following: Overhead crane Grease pit Two 20-ton jacks Transmission floor jack Jack stands Engine stands Cutting torch Welder Drill press Hydraulic press Battery charger Air compressor with hoses M-7 M.4.3 MAINTENANCE SCHEDULE The schedule to be used for the maintenance of tractors and trailers is given In Table M.4.1. If the manufacturers recommend more frequent maintenance, the manufacturers' recommendations will be followed. Miscellaneous maintenance to repair broken wheels, flat tires, air fittings, air lines, and other similar items will be performed as required. All in-use tractors and trailers will be inspected monthly, with the inspection recorded on special forms. These forms, which are shown in Figures M.4.1 and M.4.2, specify the items to be inspected. In addition, the trailers will be inspected semiannually and annually (or after driving 10,000 or 20,000 miles, whichever comes first); these inspections will be recorded on the form shown in Figure M.4.3. Furthermore, as described in Section M.6, the tractors and trailers will be inspected by the drivers before each trip, every 2 hours or 1 00 miles during the trip, and after the trip. Table M.4.1 Maintenance schedule for tractors and trailers to be used to transport TRU waste to the WIPP Grease every 5000 miles. Oil and filter change every 15,000 miles or as specified by manufacturer^. New brakes and wheel seals every 100,000 miles or when needed, whichever is first. New tires every 1 00,000 miles or when needed, whichever is first. Miscellaneous maintenance to include universal joints, broken wheels, flats, air fittings, air lines, etc., as required. For tractors only. If it is necessary to test welds by a nondestructive examination method, arrangements will be made with a subcontractor. If difficulty in scheduling this procedure is encountered, the weld testing will be performed as directed by the DOE. For the trailers, which will be furnished by the DOE, no maintenance beyond that considered routine or preventative will be permitted. Also prohibited for the trailers will be any modifications, cutting, welding, or drilling, unless authorized by the DOE. M-8 MONTHLY TRACTOR INSPECTION DATE: I^KE: SERIAL jj: OWNER (LESSOR): LOCATION OF INSPECTION: MODEL: YEAR: UNIT ffi. TIRE PLY:. SPEEDOMETER READING: _ NUMBER OF TIRES: DRIVER: CAB or TRACTOfc NOT Kf. xr. f~~l ® I I n« CU. (8 B-C MOONTtO) I I I I nMi AW) RETUCTOW I I ® I I nMO (orroKAj.) I I @ I 1 m:k (ornoNAL) I I © I I StAT KlTi (BOTH StAH) [~~l ® I I WlfCSHtLO WPERS I I Q I 1 K>nt (ONE woRWNC) I I I 1 KFKOSTWS I I ® I I SPttDOMfltR (WORKNC) I I Q I 1 LW AR WARMNC oevict CHASSIS OF TRACTOR: OCF. NOT KF. I I I 1 WHEELS. RMS * LUCS r~~l @ I I BATTtRY COVER r~~l @ I I wrr TANKS (2) (DRAIN 74 HOURS) I 1 @ I 1 SPfSNCS (UAN UAF) I I © IZZl ^ft^R I I @ I 1 BRAKE HOSES * LCKT LOOM I I @ I I UHAusT rrsTEM I I Q I I PARKING BRAKES I I @ I I TIRE CHAINS (IN SEASON) I I ® I 1 lAJO fVAPS (FOR BOBTAIL) UCMTS OF TRACTOR: NOT OCF. (Dcm ©en (DCZI ©cm ©cm ©czi ©czzi ©czi otr. (ZZl CUD [ZZl CZ] CZZ] (ZZl MEAD (HIGH ft LOW BEAM) U*RKIR OR CLEARANCE foc (ornoNAL) SPOT (OPTONAL) TURN SCNAIS (TRACTOR ON) TURN SCNALS (TRACTOR Of REGULAR REAR UCHT3 STOP UCKTS REAfNEW MRR0R3 REOUrC OEMS: DCF. CZD NOT DEF. cm [ZZ] CZI czi @ STEERING (SECTOR BOX) STEERING TIRES (CUTS. SMOCTTH) OTHER TIRES (CUTS. SMOOTH. ETC.) UJC BOLTS (FRONT-ONE MSSNC) UJG BOLTS (BACK-TWO MSSINC) FIFTH WHEa (LOOSE UJUNHNC) BRAKE DRUkO (CRACKED OR BROKEN) I I WIfOSMELDS (BAD FITS OR CRACKS) I I AIR HOSES & LOOM (LEAKS OR CUTS) I I AIR COkFRESSOR (CMOKNC OR LEAK) AIR ERAKE TEST: MAXIMUM AIR PRESSURE: AMOUNT OF LOSS/1 MIN: CONDITION AND APPEARANCE: (CHECK ONE) EXCELLENT: D 0000:0 FAIR:D POOR:n OIL SAMPLE TAKEN: D YES D NO REPAIRS TO BE MADE BEFORE DISPATCH: I HEREBY CERTIFY THAT I HAVE CAREFULLY INSPECTED THE EQUIPMENT LISTED ABOVE AND THAT THIS IS A TRUE AND CORRECT REPORT OF THAT INSPECTION. SIGNATURE OF INSPECTOR SIGNATURE OF DRWER FIGURE M.4.1 EXAMPLE OF MONTHLY TRACTOR INSPECTION FORM M-9 V.-HITE - S.LC. RNK - SHOP COPT TRAILER No: Monn ! TRAILER/DOLLY INSPECTION MAKE: LENGTH: SERIAL No: NUMBER OF TIRES:. TIRE PLY: LOCATION OF INSPECTION: INSPECTOR: TIRE SIZE: (Fun Nam*) o o a zo I j j I OEARANCE UUP I 1 j I M«KER IM4> OR REFLECTOR j 1 I 1 WNDtMC GEAR I I I I HOSE CONNECnONS I 1 j j BRAKES AM> UNINSS I I j I NTERfcCtXArE UARKER UW I 1 I ' NTERkCWATE REFLECTOR CZD CZI TIRES I [ j WHEELS, RWS. LUGS I I I ' SPRINGS ANO MANGERS Lite Plug .Ate — -*- CLEARANCE 1>U.(P IMRKER LA».f> EMERGENCY BREAKAWAY VALVE 5TM WHEEL Pl>TE AtO PIN TIRES WHEELS. RMS, LUGS SPRINGS 1 I 1 I TIRES I I I I BRAKES At© UNINCS I I I I IMRKER \Mei I 1 I I POORS I I I I REAR EM> I I I I STOP LAMP j I I I TAIL LAMP I I I 1 REFUCTOR I I I I TURN SCNALS I i I I VtJDauP I I I 1 WIRING Out za 1=] cm CZI CD en czD UNINCS AND »WKES ' I I I NTERKDtATE MARKER LAI*» I I I I NTERkCOlATE REFLECTOR ' ' I I AIR MOSES NINGS AND BRAKES RED MARKER LAMPS UNDERCARRIAGE czD cm czD cm czD cm cm cm cm cm tizi czi d] CZI CZI CZI STOP LAMP I 1 I I TAIL 1>MP 1 I I 1 FUCrCR I I I I TURN SCNALS I I I I MUD FLAP I I I » REPAIRS (Circle defective items when corrected and check not defective box) AIR LOSS TEST WITH ALL SERVICE BRAKES APPLIED; NOTE LBS. IN 1 MINUTE Moximum permissible oir loss must not exceed two pounds per minute. Any oudible oir loss must be corrected immediately. REPAIR SECTION UST ALL REPAIRS MADE UST ALL PARTS OR EQUIPMENT INSTALLED Dote Repoirs - EXPLAIN - Attach extra sheet if necessary. Replacement or equipnnent installed MAINTENANCE AND SERVICING Itenn Date Lu brie ted Woshed (Steam Cleaned) Pointed I hereby certify that the mechanical defects Indicated above have been corrected. Dote Mechanic (Full Nome) FIGURE M.4.2 EXAMPLE OF MONTHLY TRAILER INSPECTION FORM M-10 TRUPACT TRAILER INSPECTION FORM Odometer Reading Date Equipment No. Make, Model Inspector 10,000 Miles/6 Mo. 20,000 Miles/12 Mo. Condition code: Initial if item is satisfactory. X indicates maintenance required. A check mark (J) indica t es service was performed. . 10 11 12 13 14 15 16 17 18 19 20 22 23 COND CODE 10.000 MILE INSPECTION Lubricate (per manufacturer specs) Check wheel seals for signs of leakage Inspect all air line assemblies, alad band gaskets for looseness . damage and routing Inspect air tanks for security and moisture Brake valves (visual condition, leaks) Service brakes (slack adjuste r travel) Visually check brake lininas from b acking plate side for wear ana looseness Apply brakes and check for leaks ( ?- psi per minute maximum) Electric brakes check operation and security Tires for remaining tread, unusu al wear, inflation, cuts and ' ** ' ' - , -, • n / -^ '^ 11 \ separations (minimum tread depth a llowable is 2/32" Hub oilers (level and leaks) Lug nuts and rim clamps (presence and no evidence of looseness Frame (cracks and paint condition ^ magnu flux suspect areas) Verify suspension system operation al; check landing gear and shoes (operation, security and, condition) Pull tongue/hitches (cracks, security and damage) — Check operation and condition of locking mechanism and kmgpm Check operation and condition of l anding gear and leveling ^acks Electrical wiring (condition, c hafing, and routing) System lighting (clearance, stop , turn, flasher and brakes) Check spare tire in Step 10 and v erify operation of tire lock Lubricate lock Check mud flaps for physical condition Check placard holders FIGURE M.4.3 EXAMPLE OF ANNUAL AND SEMIANNUAL TRAILER INSPECTION FORM M-11 TRUPACT TRAILER INSPECTION FORM Odometer Reading Date Equipment No. Make, Model Inspector 10,000 Miles/6 Mo. 20,000 Miles/12 Mo, CONDI CODE 2 0,000 MILE INSPECTION (Also perform 10.000 Mile Items' Visually inspec t condition of wheel bearings (clean and repack^ Inspect brake drums and lining (min. lining thickness is 3/32 above rivets^ Check kingpins for cracks fusing mag, particles) Check axle spin dles for cracks fusing mag, particles) Check kingpin and cou pler base ruse go/no go gauge to measure kingpin? clean couple r base and check for cracks and anomalies) Check swing beam bus hings and pins (maximum clearance between pin and bushing is .125 inches) Check trailer d eck for damage and attachment to frame FIGURE M.4. 3. (CONCLUDED) EXAMPLE OF ANNUAL AND SEMIANNUAL TRAILER INSPECTION FORM M-12 M44 QUALITY ASSURANCE AND CONTROL The trucking contractor will implement a quality assurance (QA) program that meets the QA requirements of the DOE. Procedures for the QA program will be developed, and personnel will be trained In their implementation. In addition, quality control procedures will be implemented. The trucking contractor will be responsible for ensuring the accuracy and reliability of measurements, tests, and maintenance procedures performed at the maintenance facility through the use of inspection, measuring, and test equipment of the range, accuracy, and type necessary to determine conformance with established requirements. To the extent required by established procedures, test equipment, gauges, and tooling will be calibrated by an approved standards laboratory. Items requiring calibration will carry readily visible labels showing their calibration status and will be recalibrated as necessary. Items with an expired calibration date will be segregated to ensure that they will not be used for maintenance or inspection. All replacement parts must conform to manufacturer's specifications for replacement parts and warranted by the maker. The supplier of parts will be required to provide a copy of the warranty at the time a part is delivered for the first time. For subsequent deliveries, the supplier will be required to submit a statement that the part conforms to the original warranty. The warranty and the subsequent quality assurance statement will be kept on file at the maintenance facility. Before it is placed in inventory or installed, each part will be inspected by the mechanic. The mechanic will be responsible for ensuring that all parts received conform to the warranty requirements. The packing slip or other document that accompanies the part will be stamped "Accepted by" and initialed by the mechanic and given to the dispatcher for review. Material or equipment that does not meet established requirements will be withheld from use until It has been appropriately repaired or reworked. All nonconforming items will be segregated and properly tagged to ensure that they will not be used. All providers of services will be required to supply documentation that the service meets accepted or required standards applicable to the service being rendered. They will be given a notice of requirements and will be required to certify that their work will be, and has been, conducted according to required standards by qualified personnel. Before authorizing any work, the trucking contractor will verify that the service provider can meet all requirements. The trucking contractor will verify compliance of the QA program by conducting audits at least every 6 months. The audited organization will verify and document the actions taken to satisfy any recommendations made by the auditors. The results of the audits will be documented and a copy sent to the DOE Transportation Representative. The QA program will include the requirement that records furnishing evidence of quality assurance be prepared and maintained; examples of such records are reports on audits, inspections, maintenance, and training. The detailed requirements for the control of the QA records will be included in the trucking contractor's QA procedures. M-13 At a minimum, these procedures will address legibility, retention, distribution, maintenance, transmittal to the WIPP, and protection against damage or loss. At least once a month, the maintenance records and the certification of parts and services provided by other firms will be reviewed by the dispatcher to determine that all standards are being met. If the dispatcher finds that a part or service was not properly certified, the use of that part or service will cease immediately. The provider of the part or sen/ice will be notified in writing and required to furnish certification. If certification is not immediately furnished, the provider will be removed from the list of acceptable providers. If noncertified parts have been installed, the dispatcher will order an immediate inspection of the part to determine whether the part is adequate. If adequacy cannot be ascertained, the part will be replaced. In the event of a noncertified sen/ice, the dispatcher will order an immediate review, and the service will be repeated if necessary. The dispatcher will conduct random inspection to verify the adequacy of repairs performed by employees and by providers. M.4.5 RECORDS In addition to the QA records discussed above, a record file will be maintained for the inspection sheets and shop tickets for each tractor and trailer. Parts-inventory cost sheets will be attached to each shop ticket (see Figure M.4.4), All records will be prepared in triplicate. One sheet will be placed in the file mentioned above, one sheet will be fon/varded to the contractor's home office, and one sheet will be filed at an off-site location. M.4.6 SECURITY Security for the maintenance facility will be provided by the following physical features and by personnel procedures. The site will be surrounded by a 6-foot-high chain-link fence with barbed wire at the top. Access will be allowed only for authorized personnel, who will be admitted through a single gate controlled by personnel inside the facility. Floodlights will be used to illuminate the shop, office, fueling, and truck storage area. The site will be occupied at all times (24 hours a day, 365 days a year) by maintenance or dispatching personnel or by a security guard. All deliveries will be accepted at the gate. If a maintenance service is to be provided by a subcontractor, the service provider will be accompanied by an authorized employee of the maintenance facility. No unauthorized access by the public will be allowed at any time. M-14 NAME OF PART LABOR DESCRIPTION: DAWN TRUCKING COMPANY P.O. BOX 204 FARMINGTON, NEW MEXICO 87499 44168 DATE: MECHANIC'S NAME UNIT# SPEEDOMETER READING M. TO ,M. TOTAL HOURS. san iuan repro Form 295-3 FIGURE M.4.4 EXAMPLE OF DAWN TRUCKING SHOP TICKET M-15 M.5 DRIVERS It is estimated that 30 drivers will be needed for the trucking program, and the trucking contractor will ensure that only qualified drivers are hired. The contractor, who is an equal opportunity employer, will locate qualified drivers by posting job openings in Job Service centers in all communities near the WIPP site, including Hobbs, Carlsbad, and Roswell, as well as major cities in New Mexico and western Texas. In addition, the contractor may place advertisements in trucking publications. Drivers will be selected on the basis of ability and experience. M.5.1 DRIVER QUALIFICATIONS To qualify initially, applicants will have to meet the following requirements: they must be citizens of the United States and at least 25 years of age; they must have logged at least 100,000 miles in driving semi-tractor trailers, must have at least 2 years of uninterrupted experience in driving commercial semi-tractor trailers during the last 5 years, and may not have any moving violations (including chargeable accidents) in the past 3 years. The driver-qualifying process will consist of the following: Completing an application for employment Initial interview Verification of employment ~ including years of service and mileage logged Check of driving record, including possession of a Commercial Driver's License A test, given by qualified personnel, that examines performance in the following: ~ Pretrip inspection ~ Coupling and uncoupling of tractor and trailer - Placing tractor in operation ~ Use of tractor controls and emergency equipment ~ Operating the tractor in traffic and while passing other vehicles ~ Turning the tractor - Braking and slowing the tractor by means other than braking (shifting gears) ~ Backing and parking the tractor Drug screening M-16 • Physical examination • Written test on Federal motor-carrier safety regulations and hazardous materials regulations in accordance with 49 CFR 391 .35 • Driver-profile evaluation. When a driver has successfully completed this qualification process, a written report on the driver will be sent to the DOE for approval (see Figure M.5.1). If approved, the driver will be trained as described in the next subsection. M.5.2 DRIVER TRAINING PROGRAM Every driver hired by the trucking contractor will have to complete a training program in accordance with the requirements of 49 CFR 177.825. In addition, every driver will receive training to meet the requirements of 49 CFR Part 397. The training to meet the requirements of 49 CFR will be conducted by the Colorado Safety Institute in Denver. However, if necessary to meet scheduling requirements, an alternative qualified source of training may be used. In addition, every driver will be trained to meet special DOE requirements pertaining to the specific characteristics of the TRUPACT-II shipping containers, the transportation of radioactive materials, monitoring equipment, emergency response, and public relations. In addition, the drivers will be required to attend a training class conducted by the Transportation Safeguards Division of the DOE's Albuquerque Operations Office. This training will be comprehensive, requiring approximately 68 hours. One instructor will be provided for each two drivers. The training will include driving a WIPP tractor-trailer unit carrying TRUPACT-II containers with simulated loads. Before the actual shipment of any waste, multiple dry runs from each waste site will be conducted as part of a series of preoperational checks designed to provide experience and hands-on training to the drivers (see Appendix D.2.3.2). M-17 DATE: NAME:. SS#: ADDRESS: DRIVER QUALIFICATIONS CONTRACT NO. DE-AC04-89AL51527 DOB: DATE AND SOURCE OF TRAINING AS REQUIRED BY 49 CFR 177.825: SEE ATTACHMENTS FOR: (a) Verification of 1 00,000 miles of semi-tractor trailer combination driving experience. (b) Evidence that this driver has had two years of uninterrupted semi-tractor trailer commercial driving experience during the last five years. I DO HEREBY CERTIFY THAT THE ABOVE NAMED DRIVER IS A CITIZEN OF THE UNITED STATES OF AMERICA. I DO HEREBY CERTIFY THAT THE ABOVE NAMED DRIVER DOES MEET THE REQUIREMENTS OF 49 CFR 391 , THE COMMERCIAL MOTOR VEHICLE SAFETY ACT. AND PARAGRAPH 5.2 OF THE DAWN MANAGEMENT PLAN. SIGNED: FIGURE M.5.1 DRIVER QUALIFICATION FORM M-18 M.6 PROCEDURES USED IN WASTE TRANSPORTATION M.6.1 RESPONSIBILITY FOR DAILY OPERATIONS The manager/dispatcher at the Hobbs maintenance facility will be responsible for the daily operations of the trucking contractor. The dispatcher will receive and review trip schedules furnished by the WIPP. These schedules will be furnished for inten/als of no less than 6 weeks. If there are problems about the schedules, the dispatcher will immediately communicate with the WIPP to resolve the problems. The dispatcher will prepare and distribute a 30-day schedule to all drivers. If a driver notifies the dispatcher that there are problems with the schedule, the dispatcher will resolve the problem. The dispatcher will be reachable by beeper or telephone at all times when not in the dispatch facility. M.6.2 NUMBER OF DRIVERS Two qualified drivers will be used for each shipment of TRU waste. If a driver becomes incapacitated along the way, the alternative driver will ask and receive appropriate instructions from the dispatcher before proceeding. M.6.3 SECURITY Standard security requirements for materials in transit, as specified in DOE Order 1540.1, will be applied to the TRUPACT-II shipping containers in both the loaded and unloaded condition. Constant surveillance will be provided for each shipment (Subsection M.6.7), and the drivers will know the procedures to be followed in the event of a deliberate obstruction of a shipment. In addition, the location of each TRU waste shipment will be known at all times, via the TRANSCOM satellite-based tracking system (Section M.8). M.6.4 PROCEDURES TO BE FOLLOWED BEFORE THE START OF THE TRIP The drivers will report to the dispatch center in Hobbs 1 hour before the scheduled time departure. The driver will check in and receive trip routing instructions. The dispatcher will verify that the drivers have arrived to review the route to be taken for the trip. The routes to be taken are the routes defined as "preierred" in Federal regulations. The two drivers assigned to the trip will review the trip route together. If they have any questions, they will discuss them with the dispatcher. M-19 The drivers will obtain a copy of the pretrip inspection form (Figure M.6.1) and the trip report form from the previous trip. They will inspect the truck and the trailer, paying particular attention to any items mentioned as possibly defective in the post-trip report. The drivers will sign the pretrip report if the tractor and the trailer meet requirements. The inspection will include all extra equipment. If their inspection of the tractor and trailer shows that an item or items do not meet the required standards, the drivers will notify the dispatcher. The dispatcher will decide whether the tractor and trailer are to be dispatched in their current condition or whether further maintenance is required. If the dispatcher decides to dispatch the tractor and trailer without further maintenance, the drivers have the option of noting their concurrence or nonconcurrence with the decision of the dispatcher. If the dispatcher decides to use another tractor or trailer, the drivers will carry out the same inspection routine. M.6.5 PROCEDURES TO BE FOLLOWED AT THE WIPP SITE At the WIPP site there will be two trailer-parking areas. Parking Area A will be for trailers incoming with loaded TRUPACT-II shipping containers and trailers that have been inspected by the trucking contractor and are ready to be loaded. Parking Area B will be for empty trailers that require inspection or maintenance and for trailers that are ready for shipment and are loaded with empty TRUPACT-II containers. At the WIPP site the drivers will present the necessary identification and documentation and receive the shipment documentation, including a manifest which, for mixed waste shipments, conforms to the requirements of 40 CFR Part 263. They will then proceed to the trailer-storage area. At the trailer-storage area, the drivers will leave their tagged empty trailer in Parking Area A and verify that the trailer (from Parking Area B) loaded with empty TRUPACT-II containers has been tagged as ready for service. The drivers will then inspect the trailer, using the trailer-inspection form. As part of the pretrip inspection, the drivers must ensure that the permanently affixed flip-type placards properly signify whether the trailer is carrying a load containing radioactive material or is empty. If the trailer meets ail inspection requirements, the drivers will sign the trailer-inspection sheet and depart from the WIPP site. The departure will follow the correct procedures for notification and departure. If the trailer does not meet the required standards, the drivers will notify the WIPP and the dispatcher. The drivers will then await a decision by the WIPP and the dispatcher concerning the departure of the trailer. M-20 DAWN TRUCKING Driver Vehicle Inspection TRArrOR nATF MM FAHF (^) CHECK ANY DEFECTS NOTED BELOW PARKING (HAND) BRAKE WHEELS AND RIMS STEERING MECHANISM EMERGENCY EQUIPMENT LIGHTS AND REFLECTORS ENGINE TIRES TRANSMISSION HORN CLUTCH WINDSHIELD WIPERS EXHAUST REAR VIEW MIRRORS BRAKES COUPLING DEVICES COOLING AND OIL PRESSURE ACCESSORIES OTHER EXPLAIN IN DETAIL ANY DEFECTS CHECKED (TRACTOR ONLY) LAST P.M. (DATE)— IF NO DEFECTS-WRITE "NONE" EXPLAIN IN DETAIL ANY TRAILER DEFECTS TRAILER NO. TRAILER NO. DRIVER'S SIGNATURE DATE 1 HAVE INSPECTED THE ABOVE UNIT AND REPORTED ALL DEFECTS KNOWN TO ME REPAIRMAN'S SIGNATURE DATE 1 HAVE MADE ALL NEEDED REPAIRS OF THE DEFECTS REPORTED ON THIS UNIT Mti iutn repro Form 29S-47 FIGURE M.6.1 EXAMPLE OF DRIVER'S VEHICLE INSPECTION FORM M-21 M.6.6 GENERAL PROCEDURES TO BE FOLLOWED DURING THE TRIP The drivers must use the preferred route for shipments unless a deviation is permitted under the provisions of 49 CFR 177.825. A deviation is permitted by 49 CFR 177.825 under the following circumstances: 1) Emergency conditions that would make continued use of the preferred route unsafe 2) To make necessary rest, fuel, and vehicle-repair stops (stops will be along the preferred route) 3) To the extent necessary to pick up, deliver, or transfer a highway route controlled quantity package of radioactive materials. Any required deviation will be reported to the DOE's representative at the WIPP before the deviation occurs. Any unauthorized deviation from the preferred route will result in penalties, as discussed at the end of this section. Drivers may alternate driving shifts of approximately 5 hours. Thus, the vehicle will be constantly moving unless stopped for inspection, fueling, or weather. When circumstances require an extended stop, the driver will ensure that the shipment is parked in a safe manner. M.6.7 CONSTANT SURVEILLANCE One driver will keep the tractor and trailer under constant surveillance at all times. Constant surveillance is defined to mean that when the vehicle is not being driven, it must be attended at all times by a driver or a qualified representative of the trucking contractor. A vehicle is "attended" when at least one driver is in the tractor, awake, not in a sleeper berth, or within 1 00 feet of the vehicle and has the vehicle within his or her constant unobstructed view. If an extended stop is necessary, a driver must keep the shipment in full view and stay within 1 00 feet of the shipment at all times. The trailer with the TRUPACT-II containers must always be connected to the designated tractor during shipment except when stopped at a DOE facility for loading, unloading, or en route to maintenance. M.6.8 INSPECTIONS DURING THE TRIP The drivers will park the vehicle in a safe place every 2 hours of travel time or 100 miles, whichever is less, and inspect the vehicle. Deficiencies will be corrected at this time or at the next available repair area. The items to be inspected include the tires, tiedowns, labeling and placarding required for the transportation of radioactive materials, and the antenna used for the TRANSCOM M-22 vehicle-tracking equipment (see Section M.8). Items found to be nonconforming will either be corrected at this time or at the next available repair area. If a tire is found to be flat, leaking, or improperly inflated, the tire will be changed or properly inflated. The drivers will also inspect the vehicle lights if lights will be needed before the next stop. Hose connections will be checked, and a visual inspection of the entire vehicle will be made. "me DOT regulations in 49 CFR 397.17 (Transportation of Hazardous Materials: Driving and Parking Rules") require only tire inspections every 2 hours on vehicles carrying hazardous materials. The DOE has expanded this inspection requirement to include other components and to include unloaded vehicles. M.6.9 PROCEDURES AT THE WASTE SITE On arrival at the waste site, the drivers will stop at an inspection point where the driver and shipment documentation will be checked by site security before the tractor and trailer are permitted entrance. Specific items to be verified are the bill of lading, tamper- indicating devices, and the serial numbers of the TRUPACT-II shipping containers. The drivers will then proceed to the trailer-parking area and drop off the trailer with the empty TRUPACT-II containers. The drivers will undertake an after-trip inspection of the trailer. They will then proceed to the location of the trailer with loaded TRUPACT-II containers, or, if at a low-volume site, find out when they should return to pick up the trailer after it has been loaded. The drivers will receive trip documentation and inspect the trailer, using the trailer- inspection form. The drivers will also inspect the tractor before departing from the waste site. The drivers will follow the approved departure procedure when leaving the site. The drivers will then proceed to the WIPP site, using the same routes and procedures used with the empty TRUPACT-II shipping containers. M.6.10 PROBLEMS DURING THE TRIP If the dispatcher is notified by the driver of a problem during the trip, the dispatcher will notify the DOE's representative at the WIPP. If the WIPP notifies the dispatcher that a problem exists, the dispatcher will immediately contact the drivers to ensure that procedures are being followed and to obtain firsthand information on the situation. The dispatcher will decide on the best course of action and notify the WIPP of the decision. If the WIPP concurs, the decision will be implemented. If the WIPP does not concur, further discussions will take place. When notified of a mechanical problem that prevents the tractor or trailer from moving, the dispatcher will immediately make arrangements to rectify the situation after consultation with the WIPP. If a leased tractor is to be used, the dispatcher will consult the list of locations where tractors are available for leasing from a qualified leaser and determine the most convenient location. The leaser will be called and asked to dispatch a tractor that will allow the shipment not to exceed a total weight of 80,000 pounds. The WIPP and the drivers will be notified of the expected time of arrival. M-23 All drivers will carry full Instructions for actions to be taken in the event of an accident. The procedures to be followed after an accident are discussed in Subsection M.7. M.6.11 DEUVERY OF WASTE AT THE WIPP SITE On arrival at the WIPP site, the driver will stop at an inspection point where the driver and shipment documentation must be checked by site security before the shipment is permitted into the secured area. Specific items to be verified are the bill of lading, tamper-indicating devices, and the serial numbers of the TRUPACT-II shipping containers. Shipments will have a radiation survey performed in the designated secure area before entry into the site. The drivers will be badged and proceed to a receiving- inspection position in the radioactive-materials area. When a shipment arrives at the WIPP site, one driver will remain with the vehicle at all times. The driver will position the trailer as required for further processing in one of the parking areas. After the trailer has been removed, the tractor and drivers will be released. If an empty trailer is available, the drivers will pick up the empty trailer from Parking Area B for delivery to the maintenance facility. The drivers will then return to the maintenance facility with the tractor or tractor and trailer. M.6.12 AFTER-TRIP REPORT At the conclusion of each round trip, the drivers will complete the driver's vehicle- condition report for the tractor and trailer. They will review the report with the maintenance supervisor. The drivers will be encouraged to present their observations on the performance of the vehicle (tractor and trailer). M.6.13 PENALTIES FOR DRIVERS If the drivers fail to follow the prescribed procedures, they will be subject to penalties. For an unauthorized deviation from the preferred route, the penalties will be as follows: • First time ~ written warning and 2 weeks' leave without pay • Second time ~ termination of the driver's employment. A failure to maintain adequate records will result in the same penalties as deviating from the route. The failure to maintain constant surveillance of the vehicle will result in a termination of the driver's employment. A chargeable accident will result in a termination of the driver's employment. A moving violation will result in a termination of the driver's employment. M-24 M.7 PROCEDURES FOR ACCIDENTS AND INCIDENTS All drivers will carry full instructions for actions to be taken in the event of an accident. These instructions will include the procedures for obtaining local, State, or Federal assistance if technical advice or emergency assistance is needed. The TRANSCOM equipment (Section M.8) will provide a communications capacity that can be used in any emergency. The accidents to be reported are those specified in the applicable Federal regulations, 49 CFR 171.15 and 171.16, the general requirements of 49 CFR Part 394, and the requirements of DOE Order 1540.1. All accidents, no matter how minor, will be reported to the traffic manager of the waste site, the WIPP, and the dispatcher. Accident reporting will follow normal procedures (49 CFR Part 394) for minor accidents that involve no obvious or suspected damage to the TRUPACT-II shipping containers. In the event of a Type A accident (as defined in DOE Order 5484.1), it will be necessary to notify the DOE Headquarters Emergency Operations Center, and this notification will be made through the Albuquerque Operations Office. The trucking contractor will notify the DOE's Albuquerque Operations Office, the U.S. Department of Transportation, the WIPP, and the shipper in the event of fire and damage in excess of $5,000, breakage, spillage, or suspected contamination with radioactive material, as required by 49 CFR 171.5 and 171.861. When notified of an emergency situation, the dispatcher will immediately contact the WIPP. If action is needed by the dispatcher, such action will be taken with the concurrence of the WIPP. These actions may include, but are not limited to, the following: • Having the vehicle repaired • Dispatching a replacement tractor • Sending replacement drivers • Coordinating a route deviation • Authorizing shipment of replacement parts. The dispatcher will maintain a log of actions taken during the emergency, including the time of each action. A copy of the record will be sent to the WIPP. If the drivers perceive a potential obstruction because of a public demonstration, the drivers will immediately notify the local law enforcement agency and the WIPP and describe the situation. The WIPP will advise the drivers as to what action to take. If it is determined by the drivers that the trip should not continue, the drivers will move the tractor to the most secure nearby location, if feasible, and remain with the vehicle. M-25 If it is determined by the drivers that the tractor and trailer cannot be moved because of a deliberately placed obstruction or public demonstration, the drivers will do the following: 1 ) Notify the WIPP immediately 2) Notify the local law enforcement agency or the State highway patrol 3) Remain in the tractor with the doors secured. M-26 M.8 SHIPMENT TRACKING AND COMMUNICATIONS M.8.1 SHIPMENT TRACKING The location of each TRU waste shipment will be monitored in order to maintain shipping and receiving schedules and to learn of any unplanned deviation from the schedule or preferred route. This monitoring will include the status of the shipment at the WIPP site or at the waste site as well as location during transit. The primary method for monitoring or tracking TRU waste shipments will be the TRANSCOM locating system. TRANSCOM will use a land-based Loran positioning system to obtain exact data on the longitude and latitude. It will have a transmitter to transmit the Loran C data via satellite to the TRANSCOM Control Center at Oak Ridge, Tennessee, which will be linked to the Central Communications Center at the WIPP (see Appendix D). The transmissions will be converted to location data by the TRANSCOM central computer. TRANSCOM will provide a two-way digital means of communication. However, with the TRANSCOM system providing routine data, communication by the driver will be required only in the event of significant schedule impacts, such as accidents or delays that affect the delivery schedule by 2 hours or more. M.8.2 BACKUP COMMUNICATIONS In the event that the TRANSCOM location system is not available, telephone communications will be used, and the drivers will use the mobile telephone provided. Telephone communications will also be used by the dispatcher and by the waste site to report to the WIPP. To facilitate telephone communications, 800 numbers will be available. The required reports will be as follows: • The drivers will be required to make a telephone call to the WIPP every 2 hours and when crossing State borders, or as soon thereafter as practical, to report their location. • Any delays and the reason for delays in transit longer than 2 hours will be reported by the trucking contractor to the WIPP, who will in turn relay the information to the waste site. • The waste site will notify the WIPP at the time the shipment leaves the site. The notification will include the tractor and trailer numbers, the serial numbers of the TRUPACT-II containers, the drivers' names, the bill-of-lading number, the shipment weight, the route, the date and time the vehicle departed, and the expected arrival time. M-27/28 APPENDIX N RE-EVALUATION OF RADIATION RISKS FROM WIPP OPERATIONS N-i/ii TABLE OF CONTENTS Section N.1 INTRODUCTION N.2 REVIEW OF RECENTLY PUBLISHED RADIATION RISK EVALUATIONS N.2.1 BEIR-IV N.2.2 UNSCEAR N.3 REASSESSMENT OF RISKS FROM WIPP OPERATIONS . . . N.3.1 Methodology Selected N.3.2 Scenarios Selected N.3.3 Public Health Effects N.3.4 Genetic Effects REFERENCES FOR APPENDIX N LIST OF TABLES Table N.3.1 Estimated excess fatal cancers caused by WIPP operations during the Test and Disposal Phases N.3.2 Lifetable for population dose and risk resulting from routine emissions N.3.3 Estimated excess genetic effects caused by WIPP operations Page N-1 N-3 N-3 N-3 N-5 N-5 N-8 N-9 N-1 6 N-1 8 Page N-9 N-10 N-1 7 N-iii/iv N.1 INTRODUCTION Since the supplemental risk assessment process was initiated, there have been two new evaluations of the risks posed by radiation exposure published (BEIR, 1988 and UNSCEAR 1988) ^ In response to comments made by the DOE during its internal review of the draft SEIS, this appendix has been prepared to evaluate the extent to which these recent studies may affect the estimation of risks reported in this SEIS. The selection of a risk estimator to evaluate the radiation-induced human health effects of WIPP operations is discussed in Subsection 5.2.2.1. These estimated health risks are summarized in Table 5.14 for transportation-related exposures and in Tables 5.29 and 5 30 for WIPP routine and accident-related exposures, respectively. To establish that the risk estimators utilized provide a conservative estimation of health risk a comparison is made between certain reported health risks and those which would be predicted by a rigorous application of data provided by the newly available studies. Based upon data from the BEIR-III report (BEIR, 1980), risk estimators for both cancer incidence and genetic effects have been developed to estimate health effects associated with the calculated doses to the population and individuals. For cancer incidence, a risk estimator of 280 fatal cancers per million person-rem of radiation (external dose plus committed effective dose equivalent) received by the affected population has been used For genetic effects, a risk estimator of 257 genetic effects per million live-born offspring for each additional rem of radiation received by the gonads of the affected population has been used. 1 On December 20 1989, the National Research Council's Committee on the BEIR issued a report on the health effects of exposure to low levels of ionizing radiation (BEIR 1989) This report includes information and analyses from the BEIR-IV report (BEIr' 1988) that are appropriate for cancer and genetic risk assessment along with the delayed health effects that are induced by low linear energy transfer (LET radiations such as x-rays and gamma radiation. These health effects include fatal cancer induction (carcinogenesis), genetic effects, and retardation from in utero exposure Quantitative risk estimates based on statistical analyses of the results of human epidemiological studies and animal experiments are presented in the BEIR- V report A significant portion of the BEIR-V report deals with carcinogenesis in humans because of the extended follow-up in major epidemiological studies (e.g., Japanese atomic-bomb survivors and radiotherapy patients) and the revision of the dosimetric system for the Japanese atomic-bomb survivors. The report presents risk factors that are higher than proposed in the BEIR-III report (BEIR 1980) The BEIR-V report estimates that 800 extra cancer deaths would be expected to occur during the exposed population's remaining lifetimes if 100,000 people of all ages were exposed to a whole body dose of 10 rad (or 10 rem) of N-1 gamma radiation in a single brief exposure. These 800 excess cancer deaths are in addition to the nearly 20,000 cancer deaths that would occur in the absence of the radiation. This corresponds to a risk factor of 8.0 x 10"^ excess fatal cancers per person-rem (this SEIS used 2.8 x 10"^ excess fatal cancers per person-rem). The 90 percent confidence limits, based solely on sampling variation, for increased cancer mortality due to an acute whole body dose of 10 rem range from about 500 to 1,200 (mean 760) for 100,000 males of all ages and from about 600 to 1,200 (mean 810) for 100,000 females of all ages. The report also recommends using the relative risk model (as used in Subsection N.3) instead of the constant absolute or additive risk model. The report recognizes that the assessment of carcinogenic risks that may be associated with low doses of radiation requires extrapolation from effects observed for doses exceeding 10 rad and is derived from assumptions about dose-effect relationships and the mechanisms of carcinogenesis. In the analysis of the epidemiological data for the atomic-bomb survivors, the survivors receiving less than 0.5 rad serve as a control group for the survivors receiving more than 0.5 rad. The report also recognizes that its risk estimates become more uncertain when applied to very low doses; however, the risk estimates could either increase or decrease. For low-LET radiations such as gamma rays, the consensus is that cell survival is enhanced by a decrease in dose rate or separation of the dose into several fractions. To apply the models derived from the data on acute exposures, the dose rate effectiveness factor must be considered. The BEIR-V report indicates that it may be desirable to reduce the estimates given above by a factor of 2 for application to populations exposed to small doses at low dose rates because of the dose rate effectiveness factor. The report recognizes many uncertainties in its analyses. These include the application of results from a Japanese population (with different naturally occurring cancer rates) to a United States population, the certification of the cause of death, time- and age-related effects, and the shape of the dose-response curve. It also recognizes that direct estimates of the lifetime risk can be obtained only after the exposed population has been followed for a lifetime; however, the Japanese survivors (one of the populations followed for the longest time) have been followed for only 40 years. The report also states that studies of populations chronically exposed to low- level radiation (e.g., those residing in regions with elevated natural background radiation) have not shown consistent or conclusive evidence of an associated increase In the risk of cancer. The risk factors presented in BEIR-V, which became available as this SEIS was in the final stages of completion, are not incorporated in the risk estimates. The DOE will have to study the report thoroughly to determine any warranted changes in risk estimation methods for the generally low dose/low dose rate circumstances analyzed in this SEIS. The purpose of this SEIS, however, is to provide environmental impact information for deciding whether to proceed to the Test Phase (Proposed Action or Alternative Action). In this context, BEIR-V is not significant because 1) the likely increases in risk estimates are relatively small; 2) they affect all alternatives, including No Action; and 3) the DOE will issue another SEIS-using the then current risk assessment methods-before a decision to enter the Disposal Phase, during which most of the radiological impacts associated with the WIPP are predicted to occur. N-2 N.2 REVIEW OF RECENTLY PUBLISHED RADIATION RISK EVALUATIONS Two recently published evaluations of the risks posed by exposure to ionizing radiation contain data relevant to the radionuclide distribution for the WIPP. These studies are reviewed in terms of determinations and recommendations associated with predicting human health risk from exposure to alpha-emitting radionuclides. N.2.1 BEIR-IV In January 1988. the National Research Council's Committee on the Biological Effects of Ionizing Radiations (BEIR) issued a report reviewing available information on the health risks of alpha-emitting radioactivity which has deposited inside the human body (BEIR. 1988). This information is directly relevant to the WIPP, since virtually all of the radionuclides present in TRU waste are alpha-emitters. In their review the BEIR Committee determined that the effects of internally-deposited TRU radionuclides occur predominantly in three organs: the bone, the liver, and the lung Based on data from animal studies as well as limited human exposure data, the BEIR Committee recommended latency periods (i.e., the time between exposure to radiation and the onset of cancer) and risk factors for these organs as follows: Organ Latency Period (years) Fatal Cancer Risk (deaths/million person-rad) Bone 5 Liver 20 Lung 5 300 300 700 For the bone risk factor, the absorbed dose used is the mean bone dose. N.2.2 UNSCEAR In 1988 the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) issued the latest in a series of reports to the General Assembly, providing a comprehensive assessment of the sources, effects, and risks of ionizing radiation (UNSCEAR, 1988). In this report, the Committee reviewed available data on radiation exposures and risk estimates. N-3 The Committee recommended a range of risks for radiation-induced fatal cancer. Adjusting for the effects of low doses/dose rates as prescribed by UNSCEAR, the absolute lifetime risk of radiation is 200 to 250 fatal cancers per million person-rad. Latency periods were given as a minimum of 2 to 5 years between exposure and the onset of either leukemia or bone cancer and 1 years for all other types of cancer. These values are similar to those proposed In the BEIR-III report and used in this SEIS. N-4 N.3 REASSESSMENT OF RISKS FROM WIPP OPERATIONS Based on the information in the reports discussed in Subsection N.2, a reassessment of the risks posed by WIPP operations was performed. The approach used is patterned after the RADRISK computer code (ORNL. 1980), and could be applied to any aspect of the WIPP where radiological dose assessments are performed, including the transportation risk assessment. To establish that the risk estimators used in this SEIS remain conservative, facility operational impacts were selected for reassessment. N.3.1 METHODOLOGY SELECTED The methodology selected for this assessment uses a life table approach to predict the estimated lifetime risk of fatal cancer from exposure to radiation/radioactivity emitted during the operation of the WIPP. The reassessment calculates the effects of exposure to two types of radiation: • Low Linear Energy Transfer (LET) radiation (such as gamma and beta radiation), because of its penetrating nature, can cause damage from either outside the body, from external sources, or inside the body, once ingested or inhaled • High-LET radiation (such as alpha particles or neutrons) is primarily made up of less penetrating alpha radiation, which can cause damage once inside the body. Low-let radiation exposure risk at the WIPP during normal operations is associated almost completely with WIPP occupational workers who are subject to external exposure to gamma radiation while handling the waste containers (primarily the CH TRU shipping containers and containers of TRU waste). WIPP employees and the off-site population can also be exposed to gamma and beta radiation from a plume of radioactivity released in the event of a postulated accident. Radiation doses to low-LET radiation are described in Subsections 5.2.3.3 and 5.2.3.4. The prediction of fatal cancers associated with low-LET radiation exposure uses relationships between absorbed dose and risk developed in the BEIR-III report (BEIR, 1980). The relationships selected use a linear quadratic form to express the relationship between absorbed doses and the risk of cancer: 1) Leukemia and bone cancer (BEIR-III, Table V-16) 2) All other types of cancer (BEIR-III, Table V-19). The relationships were combined to generate the formulae used in the lifetable. N-5 In accordance with BEIR-III, a 10-year latency period is assumed for low-LET radiation prior to the onset of cancer. Any radiation-induced cancer will not begin to develop until the end of this latency period. In the eleventh year, the risk would be related to the exposure in the first year; the risk in the twelfth year would be related to the exposure in the first and second years; risk in subsequent years would be evaluated in the same manner. Once the latency period had passed, an exposed individual would have a risk of radiation-induced cancer for the remainder of his/her lifetime. If the exposure is continued, the risk would continue to increase. When the exposure is stopped (e.g., by termination of WIPP operations), the risk would continue to increase for the length of the latency period and thereafter would remain constant. Specifically for low-LET radiation and 25 years of operation, the risk of radiation-induced cancer would begin in the eleventh year and continue to increase until the thirty-sixth year, when it would become constant for the duration of the individual's lifetime. The risk, in the thirty-sixth and following years, would be dependent on the total exposure during the 25 years of operation. The dose equivalents caused by high-LET radiation exposure to WIPP waste are the result of inhaling, and to a lesser extent, ingesting alpha-emitting radioactivity. They are expressed in terms of committed effective dose equivalents (CEDE's), which provide a measure of the damage done to the body over a 50-year period due to an intake in a single year. These CEDE's are described in Subsections 5.2.3.3 and 5.2.3.4. To assess the impact of these CEDE's on human health, they are converted into organ doses to the bone, the liver, and the lung as identified by the BEIR-IV report (BEIR, 1988). The prediction of fatal cancers associated with high-LET radiation is accomplished through a series of steps: 1) The conversion of CEDE's to annual effective dose equivalents 2) The conversion of annual effective dose equivalents to annual organ dose equivalents 3) The prediction of fatal cancers for each organ 4) The summation of the organ fatal cancer risks to predict the total risk of cancer. The waste going to the WIPP will contain a variety of radionuclides which emit high- LET radiation. In an attempt to simplify the evaluation of the various types of radionuclides, the SEIS uses the concept of the "Plutonium-239 Equivalent Curie (FE- CI)." This concept, described in Appendix F.2, uses the ratio of effective dose equivalent conversion factors between a radionuclide and plutonium-239 (Inhalation Class W) to convert each radionuclide's concentration into an equivalent concentration of plutonium-239(W). All analyses then treat the waste as though plutonium-239(W) were the only radionuclide present. The dose conversion factors used are for the N-6 inhalation pathway, using a 1.0 micron aerodynamic median activity diameter (AMAD) and a 50-year commitment period (Dunning, 1986). Since the retention time for plutonium-239 in the human body is so long (ICRP, 1979), this methodology assumes that the radioactivity remains in the organ of interest for an indefinite period. Thus, the 50-year CEDE's are converted to annual effective dose equivalents simply by dividing by 50. Further, the annual effective dose equivalents are assumed to continue throughout the population's lifetime (i.e.. they do not stop at the end of the 50-year period). To obtain the dose equivalent to the three specific organs of interest (bone, liver, and lunq) each annual effective dose equivalent is multiplied by the ratio of the organ CEDE dose conversion factor to the effective dose conversion factor for plutonium- 239(W) (Dunning, 1986). To ensure that this approach was conservative, the conversion factors from effective dose equivalent to organ dose equivalent were calculated for all organs of interest^ For the liver and the bone, the assumption that all the activity was plutonium-239(W) was found to be conservative. For the lung, however, there were two radionuclides (uranium-233 and californium-252) which have higher conversion factors. To account for this difference, the conversion factor from effective to organ dose equivalent for the lung was adjusted based on the anticipated concentrations of these two radionuclides in the waste. One additional adjustment had to be made. The risks of bone cancer are expressed in terms of the mean bone dose. The organ dose equivalent conversion factor used for bone in this SEIS considers the endosteal cells only. To calculate risks, the mean bone dose risk estimator has to be converted to an endosteal dose risk estimator. The conversion was accomplished using the bone dosimetry model published by the International Commission on Radiological Protection (ICRP. 1979). Once these conversions are made, the number of excess fatal cancers can be predicted using the risk factors and latency periods contained in the BEIR-IV report (see Subsection N.2.1). The reassessment evaluated risks from both routine WIPP emissions and postulated accidental releases. For routine emissions, the reassessment follows a cohort of people (evenly distributed between the two sexes) through a 109-year lifetime. A People in this cohort are assumed to be simultaneously liveborn at the time the WIPP goes operational. The cohort is exposed to radioactivity/radiation for the 25 years of WIPP operations. The first 5 years are associated with the WIPP's Test Phase. The remaining 20 years are associated with the WIPP's Disposal Phase. For each year of the cohort's lifetime, the lifetable takes the following steps: 1) Given the population existing at the beginning of the year, the total background mortality, the total background cancer mortality, and the background mortalities for bone, liver, and lung cancer are calculated. N-7 2) The high-LET annual effective dose equivalents associated with the WIPP are converted into bone, liver, and lung dose equivalents, and the number of predicted excess fatal cancers is calculated based on those dose equivalents and the starting population. Latency periods are built into the calculation for each type of cancer. 3) The low-LET annual effective dose equivalent associated with the WIPP is converted into an annual predicted number of excess fatal cancers using the starting population and the dose equivalent (if any). The risk in subsequent years due to a given year's detriment (the actual external plus the CEDE) is corrected to reflect the decrease in the cohort population over time. A latency period is also built into this calculation. 4) The population surviving at the end of the year is calculated by subtracting the background mortality and the predicted numbers of excess fatal bone, liver, lung, and low-LET cancer from the population living at the beginning of the year. At the end of the 109-year lifetime, the excess number of fatal cancers was totalled. The reassessment also calculated predicted excess fatal cancers from effective dose equivalents received by individuals during postulated accidental WIPP releases. The reassessment follows a cohort of people (evenly distributed between the two sexes) through a 109-year lifetime. All people in this cohort are assumed to be simultaneously liveborn at the time of the postulated accident and exposed to radioactivity/radiation from the accident event. Deaths are calculated as described above for routine operations. At the end of the 1 09-year lifetime, the excess number of cancer deaths was totalled and divided by the number of people assumed for the cohort to arrive at the excess fatal cancer risk to an individual. N.3.2 SCENARIOS SELECTED In order to make health effects comparisons between results obtained utilizing the SEIS methodology and those calculated using the more rigorous approach described above, four dose consequence calculations were selected. These four calculations are not all inclusive but are representative of the full range of exposure pathways, radiation types, and individual and population assessments addressed by the SEIS. 1) The collective CEDE received by the off-site population during normal operations (see Table 5.23) 2) The collective CEDE received by the WIPP's employee population (waste handling crew) during normal operations (see Table 5.24) 3) The highest predicted CEDE to a member of the public, that is associated with postulated accident C-1 (see Table 5.28) N-8 4) The highest predicted CEDE to a WIPP employee, that associated with postulated accident C-3 (see Table 5.28). For each of these scenarios, the total number of predicted fatal cancers was calculated. Similar values for excess fatal cancers were calculated based upon the SEIS health effects estimates of 280 fatal cancers per million person-rem of population detriment. N.3.3 PUBLIC HEALTH EFFECTS The total numbers of predicted excess fatal cancers using the two assessment methodologies are shown in Table N.3.1. The table shows that the estimated health effects associated with WIPP operations as reported in this SEIS overstate estimates obtainable from the latest available recommendations for assessing human health effects associated with radiation exposure. An example of the lifetable analysis is presented in Table N.3.2 for the population risk resulting from routine WIPP emissions. TABLE N.3.1 Estimated excess fatal cancers caused by WIPP operations during the Test and Disposal Phases Scenario SEIS methodology BEIR-IV methodology Off-site population due to routine WIPP emissions^ WIPP employee population during routine WIPP operations^ Maximum off-site individual due to postulated WIPP accident C-10 Maximum worker due to postulated WIPP accident C-3 6.8 X 10'^ 1.0 X 10 -1 4.8 X 10 1.7x1 0" -4 3.2 X 1 0' 4.1 X 10" 1.6 X 10" 5.6 X 10' 3 Population risks are expressed as the total number of excess fatal cancers in the entire population. Individual risks are most easily interpreted as the excess risk of an individual contracting a fatal cancer (e.g.. 4.8 x lO'^ represents 48 chances in 100,000). ^Off-site population is 112,966 people living within 50 miles of the WIPP. ° Employee population is 18 radiation workers. N-9 o if ■^ O T- »-■ o ^• ^ o ^ ^' O .r^ o y ^ o »- ^" o ^' o Ji .- o ». -^ ci .-: o ^ d ^ r. P <» o «■ o % -o P h. ■^ -o 9, ■? ° o p ^ « fi 2 fi E e I S| 8.E 8. ^1 2 S a • 3ii| Jj( ? 8£ ^§1 iill fi^ 11 f^l -t '^ -o "o o "o o ° ° o 2 o Q u u u o CM i:°°ooooooo CM « CM M M 2°°222200ooooo r^C\(CMCMCMC>4CMCM CM c\i CM CM CM CM 'h '=i> % % "h 2222000000000 Of (4 CM CM CM CM 'o % 1 -o % % 2 2 o O) ot o> o> o> o> =i> % p ■» " o O o ■o o O) Ji ^ V! " N-10 II ►-8 35 -o-oIj'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^'^ -■ rj ^ n r^0-0'0.0-«KO<>><><>'»<>'<>'<>'<>'?'*<>'?'?'?' "^ 'o 'o o 'o 'o 'o o 'o 'o 'o o o o o o o o o o o o o „0'-«o«>«o s S s S S M § ^ S s t M. S 8 M. § ? § S S 8 8 8 8 8 8 8 8 8 8 8 8 8 8 S 6 S & & & 8 8 8 'o 'o o o o o 'o 'o 'o o o o o 2 g 2 o '° 2 2 2 2 9 XMKKXXXXKXKXKKKXKKXXWXK h.o>»-c9**o>0)0>o>o>o>0)0>0) ^ ^ ^ O O ^ O ^ o w O O O O O O O O O O O O O XXXXXXXXXXx^XXXXXXXXXXX OOOOOOOqPP^tujv|CJMNf>lmiA»niAiAmiAiniAifiiAinirt»r%ir>tA 'o o o o o o o o o o o o o 2 o o o o o o o o o XXXKKKKXKKXXKXXXXXXXXXK jN t\if>lf\iNC^fJlrJMM o 'o 'o o o o 'o o 'o 'o o o o 'o o o 'o o o o o o o XXXXXXXXKXXKXXXXXXXKXKX 'o o "o O O O O o o o o o o o 'o o 2 2 2 2 2 2 2 xxxxxxxxxxxxxxxxxxxxxxx 'o 'o O o o o o o o o o o o o o o o o o o 2 2 2 xxxxxxxxxxxxxxxxxxxxxxx cj»o>o>oic»o»oio>oio>o>0)d>o>o»o>o>o»o>o>o>o>o> ■O^-O-O'O'O-O'O'O-O-O-O-O'O-O'O-O'O-O-O-O-O'O o o 'o O O o o o o 'o o o o o o o o o o o o o 2 xxxxxxxxxxxxxxxxxxxxxxx doioioioioioioioioioioioioioioioioioioioioio) 'o o o o o o o o o o o o o o o o o o o 2 2 2 2 xxxxxxxxxxxxxxxxxxxxxxx O O O O O o o o o o o o 'o o o o o o o o 2 2 2 xxxxxxxxxxxxxxxxxxxxxxx o 'o o o 'o o o o 'o o o o o o o o o o o o o 2 2 xxxxxxxxxxxxxxxxxxxxxxx *«2t22aSiRi53S«8t3 ?) r> CO to rt fo O N-11 II -a 8 •£ o o q 9 o c w ?^ s. 1 1 S a. m 3 5^° 5 '^'^%'>bV1>%'^'^%'^'^'^'^'^•^%'^'^'^•^'^•^ o o ra M t^. nOTO>o>%9cno> o> o> o> o> o> o> CMCMCMCMCMCXCMNNCMCMCMNNCMCMMMCMM CMCMCMCMCMfJCMCMCMCMCMN tr\ \rk tr\ tf\ CMCMCMCMCMPjCMCMCM CM CM CM CM ( CM CV( CM CM CM CM CM M CM CM CM CM CM CM CMCMCMCMCMCMCMCMCMCNjCMCMCMCMCMCMCMCMCMCM CM CM CM CM CM CM o o o o o o o o o o o o o o o o o o o o o o o X X X X X X X X X X X X X X X X X X X X X X X c» oi oi oi oi oi Cji oi oi oi oi oi C7i oi oi 0)OtO>0)0)0)0)Ot X X X X X X X X X X X X X X X X X X X X X X X oi (ft CO S S ? 5 o>o>oto)o>wwo>o)o>o>o>o>o>o)o> 39$$!;9?S»»S;a!S»lnKS N-12 S-A*^: I in io t^ to ai o '■ ^ n n 't 't '^. '^. U '^ '^ ',-■ '■-. 'oi 'n 'r. 'o O O O O 'o O O O O O O O O p cJcJrIfJcJoJrJcJ ni oi cvi rg <\l fJ fJ rJ (\i ci (vi rv] •- eo Q ^ €0 S>- o » K n t\ ot_ ^ S S 9 v ( N Oi oi !<^ cJcJrJCNiNrJrJNoiNCMNNNCMCMCMOlOICNlCJNCa irtir»ir»ir»u>inmu^ir»inirt»nir»t/%irtm»r>inirtinini^»n o o o o 'o o o o o o o o o o o o o o o o o o o xxxxxxxxxxxxxxxxxxxxxxx 0ICilCg«NN(0(NNOjCg<>IO(Oj)NN Nc3<\i«>JN01CJcvi«WP'M»>lNol0J o> o> o> O) evjeMe>jeMevje>je>4eMCMeMeMevieMeMCM»>jO) cgegeMcgegegegegegegcgegcgcgcgcMeM t^ a eM Ol CM (M eg eg eg eg ! MeMeMeMeMe>jev» egegegegegeMegcgeg e>( CM e\j e\| e\( CM CM eg CM N CM eg CM CM CM CM CV) CMCVieMCMeMCMCNjCMCMCMCMCMCMeMCMeMCMOl egcMcgcgcgegeMcgcMcgcgegcgegegcgegcg 0| CM eg eg eMCMeM^leMeMeMCMeMCMCMCMCMeMCMe^Jc^lCMC^lCMe^I^lcM cgegcgcgcgcgcgcgcMcgcgegcgcgcgcgcgegcgegegegcg CVleMCMCMCMCMCMCMNCMtMCMCMCMCMCMCMCMCMCMCMeMCM cMCMeMcgcgcgcgegegcMcgcgcgcMtgrgcgcgcgcgcgcgcg X >< X K X X X X X X X X X X X X X X X X X X X oi o> oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi oi o •« •o •O •o •o •o •e •o -o -o •o •o •o ■o -o •o >o ■o ■o •o ■c -o o o o o o o o o o o o o o o o o o o o o o o o X X X X X X X X X X X X X X X X X X X X X X S 3 8 o S S 8 N-14 s->rx< UJ O 17 I "o "o t t ol - M Cj s s a 2 .» >» s» ^ o o o 'o II o> C* Ot ^ CM N CM N trt i/> lA in CM CM CM r^ CM csi rj CM CM CM (M M CM CM c\i rti CM r^ o> o> o> en Ot Ot Ot Oi 2 £ • ■" ^ C 3 lit ^•5 ■D 6. " c - ^ ? I ill N-15 N.3.4 GENETIC EFFECTS The references mentioned in Subsection N.2 also discuss the genetic effects of radiation exposure. Based on the data currently available, the following genetic risk factors for subsequent generations apply to WIPP radiation doses: Type of Radiation Low-let (UNSCEAR, 1988) High-LET (BEIR, 1988) Genetic Risk Factor (per million live offspring per rad) 120 600 Using these risk factors, the genetic risk caused by WIPP emissions (both routine and accidental) were calculated. For high-LET radiation, the calculation involved three steps: 1) Converting the CEDE for each scenario into a committed dose equivalent (CDE) to reproductive organs (testes and ovaries) 2) Dividing the CDE by the quality factor for alpha radiation (20) to convert dose equivalent to absorbed dose (rem to rad) 3) Multiplying the committed dose by the genetic risk factor to obtain the risk to subsequent generations. For low-LET radiation, the first two steps were not necessary since the dose equivalent is uniform over the whole body and the quality factor for low-LET radiation is 1 . The results of these calculations are shown in Table N.3.3. These risks were then compared with the risk of fatal cancer associated with the particular scenario. The ratio of the genetic risk to the excess fatal cancer risk is also shown in Table N.3.3. In all cases, the risk of genetic effects was less than 93% of the cancer risk. The major factor affecting the magnitude of the risk was the low-LET contribution. For the transuranic elements present in the waste at the WIPP, the CDE to reproductive organs is a fraction of the CEDE. This fact and the large quality factor for alpha radiation were the principal reasons for the lower contribution of high-LET radiation. These results support the conclusion made in this SEIS that the risk of fatal cancer provides the most conservative measure of the health effects caused by WIPP operations. N-16 TABLE N.3.3 Estimated excess genetic effects caused by WIPP operations Scenario Off-site population due to routine WIPP emissions WIPP employee population during routine WIPP operations*^ Maximum off-site individual due to postulated WIPP accident C-10 Maximum worker due to postulated WIPP accident C-3 Ratio of excess Excess genetic effects genetic effects^ to excess fatal cancers^ 5.6 x 10-^ 0.02 P 3.8x10'2 0.93 7.1 x 10' 2.6 x 1 0" 0.04 0.05 ^ Population risks are expressed as the total number of excess genetic effects appearing in live-born offspring in all future generations of the exposed population. Individual risks are most easily interpreted as the excess risk of a genetic effect appearing in the live-born offspring in all future generations of the exposed individual. ^ Excess fatal cancers taken from Table N.3.1, BEIR-IV methodology. The ratios presented compare to the 0.918 risk estimator used in this SEIS. ^ Off-site population is 112,966 people living within 50 miles of the WIPP. ^ Employee population is 18 radiation workers. N-17 REFERENCES FOR APPENDIX N BEIR, 1989. Health Effects of Exposure to Low Levels of Ionizing Radiation . National Academy of Sciences, Committee on the Biological Effects of Ionizing Radiations. BEIR, 1988. Health Risks of Radon and Other Internally Deposited Alpha Emitters . National Academy of Sciences, Committee on the Biological Effects of Ionizing Radiations. BEIR, 1980. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 1980 , National Academy of Sciences, Committee on the Biological Effects of Ionizing Radiations. Dunning Jr., D.E., 1986. Estimates of Internal Dose Eguivalent from Inhalation and Ingestion of Selected Radionuclides . WIPP-DOE-176, Rev. 1, prepared for the U.S. Department of Energy. ICRP (International Commission on Radiological Protection), 1979. Annals of the ICRP . ICRP Publication 30, Limits for Intake of Radionuclides by Workers, Pergamon Press, Washington, D.C. ORNL (Oak Ridge National Laboratory), 1980. Dunning Jr. D.E. et al, A Combined Methodology for Estimating Dose Rates and Health Effects from Exposure to Radioactive Effluents . ORNL/TM-7105, December, 1980. UNSCEAR, 1988. United Nations Scientific Committee on the Effects of Atomic Radiation, Sources. Effects, and Risks of Ionizing Radiation . 1988 Report to the General Assembly, with Annexes. N-18 ^^SK^'-y y APPENDIX O TEST PLAN SUMMARY 0-i/ii i \ ^^;<%i TABLE OF CONTENTS Section Page 0.1 INTRODUCTION 0-1 0.1 .1 Executive Summary 0-2 0.1.1.1 Objectives of the WIPP Test Phase 0-2 0.1.1.2 Description of Test Phase Activities 0-3 0.1.2 Background 0-6 0.1 .3 Proposed Testing 0-8 0.2 APPROACH 0-11 0.3 TEST DESCRIPTIONS 0-15 0.3.1 Bin-Scale Tests 0-15 0.3.1.1 Bin-Scale Test Objectives 0-16 0.3.1.2 Bin-Scale Test Summary 0-17 0.3.1.3 Bin-Scale Test Phases and Schedule 0-21 0.3.1.4 Bin Preparation and Transportation 0-21 0.3.2 Alcove 0-22 0.3.2.1 Alcove Test Objectives 0-22 0.3.2.2 Alcove Test Summary 0-23 0.3.2.3 Alcove Test Phases and Schedule 0-28 0.3.2.4 Waste Preparation and Transportation 0-29 0.4 UNDERGROUND TEST OPERATIONAL SAFETY 0-30 0.4.1 Emplacement Safety Concerns O-30 0.4.2 Test Operational Safety Concerns 0-31 0.4.3 Mine Safety Concerns 0-32 0.5 POST-TEST OPERATIONAL SAFETY 0-33 0.5.1 Bin Retrieval 0-33 0.5.2 Alcove Retrieval 0-34 REFERENCES FOR APPENDIX O 0-35 0-iii LIST OF FIGURES Figure Pafl© 0.3.1 Location of bin-scale test, plan view 0-18 0.3.2 Location of test alcoves, plan view 0-24 LIST OF TABLES Table Page 0.3.1 Estimated number of bins 0-19 0-iv 0.1 INTRODUCTION This appendix describes the underground tests using TRU waste proposed at the WIPP during the Test Phase. This appendix has been prepared in response to comments that requested additional details on the proposed Test Plan, especially as to how the Test Plan relates to the Proposed Action. As noted in Subsection 3.1.1.4, the initial step of the Proposed Action is to conduct a Test Phase of approximately 5 years. The Test Phase has two distinct elements: 1) the Performance Assessment and 2) the Integrated Operations Demonstration. These elements continue to evolve. At this time, the Performance Assessment tests using TRU waste would be composed of laboratory, bin-scale, and alcove tests, and plans on such issues as waste source, type, and volumes for the initial phase of tests are nearing finalization (DOE, 1989a). Waste requirements for the integrated operations demonstration remain uncertain. The DOE, in December 1989, published a detailed phased plan for the Test Phase (DOE, 1989a) that focused on the methods and activities required to demonstrate compliance with the long-term performance standard of 40 CFR 191, Subpart B. In addition, several of the tests planned for the Test Phase would provide data that would be used to support WIPP's demonstration that there would be no migration of hazardous constituents of the waste, as required under the RCRA Land Disposal Restrictions (40 CFR 268). A separate, detailed plan would be developed to describe in detail the Integrated Operations Demonstration. As discussed below, the DOE believes that the analyses in this SEIS bound the potential impacts that would be estimated to arise from any such waste requirements decision. During the Test Phase, the DOE proposes to transport to and emplace in the WIPP limited quantities of waste; the specific quantities of waste emplaced would be limited to that deemed necessary to achieve the objectives of the Test Phase. For purposes of bounding the potential impacts of the Test Phase in this SEIS, the DOE assumes that up to 10 percent of the volume of TRU waste that could ultimately be permanently emplaced at the WIPP would be emplaced during the Test Phase. The actual amount of waste proposed for the Test Phase would likely be less than that assumed for purposes of analysis in this SEIS. It is also assumed for purposes of bounding the impacts that waste would be shipped from all 10 facilities, although it is now likely that only waste from Rocky Flats Plant and the Idaho National Engineering Laboratory would be used during the initial phases of the proposed Test Phase. Subsets of the Proposed Action include conducting the Test Phase with bin-scale and/or alcove tests without the Integrated Operations Demonstration and the conduct of these tests with lesser volumes of waste than assumed in the SEIS. The impacts of these subsets would be bounded by the analysis of the Proposed Action in this SEIS. 0-1 0.1.1 EXECUTIVE SUMMARY The following has been derived with modification from the Executive Summary of the proposed Test Plan (DOE, 1989a). 0.1.1.1 Objectives of the WIPP Test Phase The purpose of the Test Phase is to further the intent of Congress to demonstrate safe and environmentally acceptable disposal of defense wastes and thereby establish a permanent disposal facility for TRU wastes. The activities that will provide the needed information include experiments, analyses, and operations at the WIPP facility. Although the initial part of the Test Phase is well defined, experimental programs will evolve with increasing understanding of the systems under test. The nature, scope, waste quantities, and timing of experiments and full-scale rooms recommended by various groups remain flexible. The sum total of waste for these tests would initially require approximately 2 percent by volume of the design capacity. The initial plans for the Test Phase described in this document call for the emplacement of approximately 0.5 percent by volume of the design capacity for Phases 1 and 2 of the alcove tests and Phases 1 and 2 of the bin-scale tests. These bin-scale and alcove tests will support assessment of compliance with the EPA Standard, 40 CFR 191, Subpart B, Sections 13 and 15, and the RCRA Land Disposal Restrictions, 40 CFR 268, Section 6. Additional tests will be defined based on the data acquired during the first two phases of the bin-scale and alcove tests and to incorporate potential engineered alternatives. In addition, the EPA has requested that the Project monitor the performance of the facility by emplacing waste in 2 full-scale, instrumented, backfilled, sealed rooms after an appropriate demonstration of retrieval using simulated waste. Waste requirements for these 2 full-scale room tests would be approximately 1.5 percent by volume of design capacity. The DOE will conduct a feasibility evaluation to determine the best technical approach, scope, and timing of such monitoring. The DOE will consult the NAS/NAE WIPP Panel, the EPA, the State of New Mexico, and the EEG prior to initiation of such tests. Also, waste requirements for an Operations Demonstration have not yet been determined. As suggested by several reviewers, the DOE will evaluate the operational experience to be gained through the conduct of all of the test activities and will factor this into future decisions on the scope and timing of an Operations Demonstration. Waste emplaced in the WIPP during the Test Phase would be retrievable until the DOE decides whether the WIPP should become a disposal facility. During the Test Phase, per agreement with the State of New Mexico, the WIPP would meet the applicable requirements of the EPA Standard, 40 CFR Part 191, Subpart A. The two primary objectives of the Test Phase are to demonstrate the following: 1) Reasonable assurance of compliance of the WIPP disposal system with the long-term disposal standards of the EPA Standard, 40 CFR Part 1 91 , Subpart B, Sections 13 and 15. Compliance of the disposal system would be 0-2 determined based on a performance assessment, which would include an analysis of the WIPP disposal system design and an evaluation of potential engineered alternatives. 2) The ability of the DOE TRU waste management system to safely and effectively certify, package, transport, and emplace waste at the WIPP in accordance with all applicable regulatory requirements. Acceptability of the waste management system would be evaluated by operations testing and monitoring, both individually and collectively, of the elements of the TRU waste management system. The Operations Demonstration program will be presented in greater detail in a separate document. These objectives are consistent with the Congressional guidance to demonstrate the safe and environmentally acceptable disposal of TRU waste. In addition, several of the tests planned for the Test Phase would provide data that may also be used to verify the WIPP's demonstration that there would be no migration of hazardous constituents of the waste, as required under the RCRA Land Disposal Restrictions, 40 CFR Part 268, Section 6. 0.1.1.2 Description of Test Phase Activities The objectives would be accomplished by completion of two important programs: a Performance Assessment and an Operations Demonstration. These two programs would provide the necessary information to determine compliance of the disposal system with applicable environmental requirements and to evaluate the safety and effectiveness of the TRU waste management system operations. Although Subpart B of 40 CFR Part 191 was vacated and remanded to the EPA by the U.S. Court of Appeals for the First Circuit, this Plan (DOE, 1989a) addresses the Standard as first promulgated. The 1987 Second Modification to the Agreement for Consultation and Cooperation between the DOE and the State of New Mexico (1981) commits the WIPP project to continue the performance assessment planning as though the 1985 Standard remained in effect. Compliance plans for the WIPP would be revised as necessary in response to any changes in the Standard. 0.1 .1 .2.1 Performance Assessment . The performance objective for the WIPP disposal system is to adequately isolate TRU waste from the accessible environment; the performance requirements are reasonable assurance of compliance with the 10,000- year release limits and the 1,000-year dose limits of the EPA Standard, 40 CFR Part 191, Subpart B, Sections 13 and 15. The 10,000-year performance assessment would predict cumulative releases of radionuclides to the accessible environment resulting from both disturbed and undisturbed performance of the disposal system. The 1 ,000-year assessment would predict annual doses to members of the public in the accessible environment resulting from undisturbed disposal system performance. It would not address the concentration limits established by Subpart B for special sources of groundwater, because no such sources exist at the WIPP. In evaluating compliance with Subpart B, the guidance provided in Appendix B of the Standard would be followed. To ensure that all plausible responses are identified, scenarios would be developed by coupling the individual events and processes that occur. These scenarios 0-3 would be screened on the basis of probability, consequence, physical reasonableness, and regulatory interest. Consequence analysis would be used to calculate a performance measure for each of the remaining significant scenarios. The performance measures for the scenarios would be normalized, summed, and reported as a "complementary cumulative distribution function" of release probabilities. Uncertainties in the data would be included in calculations of the performance measure for each scenario. To show that the WIPP can meet the annual dose limits set for 1 ,000-year performance, the Standard requires that releases from the undisturbed scenarios be analyzed. If any release to the accessible environment is predicted, transport along biological pathways would be modeled, and doses would be estimated. Uncertainties in the data would be included in the dose calculations. The performance assessment process would be divided into five elements: scenario screening, repository/shaft system behavior and performance modeling, controlled area behavior characteristics and performance modeling, computational system development, and consequence analysis. The combined repository/shaft system and controlled area represent the disposal system that would be assessed. 0.1.1.2.2 Disposal System Characterization Activities . Accurately simulating behavior of the disposal system requires data derived from experiments conducted in the laboratory as well as in the WIPP underground. Such scientific investigations have been conducted since 1975. These studies have resolved many technical issues and have focused attention on aspects still requiring investigation. There are four major areas of scientific investigation integral to the assessment of disposal system performance. These areas examine the behavior of the disposal room and drift system, the sealing system, structural and fluid-flow behavior of the Salado Formation, and non-Salado hydrology and radionuclide migration. Investigation of these areas involves both laboratory and large-scale underground tests. Disposal room and drift system activities would examine the interaction of TRU waste and backfill in a waste room. The combined interactions of the source term, waste containers, emplaced backfill and admixtures, brine inflow, and gas generation would be studied through laboratory testing, modeling, and in situ testing. The behavior and performance of possible backfills and additives to be emplaced in access drifts as part of facility decommissioning would also be investigated. An important parameter of the disposal room and drift system is gas generation. Gaseous products would be generated by microbial and radiolytic decomposition of the TRU waste and corrosion of the waste and waste containers. Gas generation tests with actual TRU waste would be required to characterize the behavior of the disposal system under realistic conditions. These tests would consist of laboratory tests using radioactive and nonradioactive simulated waste, three phases of bin-scale tests with CH TRU waste, and two phases of alcove tests with CH TRU waste. These tests would provide the data needed to evaluate the effects of gas generated by the waste in realistic environments for both the operational (short-term) period and the postoperational (long-term) period. The information collected in these tests would aid 0-4 the performance assessment in establishing a sufficient level of confidence in the consequence analysis to demonstrate compliance with the EPA Standard. The waste quantities required for these tests represent approximately 0.5 percent by volume of the WIPP disposal area design capacity. In addition to supporting the Performance Assessment Program, the gas generation tests would provide information to be used to verify the RCRA No-Migration Variance Petition's demonstration that the hazardous constituents will not migrate. Sealing system activities would examine seal design, system behavior, and overall performance evaluation. Seals would be developed for use in drifts to isolate waste panels, in access shafts to isolate the repository from the accessible environment, and in exploratory boreholes. Laboratory and in situ tests would evaluate behavior of potential seal materials such as crushed salt, salt/clay mixtures, and concretes. The effect of hazardous constituents of the waste on seal components would also be tested. Studies of structural and fluid-flow behavior of the Salado Formation would improve the capability to model fluid flow, hydrologic transport, waste room and drift response, and shaft closure. Healing of fractures in the disturbed zone outside excavations and around seals in shafts and access drifts would be evaluated by modeling. Effects of brine on salt creep would be examined. Laboratory and in situ tests would provide data for improving models of excavation closure, fracture behavior, permeability, and fluid-flow characteristics of the Salado Formation, and brine inflow to excavated rooms. A wide range of studies would address the behavior of penetrations through the Salado Formation, openings at the repository level, and fluid flow to and through these disturbances in the host rock. The non-Salado hydrology and radionuclide migration activities would address transport of waste to the Rustler Formation and in the Rustler Formation under present and future conditions. Laboratory studies of sorption and retardation in the Rustler Formation would be included, as well as in situ geophysical and hydrological tests from the surface. In conjunction with the performance assessment, potential engineered alternatives to the current waste disposal system design would be examined. This examination would prepare the DOE to implement any necessary changes to the design in a timely manner as a contingency if performance assessment results have a high degree of uncertainty or are unsatisfactory, or if changes are required to enhance the demonstration of no migration as required under RCRA. Examples of alternatives under consideration are waste processing, changes in the waste disposal room or panel configuration, and passive markers. Engineered alternatives would be screened for relative effectiveness using a design analysis model, and would be screened for feasibility with respect to cost, state of technology, regulatory concerns, and worker exposure. The bin-scale tests, which would use actual radioactive waste underground at the WIPP, would be scheduled in three phases. Engineered alternatives that pass initial screening would be tested in Phase 3, and if identified early enough, in Phases 1 and 2. Alternatives that seem effective and feasible would then be evaluated using the formal performance assessment process to quantify the improvement in disposal system performance. 0-5 0.1.1.2.3 Operations Demonstration . The purpose of the Operations Demonstration Program is to demonstrate safe and effective emplacement of certified waste at the WIPP facility. A separate document would be developed to describe the Operations Demonstration following the Secretary of Energy's decision as to the scope and timing of the program. Key elements of the Operations Demonstration would be waste certification and packaging at the generating/storage facilities, the operation of the transportation system, and operation of the WIPP. This demonstration would be integrated to include all elements of the TRU waste management system and would require both OH and RH TRU waste operations. Operational data needs include results from the evaluation of the safety, environmental adequacy, and effectiveness of operations that would certify, transport, and emplace waste at the WIPP. In addition, operational data would be derived from the experience gained during mock demonstrations of bin and drum emplacement and retrieval, and the emplacement of actual TRU waste for bin-scale and alcove experiments underground at the WIPP. The goal of the Operations Demonstration is to provide assurance that operations can be conducted within the limits of all applicable regulatory, technical, industrial, and managerial criteria. 0.1.2 BACKGROUND TRU waste proposed to be disposed of at the WIPP is contained in a mixture of standard 55-gal (208 L) drums and standard waste boxes (SWB). The waste results from nuclear weapons research and production. It consists of laboratory hardware (such as ring stands and other metal structures, and glassware); other laboratory waste (such as Kimwipes, tissues, and towels); protective gloves and clothing; chemicals and inorganic process sludges (generally stabilized with cement); plastic, rubber, and resin; worn-out engineered equipment and tools; and residual organic compounds. The processes by which gas may be generated include microbial action, corrosion, and radiolysis. In the short-term, these gases are generated predominantly from radiolytic degradation of the waste, and include hydrogen, oxygen (rapidly depleted in most cases), carbon oxides, and low-molecular-weight organic compounds (Zerwekh, 1979; Kosiewicz, et al., 1979; Kosiewicz, 1981; Molecke, 1979). Radiolysis of water and potentially intruding brines could also generate appreciable quantities of hydrogen (and oxygen) in the postoperational and long-term time periods. Microbial degradation mechanisms may be a major concern in both the short- and long-term time periods (Caldwell, et al., 1987; Molecke, 1979). Microbially generated gases include carbon dioxide or methane (Caldwell, et al., 1987; Molecke, 1979), potentially nitrogen from denitrification of nitrates, and hydrogen sulfide from sulfate-reducing bacteria (Brush and Anderson, 1988). Anaerobic (anoxic) metal corrosion in the postoperational and long- term periods could also generate signification quantities of hydrogen (Brush and Anderson, 1988; Molecke, 1979). No radioactive gases would be generated, with the exception of radon (T.|/2 = 3.8 days) from the decay of transuranic isotopes in the wastes. No radioactive particulates would be released, because the drums would be vented through HEPA filters. The potential for gas generation in the WIPP and its effect on the long-term performance of the repository is a primary focus of the gas generation test program. 0-6 WIPP waste emplacement operations for permanent disposal would include placement in rooms and entries within the eight panels; the rooms would be backfilled with an appropriately designed material. After being filled with containers of waste and backfilled, the panels would be sealed from the rest of the underground facility. Any net gas generated by the waste after a panel is sealed must be considered in the long- term performance assessment calculations. The performance of the WIPP disposal system includes not only the room behavior, but also the individual and coupled behavior of the panel seals, access drifts, shaft seals, disturbed zones in the rock around the excavation, and potential transport of radionuclides and hazardous waste through the upper water-bearing units to the accessible environment. Since the 1980 FEIS, changes in the understanding of factors that affect long-term performance have occurred. These are described below. • The Salado Formation is probably hydraulically saturated, with very low effective permeability in undisturbed regions. At the time of the FEIS, it was thought to be hydraulically unsaturated, with sufficient gas permeability to dissipate any gases that might be generated by emplaced waste. Thus, the estimated far-field permeability of the Salado Formation has decreased since 1 980. • Current estimates of total gas generation from degradation of emplaced waste and containers are smaller than similar estimates in the FEIS, although uncertainties exist in gas-generation rates, total volumes of gas generated, and the time periods over which gas generation might occur. • Decreased far-field permeability suggests that the WIPP repository following closure may be dominated by gas at elevated pressure, with little or no free brine within the workings. • The volumes of gas potentially generated, even in the absence of free brine, may exceed the gas-storage capacity of the waste emplacement rooms at their final state of closure under lithostatic pressure. Gas storage (or relief of pressures) is possible through 1) an expansion of the rooms, after closure, to something less than their original volume; 2) generation of a secondary zone of increased porosity from fracturing around the waste emplacement rooms, or in an incompletely removed disturbed rock zone; 3) migration of gas along open fractures within Marker Bed 139, within or around panel seals, and perhaps within stratigraphic contacts at and near the repository horizon; and 4) following transport from the panels, migration of gas into the shafts and adjacent marker beds. Thus, laboratory, bin-scale, and alcove tests are proposed to evaluate the effects of gas generation and consumption. These tests are intended to collect, interpret, and refine data necessary for performance assessment. The data resulting from the tests would reduce uncertainty in the performance assessment by verifying assumptions and providing input data on gas generation, gas depletion, and aqueous radiochemistry. 0-7 0.1.3 PROPOSED TESTING The laboratory tests would use only simulated waste (nonradioactive) or spiked waste containing a single radionuclide to assess radiolysis and effects of compaction. This appendix addresses only underground tests using actual TRU waste; a brief description of the laboratory tests is presented in the Draft Final Plan for the Waste Isolation Pilot Plant Test Phase: Performance Assessment (DOE, 1989a). The bin-scale tests would use CH TRU waste specially prepared and modified to provide both repository relevant gas and brine-leachate radiochemical data. (Bins are specially produced, instrumented containers that will hold the equivalent of about 6 drums of CH waste.) The bin-scale tests would confirm and extend similar past and current laboratory test results. Bin-scale tests would provide the results of a scaled verification and evaluation of the impacts of synergistic waste degradation, gas- generation modes, and the effectiveness of backfill additives designed to consume gases ("gas getters"). These tests would include a range of environments: wet, dry, with oxygen, without oxygen, backfilled with gas getters, and backfilled without gas getters. The alcove tests would use a mix of unmodified (as received) and specially prepared CH TRU waste to obtain information on the operational phase conditions and on the long-term, postoperational phase conditions. Alcove tests are the only experiments planned that can incorporate the impacts of the actual repository environment on the degradation behavior of the waste. The repository impacts are expected to include gases released from the host rock salt (e.g., nitrogen) intermixing with or influencing waste degradation modes; brine influx and consequent humidity effects; long-term waste compaction; and total encapsulation of the waste containers by backfill containing gas getter materials. The gas generation experiments would not include RH TRU waste. Experiments with CH TRU waste are expected to bound any effects of RH TRU waste, for two reasons. First, the repository would contain 4,000 to 5,000 RH canisters with an average radionuclide content of 37 curies per canister (DOE, 1989c; Table 3.3 in this SEIS). Thus, the maximum RH loading is expected to be 185,000 Ci, only 2 percent of the initial CH loading. Half of the RH radionuclides are short-lived, with half lives of less than 30 years. Second, RH TRU waste would be emplaced in individually drilled and sealed boreholes in the pillars, not in the waste panels proper. Preliminary calculations suggest that these boreholes will creep closed in about 10 years, making waste inaccessible to brine intrusion and degradation (Lappin et al., 1989). Underground testing would provide data necessary to conduct the performance assessment with a sufficient degree of confidence. Previously, gas generation was not considered a critical factor in the long-term performance of the WIPP. Calculations of gas transport out of the repository into the surrounding Salado Formation (DOE, 1980b; Sandia, 1979) suggested that permeability of the Salado Formation was high enough to allow gas to dissipate without a significant increase in repository pressure, even if the high gas production rates estimated by Molecke (1979) as upper bounds were applicable. Recent, more definitive far-field permeability calculations (Tyler et al., 1988), indicate that permeability of the Salado Formation is low enough that the 0-8 anticipated high gas production rates may significantly pressurize the repository. Thus, improved understanding of parameters such as gas generation and the repository system (backfill and host rock) behavior have become necessary to establish a realistic range of gas production rates for WIPP. Available estimates of the rates of gas production by CH TRU waste are based on laboratory studies of processes such as radiolysis, microbial activity, corrosion, thermal degradation (Molecke, 1979), and field studies of gases accumulated in the tops of drums (headspace gases) (Clements and Kudera, 1985). Another investigation of gas generation processes was reported by Brush and Anderson (1988). It was concluded that processes such as drum corrosion, microbial decomposition of cellulosic materials, and reactions between drum corrosion products and microbially-generated gases could affect the gas and water budget of the repository. These processes could consume or produce quantities of water similar to the quantities of brine that are expected to seep into the repository from the Salado Formation. The Performance Assessment must address the gas and water content of the disposal rooms because these factors could affect long-term performance calculations, especially in the human intrusion scenarios. However, obtaining gas production data representative of the total waste mix is difficult due to the extreme heterogeneity of CH TRU waste, which is the result of the wide variety of generating waste streams. A test program that will be representative would require a large number of experiments and, in large-scale tests, a significant and representative sample of the total waste inventory. Bin-scale and alcove tests are thus necessary to acquire the data for predictions of long-term gas and water content of WIPP disposal rooms and to assess their impact on repository performance. It is evident, based on all previous investigations, that a proper understanding of gas generation rates and quantities is critical to predicting the behavior and ultimate state of the repository. The TRU waste tests described in this appendix are designed to provide that understanding and help establish an acceptable level of confidence in the prediction of repository performance. These tests would also help in establishing whether modifications to the design of the disposal system are needed. Rates of gas consumption, normally controlled by radiolysis, microbial degradation, and corrosion, can presumably be increased by including gas getter materials as a backfill component. In addition, anoxic corrosion reactions that generate hydrogen require and consume water in the process. Thus, modifying the disposal room design to minimize brine inflow may limit hydrogen generation. In addition, the testing program would collect data to support WiPP's RCRA No Migration Variance Petition. Key aspects of the gas testing program related to RCRA compliance are: • To identify any hazardous components (such as volatile organic compounds) that may be released from waste. 0-9 • To gain greater understanding of potential chemical interactions that may occur between various waste types and between waste and repository host rock, brine, and alternative backfill and gas getter materials. • To observe and report on waste and repository behavior to meet monitoring requirements related to the granting by the EPA of a No-Migration Variance for the WIPP. Air monitoring of all potential releases from the bin and alcove experiments would be conducted throughout the Test Phase. • To evaluate through a combination of modeling and experimental studies, the expected structural and fluid-flow response of WIPP to internal gas pressurization. • To evaluate the potential for degradation of the seals and plugs (final design, not temporary inflatable seals) due to exposure to the volatile organic compounds in the waste. In conjunction with the performance assessment activities, the Project will examine engineered alternatives to the current waste disposal system design. It will prepare the Project to implement any necessary changes to the design in a timely manner as a contingency if performance assessment results have a high degree of uncertainty or are unsatisfactory, or if changes are required to enhance the demonstration of no migration as required under RCRA. Examples of types of alternatives under consideration include waste processing and changes in the storage room or panel configuration. Engineered alternatives will be screened for relative effectiveness using a design analysis model and will be screened for feasibility with respect to cost, state of technology, regulatory concerns, and worker exposure; they will then be tested in laboratory or larger scale experiments where possible. Phase 3 bin-scale tests will incorporate appropriate alternatives, and it is possible that some alternatives will be identified early enough to include them in Phases 1 and 2. Potentially effective and feasible alternatives will be evaluated using the formal performance assessment process to quantify the improvement in disposal system performance. 0-1 0.2 APPROACH An assessment of gas issues must consider three elements: gas production, gas consumption, and gas transport. Gas production is a function of radiolysis and chemical and biological interactions between the waste, waste containers, engineered backfill, brine, and salt. Gas consumption is controlled by the processes of radiolytic and microbial degradation and corrosion. Gas transport depends on the ability of the formation to accept the gas and allow it to disperse. The primary parameter controlling gas transport (in the absence of seal failure) is the Salado Formation gas permeability, which differs for different gases. The gas transport element can be addressed by investigations without waste, but gas production and consumption are largely functions of the waste itself; therefore, radioactive waste is needed in the testing. The approach of the bin-scale tests is to use test bins that will be large enough to contain a mixture of up to 6 drum volumes of actual CH TRU waste, drum metals, backfill materials, brine, and salt. Sources of gas generation would be introduced into the various environments created in each bin (wet, dry, with oxygen, without oxygen, with gas getters, and without gas getters). For microbial gas generation the sources would be halophilic and nonhalophilic bacteria. Drum and metallic waste materials would provide the corrosion gas source, and the radioactive component of the waste would be the source of radiolytic gas generation. The bin-scale tests would also provide an environment in which various types of gas generation may occur simultaneously. Therefore, these tests would provide a realistic, credible, and synergistic test for the gas generation rates and interactions with backfill and gas getters. Alcove tests would confirm the results of the laboratory and bin-scale tests. These tests would allow a larger, synergistic test of gas generation, waste compaction impacts, and effectiveness of gas getter material. The tests would consist of waste emplaced in five sealed, atmosphere-controlled test alcoves (each about one-quarter the volume of a waste disposal room). This testing arrangement allows lesser quantities of waste per test alcove, so that more types of test conditions can be accommodated. The waste emplaced in the alcoves would include a typical, representative quantity and mixture of waste types and waste loadings. The volume of waste required is based on both statistical evaluations and practical considerations, and is subject to change based on oversight agency concerns, initial results of the tests, modification of the tests to accommodate treated waste, and other factors. To accurately measure gas production and consumption, actual radioactive waste must be used. Data needed for the performance assessment models could be obtained from the combination of laboratory tests using small-scale, simulated waste (Brush, 1989; Zerwekh, 1979; Kosiewicz et al., 1979; Kosiewicz 1980, 1981; Caldwell et a!., 1987; Molecke, 1979), intermediate, bin-scale tests (Molecke, 1989a), and large, alcove 0-11 (field) tests (Molecke, 1989b). Resultant data from all of these experimental programs, when coupled with model development, would be used to assess the importance of gas to the performance of the repository. The laboratory-scale tests have been described in more detail by DOE (1989a) and Brush (1989). The strong interrelationship of the bin- and alcove-scale experimental programs, and the perceived benefits and disadvantages of each program are detailed below. The bin-scale tests may be viewed as larger-scale laboratory experiments, except that they would have the following advantages: 1 ) They would incorporate actual radioactive TRU waste, and also contain minor chemical components, organic compounds and solvents, and microbial contaminants which could have a very significant impact on overall gas generation and source-term radiochemistry; 2) There would be very few test simulations or required assumptions; 3) All test components, waste forms, contaminants, and possibly engineered alternative materials would be interacting in a synergistic, repository relevant environment, in which various modes of gas generation are occurring simultaneously; 4) The larger scale of the test bins, incorporating about 6 drum-volumes of waste each, would help smooth out the known nonhomogeneities among supposedly similar waste types; 5) The total test matrix could be expanded as necessary, to incorporate new waste forms, backfill and getter materials, and engineered alternatives as they are developed and are ready for testing; and 6) These tests could provide for the rapid collection of data, as compared to the alcove tests, consistent with present Performance Assessment schedules. The disadvantages of the bin-scale test program are 1) The inability to test at high gas pressures; 2) The inability to fully incorporate all repository environmental effects - as in the alcove tests; 3) The performance of bin-scale tests at the WIPP is linked to first receipt of waste; and 4) Tests can only examine limited interactions between waste types. 0-12 The in situ alcove tests would be conducted under credible, expected-case repository conditions. The major advantages of the alcove tests are 1 ) Tests would provide "real-world" data, with the fewest simulations or restraints of any of the test programs that could potentially bias the end results; 2) They would be the only tests which actually incorporate the environmental, possibly synergistic effects of the repository itself, i.e., gases and fluids released from the host rock, mine geochemistry and biochemistry, etc., on waste degradation rates and modes; 3) Assessments would determine the gas generation rates for the times of interest, and incorporate how the gases will either be produced or consumed; 4) There would be no significant scaling effects due to the size of the test alcoves; and 5) Many waste forms would be emplaced together in the same test alcove, as would be the case in an operating repository. The major disadvantages of the alcove tests are 1) The inability to test at high gas pressures because of underground facility safety concerns; 2) The limited number of test alcoves available, resulting in a limitation on test variables and test replicates that can be incorporated; 3) The combination of many waste types within each test alcove makes interpretation of the effects from each type or degradation mechanism almost impossible without comparison to other program data; 4) The large volume of each test alcove, plus the initial trapped gas (air or nitrogen), decreases the analytical sensitivity for gases of interest being produced - small changes in the quantity of produced gases may be masked; 5) The expected rates of production for individual gases, and changes in those rates, may not be clearly evident for an appreciable period of time -- when compared to gases generated and analyzed in the smaller test bins; and 6) There is no human access to the alcoves after test initiation; potential engineered modifications cannot be added after the test begins. The added degrees of experimental control, assumed increased sensitivity and selectivity for gas analyses, and the increased number of test conditions for variables to be used in the bin-scale tests ~ relative to the alcove tests ~ allows the interpretation of obtained data to be simpler and more straightforward than that from the alcove 0-13 tests. As such, the bin-scale tests provide a technically more satisfying and rapid means of obtaining data. Collecting test data from any of the test types must not be simply a monitoring or confirmatory activity. Data must be used for both analytical and predictive performance assessment modeling calculations and for comparison with smaller-scale laboratory data on simulated waste. It must be emphasized that it is the combined suite of CH TRU waste test programs (laboratory, bin-scale, and alcove) that are required to provide the full spectrum of information and expertise needed for the performance assessment program. Each test program has its own advantages and disadvantages. None of the three test programs alone can credibly produce the required information. 0-14 0.3 TEST DESCRIPTIONS A description of the proposed bin-scale and alcove tests is presented in the following subsections. The test description includes the objectives of the tests, a summary of the tests, and the transportation and emplacement operations. These descriptions are sumrriarized from Brush (1989) and Molecke (1989a and 1989b). 0.3.1 BIN-SCALE TESTS The primary purpose of the bin-scale test program is to provide relevant data and technical support to the WIPP Performance Assessment program for both predictive modeling studies and for the assessment of hazardous component release, and consequent impacts on the WIPP, in relation to EPA concerns and regulations. Specific data to be obtained include the quantities, compositions, and kinetic rate data on gas production and consumption resulting from various OH TRU waste degradation mechanisms. Similar data on potentially hazardous volatile organic compounds released by the waste and waste-brine leachate or source-term radiochemistry would also be provided. Actual OH TRU waste would be used in these tests. The degradation and interaction behavior of several representative classifications and types of waste would be tested under aerobic and anaerobic conditions representative of the Disposal Phase and the long-term, postoperational phase of the repository. Tests are intended to allow evaluation of impacts of several types and quantities of intruding brine; impacts on gas production and consumption of waste interactions with salt, container materials, backfill, and gas getter materials; and gas production resulting from synergism among various degradation modes. The tests would be controlled so that safety of personnel is maintained by the use of leak-tight bins, venting through HEPA filters, and close monitoring. In total, the first two phases of the bin-scale test program would include 116 waste- filled bins and 8 empty test bins (representing background conditions), and a contingency of 8 additional waste-filled bins. This represents a total of 608 drum- volume-equivalents (55-gal, 208 L) of actual OH TRU waste. A later phase of the test program is also defined but cannot be described in adequate detail at this time; all future test additions and contingencies would be included in this "Phase 3." The DOE has formed an Engineering Alternatives Task Force to evaluate potential waste form treatments, facility design modifications, and regulatory compliance approaches that may be evaluated during Phase 3 of the Test Phase. Phase 3 test bins would include any other alternate or processed waste forms, backfill materials, and/or getter materials that may be defined and developed in the future. These materials may be tested in Phase 1 or 2 if they are identified early enough. As indicated in Subsection 0.2, the volume of waste to ultimately be used in the Test Phase is subject to modification (a maximum 0-15 volume of 10 percent of the total waste destined for the WIPP, as analyzed in this SEIS). 0.3.1.1 Bin-Scale Test Objectives The objectives of the bin-scale tests are to 1) Quantify gas quantities, composition, generation, and depletion rates from TRU waste as a function of waste type, time, and interactions with brines and other repository natural and engineered barrier materials with a high degree of control; the experimental conditions would be primarily representative of the long-term, postoperational phase of the repository and the operational phase. 2) Provide a larger-scale evaluation and extension of the laboratory-scale test results, using actual OH TRU waste under repository relevant, expected conditions. 3) Evaluate the synergistic impacts of microbial action, potential saturation, waste compaction, degradation-product contamination, etc., on the gas- generation capacity and radiochemical environment of TRU waste. 4) Incorporate long-term room closure and waste compaction impacts on gas generation by including supercompacted waste. 5) Evaluate effectiveness for minimizing overall gas generation by incorporating getter materials, waste form modifications, and/or engineered alternatives into the OH TRU waste test system. 6) Measure solution leachate radiochemistry and hazardous constituent chemistry from saturated TRU waste interactions as a function of many credible environmental variables. 7) Determine the amount of volatile organic compounds/hazardous gases released from the TRU waste under realistic repository conditions in order to quantify releases of hazardous constituents and adequately address RCRA requirements. Reactive carbon composite filters will not be used because they could affect the behavior of these gases. 8) Provide necessary gas-generation and depletion data and source-term information in direct support of WIPP performance assessment analyses, predictive modeling, and related evaluation, and to justify pertinent assumptions used in modeling. 9) Help establish an acceptable level of confidence in the performance assessment calculations. Help evaluate pertinent modeling assumptions. Help eliminate most "what if" questions and concerns. 0-16 0.3.1.2 Bin-Scale Test Summary The bin-scale tests involve testing in multiple large, instrumented metal "bins" with specially prepared TRU waste and appropriate material additives. The "prepared" waste includes up to 6 drum-volume-equivalents of a specific type of actual CH TRU waste with added backfill materials (including salt), metal corrodants (mild steel wire mesh), and brine (to be injected at WIPP). Within each individual test bin there would be a specific type of TRU waste, either noncompacted or compacted. Any plastic bags encapsulating this waste would be "prebreached;" that is, the bags would be sliced or slashed or the waste itself would be shredded. Special preparation of the waste would occur at the generator/preparer facility. This "prebreaching" permits contact between, and interactions of, the waste with other added components within the bin, and within a time frame shorter than expected in the repository. Each WIPP test bin, after special waste preparation and filling, would be shipped to WIPP for emplacement and monitoring during the test period. These test bins are specifically designed to fit within a SWB (which is transported within a TRUPACT-2) for transportation to the WIPP and eventual post-test disposal. The test bin alone would not be used for transportation or as a terminal disposal container; the bin is for testing purposes only. Each bin would function as a nominally independent, isolated and controlled system. All of the test bins for Phases 1 and 2 would be isolated within one underground test room, Room 1 of Panel 1 (Figure 0.3.1). In Phase 3, bins may also be placed in Room 2 of Panel 1 The leak-tight bins would have a closely controlled and sealed test environment (internal atmosphere) similar to an isolated, waste-filled repository room. Each bin would be equipped with remote-reading thermocouples, pressure gages, pressure relief valves, gas flow/volume monitors, redundant gas sampling valves, and oxygen-specific detectors. Each test bin and associated instruments would be periodically and closely controlled and monitored by a computerized data acquisition system Each bin would also be equipped with integral, non-gas-sorbing high efficiency particulate air (HEPA) filters. As such, any gases sampled or released would not contain particulate radioactive contamination. The bin-scale test matrix includes combinations of the following parameters: 4 representative TRU waste materials classifications (waste types) 2 levels of waste compaction 4 types of backfill material 4 brine moistness conditions. The four waste types that have been selected for testing are High-organic/newly generated (HONG) (compacted and noncompacted) Low-organic/newly generated (LONG) (compacted and noncompacted) High-organic/old waste (HOOW) Inorganic processing sludges (PS). As noted in Subsection 0.1 , for purposes of bounding impacts it is assumed that CH TRU waste would be shipped from all 1 generator facilities. It is likely, however, that only waste from the Rocky Flats Plant and the Idaho National Engineering Laboratory would be used. 0-17 0Z9Z-S 'lj^jljc:_j E cr E E cc £ cc CO E cc M E cc £ QC 0) c Q. 081.2-S 0S6I.-S LU -I GQCO 009I.-S > o u CO 0) .2 > o o (0 »2 a> > o u w ■*^ U) v H ^ CC Q H Z ai O O (0 HI u -1 0) c CQ m -I < O CO o z (S CO UJ O UJ -i O 2 UJ CO UJ < Q. I-" 0) UJ H UJ < O (0 o r- I UJ UJ OC O I m u. o z o o o 0-18 The estimated contaminated number of bins per waste type is shown in Table 0.3.1. Other representative waste (i.e., high-activity, etc.) may be defined and tested during Phase 3. TABLE 0.3.1 Estimated number of bins PHASE 1 High-organic/newly generated (HONG) Low-organic/newly generated (LONG) Prepared sludges (PS) High-organic/old waste (HOOW) Empty/gas reference bins Total DRUM PHASE 2 VOLUMES 24 24 280 12 6 96 12 14 144 24 88 48 68 608 8 56 68 608 Most high-organic ("soft") and low-organic ("hard," primarily metal and glass) newly generated waste would be compacted at the Rocky Flats Plant. The advantage of using compacted waste In these tests is that the degradation behavior of compacted waste is expected to be very similar to regular (noncompacted) waste that has been crushed/compacted in situ by the long-term closure of the repository rooms. Thus, impacts on gas generation caused by compaction could be realistically evaluated during the course of these tests and factored into the performance assessment calculations. Other bin-scale test parameters are as follows: Moistness- Dry (expected short-term) Moistened with Salado Formation brine, about 1 percent by volume (expected case within several years) Saturated with Salado Formation brine, about 10 percent by volume (probable in the long-term) 0-19 Saturated with Castile Formation brine (possible in the case of human intrusion). Backfill (representative of the postoperational phase)- None Salt Salt (70 percent) and bentonite (30 percent) Salt, bentonite, and gas or radionuclide getter additives Salt and others (e.g., grouts or others to be defined later). The atmosphere inside selected test bins would be initially controlled and is expected to be representative of TRU waste in both the short-term post-emplacement period and later periods. HONG waste is expected to create its own anoxic (hydrogen and carbon dioxide) atmosphere primarily by radiolysis and would not require gas flushing. Similarly, no initial gas flushing for the inorganic PS waste would be conducted. The radiolytic depletion or production of oxygen from the PS waste would be quantified along with other evolved gases. The HOOW and LONG bins would be flushed with argon gas until an anoxic (no oxygen present) atmosphere is established. The study of potential anoxic corrosion of metals within the waste, as impacted by other ongoing degradation mechanisms, is one of the significant objectives of this test. All of the waste bins would be injected with (nonradioactive) tracer gases to help facilitate analysis and interpretation of the results. Gas sample collection would begin as soon as each bin is emplaced, prepared, and sealed. The samples would be analyzed with an on-site gas chromatography-mass spectrometer to determine major and minor gas concentrations and changes in gas compositions as a function of time. The major gases to be analyzed, based on earlier laboratory testing (Molecke, 1979), include hydrogen, carbon dioxide, carbon monoxide, methane, oxygen, water vapor, nitrogen, and injected tracer gases. The minor gases to be potentially measured include: volatile organic compounds (VOC), radon, ammonia, hydrogen sulfide, nitrogen oxides, hydrogen chloride, and any other detectable gases. Gas quantities and generation rates are significantly impacted by, and would be measured as a function of, • several representative classifications and types of OH TRU waste; • time (periodically, over several years); • impacts of several types and quantities of intruding brines; • impacts of waste interactions with salt, container metals, and backfill materials; • aerobic and anaerobic environment conditions representative of the operational-phase and longer-term, postoperational phase of the repository, respectively; and, • impacts of potential gas getter materials and engineered alternatives, particularly on gas consumption/production. Waste gas production also includes the synergistic effects of radiolysis, microbial degradation, and corrosion. Different test conditions are tailored so that the effects of O-20 individual environmental variables on gas production can be separated from the effects of other variables. The major gases are primarily generated or consumed by various waste degradation mechanisms occurring within the test bin or those remaining from the initial air atmosphere. The minor gases may arise in two ways: they may be sorbed on or in the waste before it is emplaced in the repository and eventually be volatilized in the repository, or they may be generated in the repository by waste degradation mechanisms. Determining whether VOCs and other hazardous gases are released from TRU waste is an important objective of the test program in order to adequately address compliance with RCRA regulations. Data and analyses would be incorporated into the performance assessment calculations, available on a near-continuous basis. 0.3.1.3 Bin-Scale Test Phases and Schedule This bin-scale test program is planned to take place in several phases. Phase 1 would incorporate test bins where all components can be presently defined. Approximately 48 waste-filled bins of different waste compositions and backfills, including replicates, would be included in Phase 1 . There would also be 8 other empty Phase I test bins used for gas baseline-reference purposes. Phase 2 tests would incorporate another 68 waste-containing bins, with more moisture conditions, with gas getter materials, and with the supercompacted high-organic and low-organic waste. Initiation of much of Phase 2 would be dependent on supporting laboratory data (Brush, 1989), particularly as to the composition of gas getters or other backfill material components and on the availability of supercompacted waste. Phase 2 tests would not be anticipated to start sooner than about early FY91 . Phase 3 of the test program, including all contingencies and additions, is under evaluation. Future needs for additional test bins and drum- volumes of actual CH TRU waste would be based on upcoming developments, preliminary test results, perceived data needs, and/or possible WIPP project decisions. Details of Phase 3 tests would be incorporated into a future, separate Test Plan addendum (Molecke, 1989a). Bin-scale testing would continue for a minimum of about 5 years, or until the data acquired are sufficient to provide confidence in the reliability of the information being obtained. At specific periods within the testing program, data would be analyzed and evaluated for input to ongoing performance assessment studies. At appropriate test intervals, data would be fully evaluated and documented in topical reports. 0.3.1.4 Bin Preparation and Transportation Safe transportation of the waste-filled test bins from the generator facility to the WIPP is a critical step in the testing program. The conceptual program design includes the following assumptions with regard to waste packaging and transportation. Two additions must be made to the preinstrumented bin before the waste is placed in the test bin. First, about a half-drum volume of backfill material would be placed in the bottom of the test bin. Second, about 6 drum-equivalents of bare, unpainted steel (mild steel wire mesh) would be placed along the bottom and side walls of the bin. 0-21 The bins would then be remotely filled with waste which would be characterized. Newly generated waste (HONG and LONG) could be loaded directly into the WIPP test bins at the generator facility. Previously packaged (drummed or boxed) waste (HOOW) could be emptied into the bins without the original waste packaging material. Sludges (PS) could be placed directly into the bins. After the waste is placed in the bins, another half-drum volume of backfill material would be sprinkled on top of the waste materials. The mated bin-lid/liner-lid combination would then be attached to the bin and sealed. The filled bin would be checked for surface contamination and, if necessary, decontaminated following standard procedures of the generator facility. The waste-filled test bins would be inserted into SWBs at the generator/preparer facility for transportation to the WIPP. The upper gas valves on the test bins (with HEPA filters) would be left in the open, gas-release position during transportation. Therefore, any gases vented would also be filtered through the redundant HEPA filter of the SWB. The SWBs would be loaded into the TRUPACT-II transportation containers and trucked to the WIPP. Removal of the waste bins from the SWBs would occur in the WIPP underground, just prior to emplacement. 0.3.2 ALCOVE The alcove tests are designed to provide data on production, depletion, and composition of gases resulting from the in situ degradation of OH TRU waste. These types of data are needed to support performance assessment of long-term repository behavior and to evaluate long-term generation and release of hazardous constituents. Data on TRU waste degradation rates are needed from testing that includes not only waste that is representative of anticipated waste to be disposed of at WIPP, but also representative of the time from emplacement to the long-term postoperational phase. These tests would enable acquisition of this data in a controlled research mode and allow multiple degradation mechanisms and impacts to be assessed. 0.3.2.1 Alcove Test Objectives The objectives of the alcove tests are to 1) Determine baseline gas quantities, composition, generation, and depletion rates for as-received, representative mixtures of TRU waste in a typical, operational phase repository room environment 2) Determine net gas quantities, composition, generation and depletion rates for a representative range of specially prepared mixtures of actual TRU waste (with and without compaction), backfill materials, gas getters, and intruding brine under representative, postoperational phase repository room conditions 3) Determine the amount of volatile organic compounds/hazardous gases released from the TRU waste under actual repository conditions 0-22 4) Provide an in situ test of gas getter effectiveness and demonstration of waste room backfilling procedures 5) Correlate large, alcove results of gas generation and interpretations with those of the laboratory and bin-scale tests of TRU waste degradation and gas production 6) Establish an acceptable level of confidence in the performance assessment calculations that include gas generation and depletion with actual in situ gas measurements and support validation of modeling assumptions. 0.3.2.2 Alcove Test Summary The primary purposes of this WIPP in situ alcove CH TRU waste test program are to provide relevant data and technical support to the WIPP performance assessment program for predictive modeling studies, and to provide in situ data for the assessment of hazardous component release and consequent impacts on the WIPP, in relation to EPA concerns and regulations. Specific data to be obtained include the quantities, compositions, and kinetic rate data on gas production and consumption resulting from various CH TRU waste degradation mechanisms. Similar data on potentially hazardous volatile organic compounds released by the waste would also be provided. This alcove test program involves, basically, the sampling and analysis of gases released from mixtures of CH TRU waste which have been emplaced within isolated, atmosphere-controlled test alcoves in the underground at the WIPP. The alcove tests would be conducted in six sealed atmosphere-controlled test alcoves. Four alcoves would be in Panel 1 and the remaining two alcoves would be in Panel 2 (Figure 0.3.2). Five of the test alcoves would be filled with waste. The sixth alcove would not have waste in order to collect "background" gases and establish baseline conditions. A test alcove would be about one-quarter the volume and one-third the length of a standard-size WIPP waste room. The test alcoves are smaller than standard rooms to increase the alcove stability with regard to short-term rock deformation and potential fracturing. (The behavior of the disturbed rock zone around full-sized rooms would continue to be examined in other experiments and by modeling during the Test Phase.) The waste used in the alcove tests would be "as received" (no special processing), compacted, and specially prepared CH TRU waste. All CH TRU test waste would be prepared and packaged at DOE waste generator facilities. "Specially prepared" waste is a waste container that has been filled with waste, backfill and metal corrodants in specified amounts. Waste types, representative of the majority of waste to be isolated at WIPP, include: High-organic/newly generated (HONG) Low-organic/newly generated (LONG) Inorganic processed sludges (PS) High-organic/old waste (HOOW). 0-23 r \0Z9Z-S I 1 I 1 I 1 I 1 I c to Ol (D in < < I- H- L E a: I I'l I I I I I L_ '081.3-S £ U) £ cc £ cc £ QC £ QC £ cc c nj a S ^ 096I.-S 0091-S < < I- < < I- CO at ui -J O s UL UJ < -I Q. CM UJ ^§ o o 3 I- r UJ u. o < o o 0-24 The approximate quantity of drums per waste type to be used in the alcove tests is based on a preliminary analysis (Batchelder, 1989) of waste currently stored at DOE waste generator facilities and extrapolated to exist through the year 2013. The required in situ alcove CH TRU waste gas data would be acquired in two phases. The alcoves in the Test Phase and the test parameters of each alcove are as follows: PHASE 1 Test Alcove 1 No waste Oxic atmosphere Dry No backfill Test Alcove 2 As-received, mixed CH TRU waste Oxic atmosphere Dry No backfill PHASE 2 Test Alcove 3 Specially prepared and noncompacted waste Anoxic atmosphere Moist, 1% brine No backfill Test Alcove 5 Specially prepared and noncompacted waste Anoxic atmosphere Moist, 1% brine Backfill: salt, bentonite, gas getter material Test Alcove 4 Specially prepared, compacted waste Anoxic atmosphere Moist, 1% brine No backfill Test Alcove 6 Specially prepared, compacted waste Anoxic atmosphere Moist, 1% brine Backfill: salt, bentonite, gas getter material The alcove tests would be conducted in two phases. Phase 1 includes test alcoves 1 and 2. Test alcove 2 (TA2) would represent expected conditions in the short-term, operational phase of the repository. Test alcove 1 is the gas baseline room. It would provide gas composition data (i.e., trapped atmosphere and gases released from the host rock) necessary for comparison with waste-filled rooms. Test alcove 2 would contain a representative mixture of about 1 ,050 drum or drum- volume equivalents of "as-received" CH TRU waste. This waste would be packaged at waste generator facilities into either standard 55-gallon drums or SWBs. Both types of containers would be vented and particulate-filtered. Alcove TA2 would be used to provide data on CH TRU waste gas generation under actual, in situ repository conditions (initial air atmosphere, dry/as-received, with no salt, backfill, or getter material in direct contact with the waste), and is specifically representative of the short-term, operational-phase of the repository. TA2 also provides the initial data for repository time t = 0, necessary for the Phase 2 tests. 0-25 Phase 2 of this alcove test program would include four alcoves, and is specifically tailored to be representative of the long-term, postoperational phase of the WIPP repository. Phase 2 test "tailoring" consists of three basic operations: alcove gas atmosphere control, waste special preparation, and brine injection of all waste. It is assumed in WIPP performance assessment that the repository will be anaerobic in the long-term, i.e., anoxic, less than 10 ppm O2. Therefore, the atmosphere in each alcove would be initially prepared and kept anaerobic. This involves nitrogen gas flushing of each alcove and the continuous use of an oxygen-gettering reactant system. The TRU waste in each Phase 2 test container would be "specially prepared" and/or packaged, as follows. There will be a specific type of TRU waste, either noncompacted or supercompacted, within each test drum or SWB. Any plastic bags encapsulating this waste would be "prebreached," e.g., sliced, slashed, or similarly prepared at the waste generator/storage facility. This operation is beneficial for both testing and transportation (within TRUPACT-II) purposes. The waste would be sandwiched between added layers of backfill materials, 70 wt% WIPP crushed salt/ 30 wt% bentonite clay, and metal corrodant materials (mild steel wire mesh). One or two unbreached plastic bags would enclose all the prebreached waste and other components within one total environment. These all-encompassing plastic bags, at the periphery of the waste container, are used for contamination control during waste packaging operations at the generator facilities. After emplacement in the WIPP, all Phase 2 TRU waste containers would be specifically moistened with about 1% by volume of Salado brine; this is to be representative of probable long-term brine intrusion. The brine is a mixture of 90% by volume of artificially prepared, and 10% of WIPP-collected Salado brine. Small amounts of brine, 2 liters/drum or 14 liters/SWB, would be injected through brine-injection septa on the top of each container into or onto the waste inside, breaching the all-encompassing plastic bags. Phase 2 test alcoves TA3 and TA5 would include "specially prepared," noncompacted waste, and TA4 and TA6 would include "specially prepared," supercompacted waste. Alcoves TA5 and TA6 would also include both backfill and gas getters, e.g., reactant, sorptive materials that encapsulate the waste. Backfill and getter materials would be emplaced over and around the waste container stacks in these two test alcoves in a fully retrievable mode. All test waste would be emplaced in such a manner to ensure that post-test retrieval is possible. Waste backfilling would be conducted for gas mitigation test purposes, as well as for operational demonstrations. If other engineering modifications to minimize TRU waste gas generation are available in the appropriate time frame, they could also be added to alcoves TA5 and TA6 for testing of their in situ efficacy. Four of the six test alcoves would be located along the northern edge of Panel 1 ; the remaining two alcoves would be located within Panel 2 (Figure 0.3.2). Two of the conventionally-mined alcoves (1 and 2) would be 13 ft high by 25 ft wide by 100 ft long. Four of the test alcoves (3, 4, 5, and 6) would be 0.8 ft higher, for a total of 13.8 ft to accommodate compacted backfill on the floor. The available volume to store 0-26 the TRU waste in each test alcove is about 32,500 ft^. The alcove would be rock bolted and wire meshed to facilitate waste retrievability and to increase operational safety. The access drifts would have a slightly smaller cross-sectional area, approximately 1 3 ft (4 m) wide by 14 ft wide, to facilitate sealing. The access drift would be 170 ft long. The height and width of the access drift are the minimum size possible to accommodate a mining machine and still allow sealing with an appropriately shaped closure seal. The closure seal would be inflatable and adequate to control pressure of up to 1 .5 pounds per square inch (psi) differential pressure without being pushed out. An internal differential pressure of 0.5 psi must be maintained within the test alcove. The test alcove seal would be constructed of materials that have a five-year durability when in contact with salt, gases and liquids expected within the test alcove and that are impermeable to air/oxygen (without generating volatile gases). The seals would contain instrumentation and access ports for the gas sampling system. Dual redundant closure seals would be placed in each access drift, in case one seal leaks while in place. The test alcoves would contain either 150 seven-packs of drums or standard waste boxes, stacked four across and three high. Test alcoves 3 and 5 would contain a mixture of specially prepared and packaged waste that has not been compacted. Test alcoves with standard, noncompacted waste would contain about 1 ,050 drums or drum- volume equivalents (210 liter or 55-gallon). Test alcoves 4 and 6 would contain similar waste that has been compacted. Test alcoves with compacted waste would contain about 350 drums of waste. Waste quantities were selected based on statistical evaluations and practical matters. Each test alcove would be equipped with remote reading thermocouples, pressure gages, and HEPA-filtered gas relief and gas volume monitoring gages. AH instruments would be connected to a computerized data acquisition system. No appreciably elevated gas pressures would be present in the test alcoves. A gas recirculation system would be installed to mix gases for sampling; it would include inlet and outlet ducts that penetrate through the inflatable seal with gas sampling ports or septa. All instrumentation and hardware access would be through a sealed access port in the test alcove seal. After the waste, backfill, instruments, hardware and seals are installed, there would be no access to the test alcoves during the tests. Tracer gases would be added to the test alcoves. Tracer gases would help monitor outflow from the test alcoves to the repository environment, and evaluation of the changes in concentration over time of these tracers would allow compensating corrections to be applied to all other gases being quantified. Separate tracers would be used in each test alcove to monitor any potential leakage from one alcove to another through fractures in the rock. 0-27 S<*^ Gas quantities, compositions, and generation rates can be significantly affected by, and would be measured as a function of, several factors: representative classifications and types of CH TRU waste, and mixtures thereof time (periodically over several years) impacts of intruding, moistening brine impacts of waste interactions with salt, container metals, and backfill materials aerobic and anaerobic environment conditions, as representative of the operational-phase and longer-term, postoperational-phase of the repository, respectively • impacts on gas consumption of potential gas getter materials that surround or encapsulate the waste containers. The waste gas production results also include synergisms between the various waste materials and degradation modes. Gases periodically collected from each test alcove would be analyzed using a gas chromatograph/mass spectrometer to determine major and minor gas concentrations, and changes in those concentrations as a function of time. This allows rates of generation and/or depletion to be determined. Evaluation of the changes in gas composition would help to determine the relative importance and kinetics of individual degradation mechanisms over time and of the subsequent impacts of degradation by- products on further gas production. The major and minor gases to be analyzed in the alcove tests are the same as those to be analyzed in bin-scale tests (see Subsection 0.3.1.2). Gas data collection would begin as soon as each test alcove is filled with TRU waste, sealed, and the initial alcove gas atmosphere appropriately prepared. These tests are expected to start providing significant data within months after test emplacement. However, due to the expected slow rate of gas generation and the lack of sensitivity due to the large, masking amount of gas atmosphere initially in the alcoves, it is expected that almost one year will be required before there is an adequate quantity and quality of data for interpretations. WIPP alcove testing would continue for roughly 5 years, or until the data acquired are sufficient to provide confidence in the reliability of the information being obtained. Data would be analyzed and evaluated for input to ongoing performance assessment studies on a near-continuous basis. Data would be fully evaluated and documented in periodic, topical reports. 0.3.2.3 Alcove Test Phases and Schedule Initiation of Phase 2 testing in alcoves TA5 and TA6 depends on supporting laboratory data, (Brush 1 989) particularly as to the composition and quantities of gas getters, other 0-28 backfill material components, or proposed engineered alternatives, tests would not be expected to start sooner than FY91 . These Phase 2 The first four test alcoves, TA1 - TA4, must be mined, equipped, and instrumented prior to the first receipt of waste at the WIPP, expected in FY90. This would be followed by sequential waste loading and filling for each alcove, alcove sealing, appropriate atmosphere preparation, and subsequent gas testing. In order to adequately meet WIPP performance assessment schedule needs, the first four alcoves must be on-line and generating data for about one year prior to the end of FY92 (DOE, 1989a). The remaining two needed test alcoves, TA5 and TA6, would be available for testing at a somewhat later date. Detailed test planning for these in situ alcove CH TRU waste tests continued through 1989. Procurement activities for necessary test equipment, instruments, associated supplies, and the actual CH TRU waste will proceed through and beyond 1990. Site preparation, including any necessary mining and test installation, also began during FY90 and will continue for one year or more. Initial data acquisition from these tests, e.g., baseline-alcove gas analyses and interpretations, is anticipated to start during FY91. Further descriptions and technical details of these WIPP in situ alcove CH TRU waste tests will be found in the Test Plan (Molecke, 1989b). 0.3.2.4 Waste Preparation and Transportation Safe transportation of the waste-filled test drums and/or standard waste boxes (SWB) from the generator/preparer facility to the WIPP is a critical step in the testing program. The conceptual program design includes the following assumptions with regard to waste packaging and transportation. The specially prepared waste is placed in a polyethylene-lined drum or SWB. About 0.5 cubic ft of backfill materia! would be placed in the bottom of the container. A special metal corrodant (a mild steel wire screen or mesh) would be inserted in the container on top of the backfill layer. The container would then be nearly filled with CH TRU waste in prebreached plastic bags. An additional 0.5 cubic ft of backfill would be placed over the waste. The waste-filled containers would be inserted into the TRUPACT-II at the generator/preparer facility for transportation to the WIPP. Gases released from the drums during transportation would be contained in the TRUPACT-II containers. 0-29 0.4 UNDERGROUND TEST OPERATIONAL SAFETY Concerns regarding operational test safety are addressed in three categories: emplacement, test monitoring, and mine safety. The major safety consideration in the first two categories, emplacement and test monitoring, is personnel exposure to radioactive and/or hazardous constituents. The safety practices during emplacement operations would be similar to those planned for normal operations. During the test monitoring and sampling activities, concerns are focused on personnel exposure during sampling and ventilation due to release of gases from the test bins or rooms. The third category, mine safety considerations, is focused on room stability and waste retrieval. 0.4.1 EMPLACEMENT SAFETY CONCERNS The emplacement operations for testing are anticipated to be similar to planned WIPP waste handling operations. WIPP waste handling operations would encompass a broad range of activities. The operating functions at the WIPP involve the handling of waste for emplacement, operation of surface facilities, and mining operations. Waste handling consists of shipping container receipt and unloading, waste handling from the surface to the underground facility, emplacement in the underground test area, and maintenance of required records. In support of waste handling activities, the surface and underground facilities would be operated in a manner to ensure operator and public safety in accordance with the "WIPP Operational Safety Requirements Administration Plan" and the "WIPP Radiation Safety Manual" (WEC, 1988a and 1988b). Unlike plans for normal operations, the emplacement operations, and subsequent sampling and retrieval, would require operators to be in the downstream ventilation air flow on a routine basis. This air flow would be monitored for personnel safety. Use of waste container handling equipment during the Test Phase would be limited to emplacement and retrieval activities. Thus, the potential for an equipment handling accident would be restricted. The operational safety requirements are based on the as low as reasonably achievable (AU\RA) principle. The ALARA techniques applied to the WIPP facilities are based on DOE Order 5480.11, as well as DOE's exposure guide (DOE, 1980a), as appropriate for this first-of-a-kind facility. Radiation exposure to plant personnel is kept ALARA by continued review of operations, training, and the functioning of the Radiation Safety and Emergency Programs Section. The WIPP ALARA program is described in Section 2.0 of the WIPP Radiation Safety Manual (WEC, 1988b). The expected radiation and chemical doses to plant personnel described in Subsections 5.2.3 and 5.2.4 of this SEIS, respectively, are based on testing with 1 percent of the total projected waste and are far below regulatory guidelines. On this basis, the dose estimates in this SEIS can be considered a conservative upper bound. O-30 0.4.2 TEST OPERATIONAL SAFETY CONCERNS Safety concerns during the testing are related to radiological safety, hazardous material safety, and ventilation. In accordance with DOE Order 5480.5 (DOE, 1986), Operational Safety Requirements would be developed as necessary to ensure control of appropriate safety parameters during the Test Phase. Operating procedures would be developed by Westinghouse Electric Corporation, the WIPP operating contractor, in coordination with Sandia National Laboratories, the in situ test coordinator, to guide the testing and monitoring activities. These procedures would be approved by the Westinghouse Radiation Safety and Emergency Programs Section. Radiological and hazardous material safety operations associated with the in situ testing of actual CH TRU waste would be guided by procedures, which would include specific monitoring and testing requirements. The program would, at the minimum, include the following requirements: • Gas or other samples taken in the testing program will be monitored for radiation and volatile organic compounds prior to being removed from the test area, a defined Radioactive Materials Area. • Appropriate personal protective equipment will be worn during sampling and monitoring activities. • Radiation Work Permits will be prepared for most of the test activities conducted with the actual waste. • Site Health Physics and Industrial Hygiene personnel will monitor sampling and other test-related activities. • Westinghouse Radiation Safety and Emergency Programs Section personnel will review sampling and monitoring procedures. The ventilation system for the WIPP underground facilities is designed to provide a suitable environment for personnel and equipment. It is also designed to remove potential airborne radioactive or hazardous material from the underground area during routine operations or through HEPA filters in the event of an accident. The ventilation system is an exhausting system in which the underground area is maintained below atmospheric pressure. The design airflow quantities are based on standard local. State, and Federal industrial and mining laws and practices. Air quantities supplied to the underground area have been determined to meet or exceed the criteria specified in the Mine Safety and Health Administration code. All gases released through the pressure relief valves on bins and alcove seals would already have been filtered through a non-gas-sorbing HEPA filter. Therefore, the potential for a radioactive release from within the bins or drums is very small. Released gases are expected to be predominantly nitrogen, with low concentrations of carbon dioxide, carbon monoxide, hydrogen, oxygen, tracers, and possibly methane and other volatile organics. These released gases would be vented to the person-access area or 0-31 directly to a mine ventilation duct to be carried away by normal mine ventilation. Separate chromatograph/mass spectrometry analyses of gases from the test bins and alcoves would provide a measure of the possible hazard of such gas released in small quantities. If necessary, samples of mine air in the immediate vicinity of the test room person-access areas may also be analyzed for safety assurances. 0.4.3 MINE SAFETY CONCERNS Guaranteeing the retrievability of CH TRU waste emplaced and related operational mine safety are major concerns in the design of the underground testing program. The test areas must remain stable and open during the Test Phase and for several more years to assure retrievability. Concerns about rock spalling, fracturing, and slabbing would be mitigated by rock bolts and wire mesh. In order to minimize the rock instability uncertainties, the roofs of the test alcoves and rooms would be supported using patterned rock bolting, which has been successfully used for stability in other portions of the underground. The rock bolt system, which was designed and installed in Panel 1 , consists of three-fourths inch diameter by ten- foot long mechanically anchored bolts. A similar rock bolting pattern would be implemented in the alcoves. Wire mesh would also be added. The support system has been designed to support the full weight of the immediate roof beam up to the first anhydrite layer in the roof. The pattern is staggered in order to increase bolt hole distance, and, therefore, reduce the potential for fracturing between holes. It is not expected that the bolting will prevent creep of the salt nor stop the fracturing and separating that have been observed in the underground. Rather, the bolting would prevent roof rock from falling, once it has fractured and has become detached. In order to maintain the gas and brine leak-tight integrity of the test room roofs, certain precautions must be taken with regard to rock bolt installation, testing and sealing procedures. Appropriate types of caulking sealant would be injected into the rock bolt holes; degassing and volatilization of the sealant material would be kept to a minimum to limit interference with subsequent gas sampling and analyses. 0-32 0.5 POST-TEST OPERATIONAL SAFETY Post-test operational safety concerns focus on three main issues: retrieval of bins, retrieval of drums, and options for disposition of the waste used in the tests. Safety concerns associated with bin retrieval include handling and processing the waste and possible exposure to radiation and hazardous materials. Radiological exposures to the workers and to the public from retrieval operations are discussed in Subsection 5.2.3 of this document. While potential drum handling accident scenarios are not different than during emplacement, the probability of container failure during handling may be higher, particularly for drums from the test alcoves because of the potential for drum corrosion or damage during the test period. In addition, retrieval of waste from back- filled rooms may be more complex resulting in a higher probability of an accident during retrieval operations. However, as discussed in Subsection 5.2.3, special procedures and provisions would be employed to reduce worker exposures in the event that retrieval of the waste is required. Disposition of the waste after the tests is subject to regulatory requirements and available disposal or storage facilities. A Waste Retrieval Plan (DOE, 1989d) is currently being developed to describe the processes, administrative controls and procedures, and organizational responsibilities that would be implemented to ensure safe and effective removal of emplaced TRU waste. 0.5.1 BIN RETRIEVAL At the end of the test period, the bins would still be filled with various combinations of CH TRU waste, backfill, and brine. Gases, potentially with radioactive or hazardous constituents, are also expected to be in the bins. The gases would be purged by flushing through the HEPA filters on the bins. The HEPA filters would remove any radioactive particulates. The gases would be vented through the facility ventilation system. Any free liquids would be removed from the bins. The waste in the bins could be further desiccated by flushing the bins with warm air or injecting sorptive materials. Disposition of the liquid and the waste is discussed in Subsection 0.5.2. Safety precautions during the post-test period would be similar to those taken during the test period (Subsection 0.4.2). Gas and liquids removed from the test bins would be monitored for radiation and volatile organic compounds prior to being removed from the test area. During all post-test activities, appropriate personal protective equipment would be worn. Site health physicists and industrial hygienists would monitor post- test-related activities. Radiation work permits would be prepared for the post-test activities conducted with the actual waste. The Radiation Safety and Emergency Programs Section personnel would review sampling and monitoring procedures in use during post-test activities. 0-33 0.5.2 ALCOVE RETRIEVAL At the conclusion of the alcove test measurements, five of the alcoves would contain various combinations of waste, backfill, drums, and gas. The injected brine is expected to be predominantly sorbed on the waste matrix materials; very little free liquid is anticipated. If a decision to retrieve waste is made at the end of the Test Phase, a contamination control area would be established in the waste retrieval chambers during waste retrieval operations. Air flow in the control area would be maintained such that workers remain in the upstream flow of the working face of the waste stack. Current plans are to continuously filter area exhausts through a single HEPA filter, reducing the concentration of particulates released to the underground exhaust shaft by a factor of 1 ,000 before release to the atmosphere. The gas atmosphere in each alcove would be purged (flushed, or simply released) into the normal mine ventilation system. The plug seals would then be removed. In the test alcoves where backfill was installed, the backfill would be removed, possibly by vacuuming as waste retrieval proceeds. Safety precautions during the post-test period would be similar to those taken during the test period (Subsection 0.4.2). Gas removed from the test bins and alcoves would be monitored for radiation and volatile organic compounds prior to being removed from the test area. During all post-test activities, appropriate personal protective equipment would be worn. Site health physics personnel and industrial hygienists would monitor post-test-related activities. Radiation Work Permits would be prepared for the post-test activities conducted with the actual waste. The Radiation Safety and Emergency Programs Section personnel would review sampling and monitoring procedures in use during post-test activities. 0-34 REFERENCES FOR APPENDIX O Batchelder, H. M., 1989. "Meetings Held at Rocky Flats and INEL," memorandum, June 12, 1989, Westlnghouse Waste Isolation Division, Carlsbad, New Mexico. Brush, L H., 1989. Test Plan for Laboratory and Modeling Studies of Repository and Radionuclide Chemistry , draft, Sandia National Laboratories, Albuquerque, New Mexico. Brush, L H., and D.R. Anderson, 1988. "Potential Effects of Chemical Reactions on WIPP Gas and Water Budgets," Sandia National Laboratories Memorandum, Sandia National Laboratories, Albuquerque, New Mexico. Caldwell et al. (D. E. Caldwell, M. A. Molecke, R. C. Hallett, E. Martinez, and B. J. Barnhart), 1987. "Rates of CO2 Production from the Microbial Degradation of Transuranic Wastes Under Simulated Geologic Isolation Conditions," SAND87- 7170, Sandia National Laboratories, Albuquerque, New Mexico. Clements, T. L Jr., and D. E. Kudera, 1985. TRU Waste Sampling Program: Volume I & II . EGG-WM-6503, Idaho National Engineering Laboratory, Idaho Falls, Idaho. DOE (U.S. Department of Energy), 1989a. Draft Final Plan for the Waste Isolation Pilot Plant Test Phase: Performance Assessment . DOE/WIPP 89-011, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1989b. Final Safety Analysis Report. Waste Isolation Pilot Plant. Carlsbad. New Mexico , draft, DOE/WIPP 88-xxx, Albuquerque, New Mexico. DOE (U.S. Department of Energy), 1989c. Radionuclide Source Term for the WIPP . DOE/WIPP-88-005, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1989d. Waste Retrieval Plan , draft, DOEA/VIPP 89- 022, Carlsbad, New Mexico. DOE (U.S. Department of Energy), 1986. "Safety of Nuclear Facilities," DOE Order 5480.5, Washington, D.C. DOE (U.S. Department of Energy), 1980a. A Guide to Reducing Radiation Exposure to As Low As Reasonably Achievable (ALARA) . DOE/EV/1 830-T5. DOE (U.S. Department of Energy), 1980b. Final Environmental Impact Statement: Waste Isolation Pilot Plant . DOE/EIS-0026, Washington, D.C. 0-35 ..<3^«y: Kosiewicz, S. T., 1981. "Gas Generation from Organic Transuranic Wastes, I. Alpha Radiolysis at Atmospheric Pressures," Nuclear Technology, Vol. 54, pp. 92-99. Kosiewicz, S. T., 1980. "Cellulose Thermally Decomposes at 70°C," Thermochimica Acta, Vol. 40, pp. 319-322. Kosiewicz et al. (S. T. Kosiewicz, A. Zenwekh, and B. Barraclough), 1979. "Studies of Transuranic Waste Storage Under Conditions Expected in the Waste Isolation Pilot Plant (WIPP)," Interim Summary Report, October 1, 1977 - June 15, 1979, LA-7931 -PR, Los Alamos National Laboratory, Los Alamos, New Mexico. Lappin et al. (A. R. Lappin, R. L Hunter, D. P. Garber, and P. B. Davies, eds.), 1989. Systems Analysis. Long-Term Radionuclide Transport, and Dose Assessments, Waste Isolation Pilot Plant (WIPP) . Southeastern New Mexico: March 1989, SAND89-0462, Sandia National Laboratories, Albuquerque, New Mexico. MSHA (Mine Safety and Health Administration), Federal Mining Code for Metal and Nonmetallic Underground Mines . Title 30, Code of Federal Regulations, Part 57. Molecke, M. A., 1989a. "WIPP Bin-Scale CH TRU Waste Tests," review draft, August 1989, Sandia National Laboratories, Albuquerque, New Mexico. Molecke, M. A., 1989b. 'Test Plan: WIPP In Situ Alcove CH TRU Waste Tests," management review draft, December 1989, Sandia National Laboratories, Albuquerque, New Mexico. Molecke, M. A., 1979. Gas Generation from Transuranic Waste Degradation: Data Summary and Interpretation . SAND79-1245, Sandia National Laboratories, Albuquerque, New Mexico. Sandia National Laboratories, 1 979. Summary of Research and Development Activities in Support of Waste Acceptance Criteria for WIPP . SAND79-1305, Sandia National Laboratories, Albuquerque, New Mexico. Tyler, et al. (L D. Tyler, R. V. Matalucci, M. A. Molecke, D. E. Munson, E. J. Nowack, and J. C. Stormont), 1988. Summary Report for the WIPP Technology Development Program for Isolation of Radioactive Waste . SAND88-0844, Sandia National Laboratories, Albuquerque, New Mexico. WEC (Westinghouse Electric Corporation), 1988a. WIPP Oper ational Safety Reguirements Administration Plan . WP-04-7, Rev. 4, Waste Isolation Pilot Plant, Carlsbad, New Mexico. WEC (Westinghouse Electric Corporation), 1988b. WIPP Radiation Safety Manual . WP- 12-5, Waste Isolation Pilot Plant, Carlsbad, New Mexico. Zenwekh, A., 1979. "Gas Generation for Radiolytic Attack of TRU-Contaminated Hydrogenous Waste," LA-7674-MS, Los Alamos National Laboratory, Los Alamos, New Mexico. 0-36 APPENDIX P TRU WASTE RETRIEVAL, HANDLING, AND PROCESSING P-i ■^>>: TABLE OF CONTENTS Section Page P.1 INTRODUCTION P-1 P.2 SAVANNAH RIVER SITE P-3 P.2.1 Retrieval and Processing P-3 P.2.2 Consequences P-5 P.2.2.1 Routine Operations P-5 P.2.2.2 Facility Accidents-Retrieval P-6 P.2.2.3 Facility Accidents-TRU Waste Processing Facility P-8 P.3 HANFORD RESERVATION P-1 1 P.3.1 Waste Characteristics and Current Management Methods P-11 P.3.2 Retrieval P-11 P.3.3 Waste Receiving and Processing Facility P-1 5 P.3.3.1 Waste Process Description P-1 5 P.3.4 Consequences of Waste Receiving and Processing Operations . P-1 8 P.3.4.1 Radiological Emissions P-1 8 P.3.4.2 Radiological Impacts P-1 8 P.4 LOS ALAMOS NATIONAL LABORATORY P-20 P.4.1 Retrieval and Processing P-20 P.4.2 Waste Storage Site P-21 P.4.3 TRU Waste Size Reduction Facility P-21 P.4.4 TRU Contaminated Solid Waste Treatment and Development Facility (TDF) P-23 P.4.5 TRU Waste Preparation Facility P-23 P.4.6 TRU Waste Nondestructive Examination and Analysis (NDE-NDA) Facility P-24 P.4.7 TRU Waste Transportation Facility P-25 P.4.8 TRU Corrugated Metal Pipe (CMP) Saw-Processing Facilities . . . P-25 P.5 OAK RIDGE NATIONAL LABORATORY P-27 P.6 IDAHO NATIONAL ENGINEERING LABORATORY P-29 P.6.1 Waste Retrieval and Processing P-29 P.6.1.1 Waste Characteristics and Current Management Methods P-29 P.6.1 .2 Environmental Effects of Current Operations P-30 P-iii Section Page P.6.1 .3 Methods for Retrieving and Handling Waste P-31 P.6.1.4 Retrieval Building and Operations P-35 P.6.1. 5 Processing to Meet WIPP WAG P-35 P.6.2 Process Experimental Pilot Plant P-36 P.6.2.1 Existing Facilities and Process . P-36 P.6.2.2 Waste Characteristics P-39 P.7 ROCKY FLATS PLANT P-40 P.7.1 Processing P-40 P.7.2 Supercompaction and Repackaging Facility Equipment Description P-42 P.7.2. 1 Hard-Waste Entry into the Supercompaction and Repackaging Facility P-43 P.7.2.2 Soft-Waste Entry and Precompaction P-43 P.7.2.3 Supercompaction P-44 P.7.3 TRU Waste Shredder Description P-45 P.7.3.1 TRU Waste Shredder Equipment Description P-45 P.7.3.2 TRU Waste Shredder Process Description P-45 P.8 BIN-SCALE TESTS P-47 REFERENCES FOR APPENDIX P P-49 P-iv LIST OF FIGURES Figure Page P.2.1 Savannah River Site TRU waste management plan P-4 P.3.1 TRU waste asphalt pad storage P-12 P.3.2 Typical caisson for TRU waste storage P-1 3 P.3.3 Typical TRU waste burial trench P-1 4 P.3.4 Waste Receiving and Processing Facility flow diagram P-1 6 P.4.1 Los Alamos National Laboratory TRU waste process flow P-22 P.5.1 Simplified diagram of Oak Ridge National Laboratory contact-handled transuranic waste management activities P-28 P.7.1 Supercompaction and Repackaging Facility process flow diagram .... P-41 P-v '^'■^^ LIST OF TABLES Table Page P.2.1 P.2.2 P.3.1 P.3.2 P.6.1 P.6.2 P.6.3 Summary of consequences from postulated accidents in the burial ground P-9 Summary of consequences from postulated accidents in the TRU Waste Processing Facility P-10 Population total-body dose commitments (man-rem) from the processing of retrievably stored and newly generated CH TRU waste at the Waste Receiving and Processing Facility P-19 Maximum individual total-body dose commitments (rem) from the processing of retrievably stored and newly generated CH TRU waste at the Waste Receiving and Processing Facility P-19 Summary of radiological consequences to the public from accidental or abnormal releases during RWMC/SWEPP operations with stored TRU waste P-32 Summary of radiological consequences to the maximally exposed worker from accidental or abnormal events during RWMC/SWEPP operations with stored TRU waste P-33 Excess cancer risks due to accidents associated with RWMC/SWEPP operations with TRU stored waste P-34 P-vi P.1 INTRODUCTION This appendix has been prepared in response to comments requesting that this SEIS evaluate TRU waste retrieval, certification, handling, and processing activities that would be conducted at the various generator/storage facilities for the purpose of preparing the waste for transport to the WIPP. In the 1980 FEIS, Subsection 9.8, and in this SEIS, Subsection 5.2.1 , waste retrieval and processing at the Idaho National Engineering Laboratory are discussed. These discussions include: 1) waste characteristics and management methods; 2) the consequences of current operations from routine handling and potential accidents; and 3) the methods used to retrieve, process, and ship waste. This appendix provides information that describes the current and planned TRU waste retrieval and processing activities at representative DOE generator/storage facilities. Many of these activities would support TRU waste certification and preparation for transport to the WIPP. However, these retrieval and processing activities would be applicable even if the No Action Alternative were implemented. For example, waste containers currently in retrievable storage on asphalt pads and covered with plastic and soil will ultimately have to be retrieved and altered (treated or repackaged) to avoid a release of materials from package degradation. Once this becomes necessary, it would be appropriate to assay the packages to better characterize the contents. Other treatments could be applied at this time as appropriate. Therefore, the processes described herein are not unique to WIPP operations. Appropriate NEPA documentation has been or will be prepared for any proposed modifications to TRU waste management activities of the various DOE facilities. This appendix also provides a description of bin and waste preparation that would occur at the generator/storage facilities prior to the Test Phase. This appendix draws upon the following documentation: • Idaho National Engineering Laboratory . A draft Environmen tal Assessment for the Process Experimental Pilot Plant (PREPP) has been prepared and Is undergoing internal review. Other NEPA documentation will be prepared for other retrieval and process facilities as proposed. • Hanford Reservation . A Final Environmental Impact Statement (DOE/EIS- 01 13), "Disposal of Hanford Defense High-Level, TRU and Tank Waste" (DOE, 1987a), was published in December 1987 and a Record of Decision was issued on April 4, 1988 (53 FR 12449). ^ Copies of preliminary drafts of documents in internal review are not yet publicly available; descriptive information and environmental consequences are preliminary and subject to change. P-1 • Los Alamos National Laboratory . A draft Environmental Assessment addressing waste retrieval, processing, and shipment to the WIPP has been prepared and is undergoing internal review.^ • Oak Ridge National Laboratory . A draft Environmental Assessment addressing CH TRU waste has been prepared and is undergoing internal review.^ A similar Environmental Assessment addressing RH waste will be prepared in 1992. • Savannah River Site . DOE/EA-0315, "Environmental Assessment on Management Activities for Newly Generated TRU Waste, Savannah River Plant" (DOE, 1988a) and a finding of no significant impact covers retrieval, treatment, and packaging for shipment to the WIPP. • Rocky Flats Plant . DOE/EIS-0064, "Final Environmental Impact Statement: Rocky Flats Plant Site" was published in April, 1980. Also, an Environmental Assessment to consider the potential environmental impacts that may occur from construction and operation of a Supercompactor and Repackaging Facility and a Transuranic Waste Shredder has been prepared and is undergoing internal review.^ • WIPP Site . WIPP 89-01 1 , "Draft Plan for the Waste Isolation Pilot Plant Test Phase, Performance Assessment and Operations Demonstration" has been prepared (DOE, 1989). The DOE believes that the waste retrieval and processing activities described herein are representative of those that likely would occur at other DOE facilities that may eventually transport post-1970 TRU waste to the WIPP. This belief is based on the following: • The similarity in retrieval and processing approaches at the various facilities and the nature of retrievable storage among facilities. • The volume of retrievably stored CH TRU waste at the six DOE facilities described constitutes 98 percent of the total retrievably stored inventory (see Table 3.1). • The magnitude of the consequences presented for the Idaho National Engineering Laboratory, Hanford Reservation, and Savannah River Site. As noted elsewhere in this SEIS, the DOE will issue another SEIS at the conclusion of the Test Phase; such a SEIS would update the information contained in this Appendix for all 10 DOE facilities and would analyze in detail the system-wide impacts (including those from retrieval, handling, processing, and transportation) of disposal of post-1970 TRU waste in the WIPP. Copies of preliminary drafts of documents in internal review are not provided. P-2 ^m^M P.2 SAVANNAH RIVER SITE P.2.1 RETRIEVAL AND PROCESSING TRU waste at the Savannah River Site is In retrievable storage on concrete pads or buried in shallow trenches. It is contained in concrete and steel boxes, concrete culverts, and galvanized steel drums covered with 4 ft of soil or tornado netting (in use since 1985). The 4-ft soil cover would be removed from the stored waste pads by earth-moving equipment to within 6 to 12 inches of the waste containers. The remaining soil would be removed with the remotely operated, HEPA-filtered soil vacuum. Drums would be removed using a shielded lifting canister. Large steel boxes and concrete culverts would be lifted from the pads and placed directly on a transport trailer for shipment to the TRU Waste Processing Facility building. Retrieved TRU waste and the newly generated TRU waste requiring processing prior to certification would be processed at a new TRU Waste Processing Facility. A flow diagram for TRU waste processing at the Savannah River Site is depicted in Figure P.2.1. The TRU Waste Processing Facility is scheduled to begin operation in 1995. Waste containers would be received at the TRU Waste Processing Facility through an airlock into a high bay storage and opening area. The TRU Waste Processing Facility would be used to vent, purge, x-ray, and assay the storage containers; size-reduce the large waste not suitable for shipment; solidify free liquids, resins, and sludge; and repackage the waste to meet WIPP Waste Acceptance Criteria (WAC). (The WAC are described in Appendix A.) Large steel boxes would be opened in this area, and plywood boxes within the large steel boxes would be removed to be processed individually. Culverts would be opened remotely, and drums would be removed and placed into a cell where they would be vented, purged with inert gas, and fitted with a filter vent before going to the verification area. Any gases vented from the drums would pass through the building exhaust system. In the verification area, drums and boxes would be assayed to determine curie content for inventory control and record purposes. Each container would then be x-rayed to verify compliance with the WAC. After being x-rayed, containers not conforming to the WAC would pass through an airlock into the remote waste-preparation cell. This cell would have lead-shielded viewing windows and a remote operator's console. All waste-preparation activities would be performed remotely with the aid of a telerobot. This robot would handle several tools, including a plasma arc torch, to size-reduce large objects. The telerobot would remove any objects identified in the x-ray process that do not meet the WAC. An electric worktable would be provided so that the telerobot can work on large, bulky objects. P-3 NEWLY GENERATED WASTE _ STORED WASTE g— CERTIFIABLE ...:. . 1 EXISTING. \ „ ^CERTIFICATION FACILITY g/; LOW-LEVEL WASTE hM^MM*ttMa«MUi TRU CERTIFIED f=^ NONCERTIFIABLE ■^ 1 ^'f^l 200 mph). The effective dose equivalent was calculated to be 2.0 rem, which is well below the DOE guideline of 25 rem. The upper-bound latent cancer risk to the total onsite and offsite populations would be about two additional deaths among the total population within 50 mi. This population is expected to experience about 110,000 cancer deaths during the same time frame from unrelated ("natural") causes. Table P.2.2 summarizes the consequences for postulated accidents in the TRU Waste Processing Facility. P-8 TABLE P.2.1 Summary of consequences from postulated accidents in the burial ground^ Effective dose equivalent Offsite Accident Curies released On-site Off-site population population (person-rem) (person-rem) maximally exposed individual (mrem) Winds^ 100 mph > 150 mph 2.1x10-2 4.2x10-2 1.6X10-"' 2.2x1 0-"" 4.4 6.3 6.3x10-2 7.3x10-2 Tornado 113-157 mph 158-206 mph 2.5x10-2 5.3x10-2 9.3 2.1x10^ 1.6x10^ 3.5x10^ 1.3x10-2 2.7 Fire Drum in culvert Drum on pad 1.7 5.0x10-^ 9.3x10^ 2.8x10^ 2.0x10"^ 6.1x10^ 4.4x10^ 1.3x10^ Drum rupture Internal pressure External pressure 5.0x10-^ 5.0x10-^ 2.8x10^ 2.8x10-^ 6.1x10^ 6.1 xlO-"" 1.3x10^ 1.3x10-"' ^ Estimated from the analysis of potential burial ground accidents reported in DPSTSA- 200-10, Supp. 8. ^ Straight winds. P-9 ^>^' ig TABLE P.2.2 Summary of consequences from postulated accidents in the TRU Waste Processing Facility^ Accident Effective dose equivalent Curies released Pu-238 Pu-239 On-site Off-site population population (person-rem) (person-rem) Offsite maximum individual (mrem) Winds'^ 100-150 mph > 150 mph Tornado 100-200 mph > 200 mph Earthquakes 0.09-0.2 g Vehicle crash Fire Drum rupture Internal pressure External pressure 4.3 8.8 4.7x1 0'^ 5.1x10^ 9.5x1 0'^ 7.3x10^ 5.2 5.7x1 0-*^ LgxIO"* 4.4x10^ 4.7x10"'' 1.5x10"^ 4.3x1 0'^ 5.0x10"^ 3.4x10^ 2.2x10-2 2.4x10-^ 1.7x10^ 8.7x10"^ 9.5x10"^ 7.3x10^ 4.3x10-^ 4.7x10-^ 3.4x10^ 4.3x10-^ 4.7x1 0"^ 3.5x1 0-"" 7.3x1 0^ 1.1x10^ 1.1x10^ 1.8x10^ 2.8x10^ 2.3x10"^ 2.5x1 0^ 2.0x10^ 4.3x1 0^ 1.1x10^ 2.1x10^ 5.5x10^ 9.3x10 4.2x10 4.3x10 2.5x10 1.1x10 1.1x10" ® Estimated from the analysis of potential ETWAF/WCF accidents reported in DPSTSA- 200-17, Rev. 1. Straight winds. P-10 m^^m^ P.3 HANFORD RESERVATION P.3.1 WASTE CHARACTERISTICS AND CURRENT MANAGEMENT METHODS TRU waste generated at the Hanford Reservation since 1970 has been retrievably stored. Most of this waste is contact-handled (OH) waste and is in 55-gal drums, stored as shown in Figure P.3.1. The containers are covered with plywood, plastic- reinforced nylon sheeting, and a 4-ft layer of uncontaminated soil to reduce surface radiation exposure rates. Hot cell remote-handled (RH) waste is stored in caissons such as those illustrated in Figure P.3.2. TRU waste unsuitable for asphalt pad or caisson storage because of size, chemical composition, security requirements, or surface radiation has been packaged in wooden, concrete, or metal boxes, and stored in dry waste trenches since approximately 1 973. Each trench is covered with plywood and vinyl plastic and backfilled with dirt (see Figure P.3.3). Newly generated TRU waste is stored in approved storage facilities. These aboveground buildings meet all Federal, State, and local regulations. P.3.2 RETRIEVAL OH TRU waste in retrievable storage trenches and aboveground buildings Is stored free of external contamination and packaged to maintain integrity for a minimum of 20 years. It is packaged so that the waste can be retrieved in an open environment without releasing airborne radioactivity. The soil overburden would be removed using conventional equipment and/or hand digging as required. Once the overburden is removed, the packaged waste would be removed by a forklift or crane. The current inventory of retrievably stored OH TRU waste would be removed and transferred for certification to a Waste Receiving and Processing Facility (Subsection P.3.3). Waste not directly certifiable would be processed within the Waste Receiving and Processing Facility to produce waste packages that would meet the WAG. Until about 1994 when the Waste Receiving and Processing Facility is scheduled to begin operation, newly generated TRU waste would be retrievably stored on pads or in buildings. Newly generated TRU waste would be retrieved and, if required, processed in the same manner as the existing retrievable TRU solid waste. After 1 994, all OH TRU waste would be processed and packaged to meet the WAG in the facility as it is generated. Special equipment would be used to recover the RH TRU waste in caissons. In the current retrieval scenario this equipment would not require an entry pit to gain access to the caissons. A recovery building would be positioned over the first caisson row and would contain a remotely operated manipulator and associated equipment. Movement of the building would require roadways. A new entry cut would be made P-11 314' PLYWOOD ^1/4" PLYWOOD VINYL COVER GRADE 55-GALLON DRUMS "^ 2' PVC PIPE BACKFILL .«^^^^S« mm^ ^ ASPHALT SLAB i FIGURE P.3.1 TRU WASTE ASPHALT PAD STORAGE P-12 91 cm DIA PIPE 10 cm CONCRETE FILLER GRADE 3.1 m ' FIGURE P. 3.2 TYPICAL CAISSON FOR TRU WASTE STORAGE P-13 5-8 m •14-20 m- 9 m 5 m 1.5 m' ^ EXISTING GRADE MINIMUM 1.3 m BACKFILL * DIMENSIONS FOR TYPICAL "DRY WASTE' TRENCH; BOXES. BARRELS, ETC. (LARGER DIMENSIONS ARE FOR CONTAMINATED 'INDUSTRIAL" SOLID WASTE TRENCH; FELLED PROCESS EQUIPMENT IN LARGE METAL OR CONCRETE BOXES). FIGURE P. 3. 3 TYPICAL TRU WASTE BURIAL TRENCH P-14 into the caisson. The retrieval operations would be controlled remotely from an auxiliary control room. A grappler housing equipped with a telescoping articulated boom would retrieve the caisson waste stored mainly in 1 -gal and 5-gal containers. An airlock and conveyor system would be used to transfer the remotely handled cask containing the retrieved caisson waste. This cask would be remotely sealed and decontaminated before placement on a truck. The cask would then be transported to a waste processing facility for conversion to a form suitable for geologic disposal. A small amount of retrievably stored and newly generated RH TRU waste would also require processing. This waste may be routed to a Special Handling and Packaging Facility designed to process RH TRU waste (not In the Waste Receiving and Processing Facility). This facility would be functionally similar to the Waste Receiving and Processing Facility, and its operations would include specific processes required to meet WAC requirements. P.3.3 WASTE RECEIVING AND PROCESSING FACILITY The major functions of the Waste Receiving and Processing Facility would include: 1) providing for examination, processing, packaging, and certification of retrievably stored OH TRU waste; and 2) providing for examination and certification of newly generated OH TRU waste for repository disposal. The Waste Receiving and Processing Facility is conceptually designed to support examination and certification (to the WAC) of CH TRU waste for permanent disposal and is scheduled to be constructed during the 1990s. Processing and packaging capabilities for CH TRU waste in retrievable storage would be provided in the Waste Receiving and Processing Facility. In estimating product costs, emissions, and volumes of waste, it is projected that 40 percent of all CH TRU waste would be reclassified as low-level waste after the TRU waste content of each pack is measured. The projected 40 percent of waste to be reclassified is based on engineering judgment and historical records. Waste process systems being considered include waste package inspection, assaying, repackaging, size reduction, compaction, sorting, shredding, and waste immobilization in grout. A conceptual process flow diagram for the Waste Receiving and Processing Facility using a shredding process without incineration is shown in Figure P.3.4. P,3.3.1. Waste Process Description P.3.3.1 .1 Receiving Dock . The first step in the waste package flow would be to offload the waste onto the receiving dock. The dock would be constructed to facilitate offloading of trucks by forklift and possibly by crane. Once offloaded, the waste packages would initially be inspected to determine whether incoming waste meets the WAC or whether further processing is required. For inspection, the receiving dock would be equipped with instruments that measure surface contamination, surface exposure rates, and physical dimensions. Waste packages with exposure rates greater P-15 RETRIEVE CH TRU WASTE " 1 SIZE REDUCTION IF NEEDED '' NDA & NDE >■ LLW 1 ^^ ^v^ A/0 <^CERT\F\f