OFFICE OF CIVILIAN RADIOACTIVE WASTE MANAGEMENT QA: N .../67531/metadc... · ACKNOWLEDGMENTS . . This document was authored by James 0.Duguid, Jerry McNeish, Vinod Vallikat, and Nelson

OFFICE OF CIVILIAN RADIOACTIVE WASTE MANAGEMENT QA: N

SPECIAL INSTRUCTION SHEET Page: 1 of: 1 I

This is a placeholder page for records that cannot be scanned or microfilmed 1. Record Date 09130197

2. Author Name(s) DUGUID JO, MCNEISH JA, VALLIKAT V, CRESAP D

10. Accession Number MOL,I99806 18.0474

3. Author Organization M+O

4. Title TOTAL SYSTEM PERFORMANCE ASSESSMENT SENSITIVITY' STUDIES OF U.S. DEPARTMENT OF ENERGY SPENT NUCLEAR FUEL

5. Document Number(s) A00000000-0 17 17-5705-000 17-01

6. Version REVISION 01

7. Document Type 8. Medium REPORT I DESIGN DOCUMENT PAPERlOPTIC

9. Access Control Code PUB

1 1. Traceability Designator VADD-VOL-03, CIDI-A00000000-017 17-5705-00017-01 ,WBS EM.4.01

12. Comments THIS ON OF A KIND DOCUMENT WITH COLOR CHARTS CAN BE LOCATED THROUGH THE RPC, THIS IS A VIABILITY ASSESSMENT RECORD-DO NOT DELETE

Exhibit AP-17.1 Q. 1 Rev. 1 1/22/96

WBS: EM.4.01

MOL. 19980618.0474

Civilian Radioactive Waste Management System Management and Operating Contractor

TOTAL SYSTEM PERFORMANCE ASSESSMENT SENSITIVITY STUDIES

OF US. DEPARTMENT OF ENERGY

SPENT NUCLEAR FUEL

September 30,1997

Prepared for:

National Spent Nuclear Fuel Program U. S. Department of Energy

Idaho Operations Office Idaho Falls, Idaho 834 10- 1 154

Prepared by: J. 0. Duguid, J. A. McNeish, V. Vallikat, D. Cresap, and N. J. Erb

TRW 1 180 Town Center Drive

MIS 423 Las Vegas, Nevada 89 134

Under Contract Number DE-ACO 1-9 1RWOO 134

WBS: EM.4.01 QA: N/A

Civilian Radioactive Waste Management System Management and Operating Contractor

TOTAL SYSTEM PERFORMANCE ASSESSMENT SENSITIVITY S W I E S

OF U.S. DEPARTMENT OF ENERGY

SPENT NUCLEAR FUEL

A00000000-017175705-00017, Rev. 01

September 30,1997

Prepared for:

National Spent Nuclear Fuel Program U. S. Department of Energy

Idaho Operations Office Idaho Falls, Idaho 834 10- 1 1 54

Prepared by: J. 0. Duguid, J. A. McNeish, V. Vallikat, D. Cresap, and N. J. Erb

TRW 1 180 Town Center Drive

MIS 423 Las Vegas, Nevada 89 134

Under Contract Number DE-ACO 1-9 1 RW00134

U.3. VEYAK 1 lV111\ 1 u r blr fin- r

SPENT NUCLEAR FUEL

A00000000-01717-570SOOOI7, Rev. 01

September 30,1997

Prepared by: Date: ZS, /%37

Prepared by: LAXc'LL Date: 9/29/4 7 JW. McNeish

Prepared by: f i fwd & I/, bLf FA Date: 5)/2 9/9 7 Vinod Vallikat

7,~ Prepared by: Date: c/26 57.7 Dale Cresap,

- National ~ p e h Nuclear

7

I . Fuel Program

b.

Prepared by: & b ate: 9/'4 /57 Nelson Erb

Approved by: p ) b & . Date: ' ,qh Ro ert W. Andrews, Manager Performance Assessment & ~ o d e l i n ~

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expresScd or implied, or assumes any legal liability or responsibility for accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infrnge privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

ACKNOWLEDGMENTS . .

This document was authored by James 0. Duguid, Jerry McNeish, Vinod Vallikat, and Nelson Erb (MTERA, Inc./Duke Engineering and Services) as part of the Civilian Radioactive Waste Management System (CRWMS), Management and Operating Contractor (M&O). Dale Cresap of the Idaho National Engineering and Environmental Laboratory (INEEL), National Spent Nuclear Fuel Program provided a first t of Chapter 3. Data in Appendix A were furnished by Larry Taylor of the National Spent Nuclear Fuel Program, INEEL, which is operated by Lockheed Martin Idaho Technologies Company (LMIT).

The document benefitted fiom reviews and suggested modifications by Henry Loo and D e ~ y Fillmore the National Spent Nuclear Fuel Program at MEEL, and for their technical support in interpretation of results. The draft document was reviewed by staff at Savannah River and Hanford as part of the NationaI Program. Comments were also provided by Abe Van Luik of the Yucca Mountain Site Characterization Project Ofice (YMSCO). These reviews greatly improved the final document. .

The work reported here was funded by the U.S. Department of Energy Environmental Management Division (EM), by the National Spent Nuclear Fuel Program under Work Breakdown Structure EM.4.01 - through the U.S. Department of Energy Office of Civilian ~adioactive Waste Management -- YMSCO # DE-AC01-91 RW00134 to TRW Environmental Safety Systems, Inc.

The support of ail of these individuals and organizations is gratefully acknowledged.

TABLE OF CONTENTS

..................................... ...... EXECUTIVE SUMMAFtY . : :

1 . INTRODUCTION ............................................... .......................................... 1.1 BACKGROUND

............................ 1.2 OBJECTIVES AND APPROACH ...... 1.3 TOTAL SYSTEM PERFORMANCE ASSESSMENT (TSPA)

.............. 1.4 RIP (REPOSITORY INTEGRATION PROGRAM)

2 . PERFORMANCE ASSESSMENT OF DOE SPENT NUCLEAR .................................................. FUEL (SW)

.................................. 2.1 SYSTEM COMPONENTS ......................... 2.2 REGULATORY CONSIDERATIONS

...................................... 2.3 SITE DESCRIPTION 2.4 REPOSITORY DESCRIPTION ............................. ;

............................. 2.5 NEAR-FIELD ENVIRONMENT ........................ 2.6 WASTE PACKAGE DEGRADATION

................. 2.7 RADIONUCLIDE TRANSPORT PROCESSES ............ 2.7.1 Engineered Barrier Release Conceptual Models

................................ 2.7.2 Geosphere Transport ....................... 2.8 BIOSPHERE AND RADIATION DOSE

.................. 2.9 BASE CASE CALCULATIONAL SCENARIO

................................................ 3 . SOURCE TERM .............................. 3.1 WASTE FORM CATEGORIES

........................... 3.2 RADIONUCLIDE MVENTORIES .................. 3.3 WASTE PACKAGE TYPES AND LOADING

................... 3.4 PHYSICAL PROPERTIES OF SPENT FUEL ..................... 3.5 WASTE FORM DISSOLUTION MODELS

................................... 3.5.1 Metallic Spent Fuel .................................... 3 . 5.2 Carbide Spent Fuel .................................. 3.5.3 Ceramic Spent Fuel

............................... 3.5.4 Commercial Spent Fuel .............................. 3.5.5 High-Level Waste Glass

...................... . 4 DOSE AT THE ACCESSIBLE ENVIRONMENT 4.1 DOSE ATTRIBUTED TO MDIVIDUAL WASTE

............................................ CATEGORIES ................. 4.1.1 Uranium Metal Spent Fuel (Category 1)

4.1.2 Uranium-Zirconium Alloy Spent Fuel (Category 2) ......... 4 .I . 3 Uranium-Molybdenum Alloy Spent Fuel (Category 3) .. : ....

.................. 4.1.4 Uranium Oxide Spent Fuel (Category 4)

ES- 1

iii

TABLE OF CONTENTS (Cont.)

..... 4.1.5 Uranium Oxide-Disrupted Clad Spent Fuel (Category 5) 4-4 4.1.6 Uranium-Aluminum Alloy Spent Fuel (Category 6) ......... 4-4

................. 4 .I . 7 Uranium Silicide Spent Fuel (Category 7) 4-4 4.1.8 High-Integrity Uranium-Thorium Carbide Spent Fuel

(Category 8) ........................................ 4-5 4.1.9 Low-Integrity Uranium-Thorium Carbide Spent Fuel

........................................ (Category 9) 4-6 4.1 . 10 Uranium and Uranium-Plutonium Carbide spent Fuel

....................................... (Category 10) 4-6 4.1.11 MixedOxideSpentFuel(Category11) .............. '. .... 4-6 4.1.12 Uranium-Thorium Oxide Spent Fuel (Category 12) ......... 4-7 4.1.13 Uranium-Zirconium Hydride Spent Fuel (Category 13) ...... 4-7

............... 4.2 DOSE ATTRIBUTED TO COMBINED WASTES 4-8 ............................... 4.3 COMPARISON OF RESULTS 4-9

................... 5 . EVALUATION OF SENSITIVITYAJNCERTAINTY 5-1 . ................. 5.1 ENGINEERED BARRIER RELEASE MODEL 5-1

............................... 5.2 TYPE OF WASTE PACKAGE 5-2 ............... 5.3 CLADDING FAILURE AND FREE WENTORY 5-3

................................... 5.4 DISSOLUTION MODEL 5-3 ........................ 5.5 WASTE FORM DISSOLUTION RATE 5-4

............................ 5.6 WASTE FORM SURFACE AREA 5-4. ....................... 5.7 WASTE PACKAGE FAILURE GROUP 5-5

................................. 5.8 SUMMARY OF RESULTS 5-5

........................... 6 . FINDINGS AND RECOMMENDATIONS 6-1

......................... ....................... 7 . REFERENCES : 7-1

......................................... APPENDIX A DOE SNF DATA A-1

LIST OF FIGURES

Location Map of the Potential Repository Site at Yucca Mountain, ...................................................... Nevada 1-9

Schematic Diagram of Components that Contribute to Waste Isolation at ................................................ Yucca Mountain 1 - 1 0

Schematic Diagram of the Ground-Water Transport Pathway from the ................................... Potential Repository to Humans 1 - 1 1

Schematic Diagram of the Models Used in a Total System Performance ............................................ Assessment (TSPA) 1-1 2

Flow of Information and Results through the Domains Modeled for the '

................. Analyses of a Potential Repository at Yucca Mountain 1 - 13

................. Hierarchy of Models Used in Performance Assessment 1 - 1 4

Diagram of Information Flow to the Total System Performance Assessment ................... Model : .................................. 1-15

Flow of Information from Process and Subsystem Models into the Repository ........................................ Integration Program (RIP) 1 - 1 6

\

Schematic Cross Section Through Yucca Mountain Showing the Potential Repository in Relationship to Faults, Stratigraphic Units, and the Ground-

............................... Water Table (AAer Ortiz et a]., 1985) 2- 1 5

Major Geologic Units, Lithology, and Hydrogeological Units at Yucca Mountain, Nevada (Modified from Montazer and Wilson, 1984 using

............................................. Sawyer 9 al., 1995) 2-1 6

Regional Ground-Water Flow System in the Vicinity of Yucca Mountain Showing Three Ground-Water Subbasins (Modified from DOE, 1988) .... 2-1 7

Idealized Geohydrologic Cross Section from Yucca Mountain to Eagle ................................. Mountain (After Czamecki, 1989) 2- 1 8

........................ Schematic Plan View of the Repository Layout 2-19

.......................... Schematic Diagram of Waste Emplacement 2-20

LIST OF FIGURES (Conk)

2.5- 1 Engineered Barrier System Processes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 1

2.6-1 History of First Pit Development Though Waste Packages . . . . . . . . . . . . . 2-22

2.6-2 Average Number of Pits Through Waste Packages in Eight Repository Failure Groups . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-22

, '

2.7-1 (a) Conceptual Mode1 for Diffusive and Advective Release from Both the Waste Package and Other Components of the Engineered Barrier System ("Drips on the Waste Form") . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 2-23

2.7- 1 (b) Conceptual Model for Diffusive Release from the Waste Package and Both Diffusive and Advective Release fiom Other Components of the . Engineered Barrier System ("Drips on the Waste Package") . . . . . . . . . . . . . 2-24

2.7-l(c) Conceptual Model for Diffusive Release from Both the Waste Package . and Other Components of the Engineered Barrier System ("Capillary Barrier Effect") . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-25

3.1-1 DOE Spent Fuel Categohes for Total System Performance Assessment (TSPA) (Modified form Stroupe, 1 997) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 14

3.3-1 Waste Package Used for Disposal of High-Level Waste and Co-Disposal of DOESNF .................................................... 3-15

Waste Package Used for Co-Disposal of DOE SNF . . . . . . . . . . . . . . . . . . . . 3- 16

Waste Package Used for Disposal of 21 Uncanistered Spent Fuel Assemblies fiom a Pressurized Water Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 7

Waste Package Used for Disposal of 44 Uncanistered Spent Fuel Assemblies fiom a Boiling Water Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-18

Comparison of Dissolution Rates for High-Level Waste Glass and Commercial, Metallic, Carbide, and Ceramic Spent Fuels . . . . . . . . . . . . . . . 3-1 9

Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 1 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 1

Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 1 Spent Fuel . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 1

LIST OF FIGURES (Cont.)

Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 1 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 2 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 ' Years from Category 2 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 2 . . . , . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 3 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 3 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 3 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 4 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 4 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 4 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 5 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Category 5 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Commercial Spent Fuel Equivalent to Category 5 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Category 6 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . .

vii


Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 6 Spent Fuel . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 6 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 7 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 7 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 7 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 8 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 1 Q0,000 Years fiom Category 8 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Commercial Spent Fuel Equivalent to Category 8 . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 1,000,000 Years fiom Category 8 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 1,000,000 Years fiom Category 8 Spe'nt Fuel Using the Radionuclide Inventory h m Duguid et al., 1997 . . . . . ; . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 9 Spent Fuel and HLW . . . . . . . . . . . . . . . . . . . .,. . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 9 Spent Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 9 . . . . . . . . . . . .

4.1-30 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 10 Spent Fuel and HLW . . . . . . . . . . . . . . .. . . . . , . . . . 4-25

viii

5 : :',

/ LIST OF FIGURES (Coat.)

4.1 -3 1 Expected-Value Dose History at the Accessible Environment Over 100,000 ................................ Years fiom Category 10 Spent Fuel 4-26

4.1 -32 Expected-Value Dose History at the Accessible Environment Over 100,000 ,

........... Years from Commercial Spent Fuel Equivalent to Category 10 4-26

4.1 -33 Expected-Value Dose History at the Accessible Environment Over 100,000 ........................ Years from Category 11 Spent Fuel and HLW 4-27

4.1-34 Expected-Value Dose History at the Accessible Environment Over 100,000 ................................ Years fiom Category 1 1 Spent Fuel 4-27

4.1-35 Expected-Value Dose History at the Accessible Environment Over 100,000 ........... Years h m Commercial Spent Fuel Equivalent to Category 1 1 4-28

4.1 -36 Expected-Value Dose History at the Accessible Environment Over 1 00,000 ........................ Years from Category 12 Spent Fuel and HLW 4-28

4.1-37 Expected-Value Dose History at the Accessible Environment Over 100,000 ................................ Years from Category 12 Spent Fuel 4-29

4.1 -38 Expected-Value Dose History at the Accessible Environment Over 100,000 ........... Years fiom Commercial Spent Fuel Equivalent to Category 12 4-29

4.1-39 Expected-Value Dose History at the Accessible Environment Over 100,000 ........................ 'Years from Category 13 Spent Fuel and HLW 4-30

4.1-40 Expected-Value Dose ist tory at the Accessible Environment Over 100,000 ................................. Years from Category 13 Spent Fuel 4-30

4.1-41 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel guivalent to Category 13 ........... 4-3 1

4.2-1 Expected-Value Dose History at the Accessible Environment Over 100,000 ................................. Years from 8,745 MTHM of HLW 4-3 1

4.2-2 Expected-Value Dose History at the Accessible Environment Over 100,000 ................................. Years fiom 7,000 MTHM of HLW 4-32

4.2-3 Expected-Value Dose History at the Accessible Environment Over 100,000 ................. • Years from 63,000 MTHM of Commercial Spent Fuel 4-32


Expected-Value Dose History at the Accessible Environment Over 100,000 years from 2,436 hdTHM of DOE SNF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom 2,436 MTHM of Commercial Spent Fuel . . . . . . . . . . . . . . . . . .

Expected-Value Dose History at the Accessible Environment Over 100,000 Years for a Repository Containing 63,000 MTHM of Commercial Spent Fuel, 2,436 MTHM of DOE SNF, and 8,745 MTHM of HLW (74,181MTHM) ............................................... Expected-Value Dose History at the Accessible Environment Over 100,000 Years for a Repository Containing 63,000 MTHM of Commercial Spent Fuel and 7,000 MTHM of HLW (70,000 MTHM) . . . . . . . . . . , . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years for a Reposit* Containing 63,000 MTHM of Commercial Spent Fuel and a Combined 7,000 MTHM of DOE SNF and HLW . . . . . . . . . . . .

Expected-Value Dose History at the Accessible Environment Over 100,000 Years for a Repository Containing 70,000 MTHM of Commercial Spent Fuel and HLW with 2,333 MTHM of Spent Fuel in Waste Package Failure Group 1 and 60,667 MTHM of Spent Fuel and 7,000 MTHM of HLW Distributed Evenly Among Eight Waste Package Failure Groups . . . . . . . . .

Expected-Value Dose History at the Accessible Environment Over 100,000 Years from One Package of Commercial Spent Fuel as Compared to One Package of Category 1,2, and 3 DOE SNF . . . . . . . . . . . . . . . , . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 1 00,000 Years fiom One Package of Category 4,5,6, and 7 DOE SNF . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from One Package of Category 8,9, and 10 DOE SNF . . . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from One Package of Category 1 1,12,-and 13 DOE SNF . . . . . . . . . . Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 1,2,3,4,5,6, and 7 DOE SNF . . . . . . . . . . . . . . . . . .


Expected-Value DO& History at the Accessible ~ n v i r o h e n t Over 100,000 YearsfromCategory8,9,10,11,12,and13DOESNF ................ 4-38

Comparison of Peak Dose at the Accessible Environment for the Engineered Barrier Release Model "Drips on the Waste Formn, "Drips on the Waste Package", and "No Dripsw (Capillary Barrier) for

............................. ............. Category 1 Spent Fuel ; 5-7

Sensitivity of Peak Dose at the Accessible Environment to Diffusion ,

Through an Invert with Assumed Properties of Crushed Tuff, Gravel, ................................. or Sand for Category 1 Spent Fuel 5-7

Sensitivity of Peak Dose at the Accessible Environment to Assumed . .................. Lifetime of the Drip Shield for Category 1 Spent Fuel 5-8

Sensitivity of Dose at the Accessible Environment to Waste Package Type ........................................ for Category 4 Spent Fuel 5-8

Sensitivity of Dose at the Accessible Environment to Percent of Fuel .......................... Cladding Failure for Category 1 Spent Fuel 5-9

Sensitivity of Dose at the Accessible Environment to Assumed Dissolution Model (Metal, Carbide, Ceramic, and Oxide) for Category 9 Spent Fuel ... 5-9

Sensitivity of Dose at the Accessible Environment to Assumed Dissolution .

Rate of Category 1 Spent Fuel Assuming the Waste Package Release Model .................................... "Drips on the Waste Package" 5- 1 0

Sensitivity of Dose at the Accessible Environment to Assumed Dissolution Rate for &egov 1 Spent Fuel Assuming the Waste Package Release Model

....................................... "Drips on the Waste Form" 5- 1 0

Sensitivity of Dose at the Accessible Environment to Assumed Surface Area of Category 1 Spent Fuel Assuming the Waste Package Release Model .

.................................... "Drips on the Waste Package" 5-1 1

Sensitivity of Dose at the Accessible Environment to Assumed Surface Area of Category 5 Spent Fuel Assuming the Waste Package Release Model

.................................... "Drips on the Waste Package" 5-1 1

LIST OF FIGURES (Cant.)

5.6-3 Sensitivity of Dose at the Accessible Environment to Assumed Surface ~ r i for Category 5 Spent Fuel Assuming the. Waste Package Release Model "Drips on the Waste Form" ...................... ................. 5-1 2

5.7-1 Sensitivity of Dose at the Accessible Environment to Waste Package Failure Group for Category 1 Spent Fuel ............ : ............... 5- 12

xii

LIST OF TABLES

Major Parameters Used in Assessing the Performance of DOE SNF in ............... a Potential Repository at Yucca Mountain. Nevada .... ; 2-26

....................... Categories and Typical Members of DOE SNF 3-20

Summary of Metric Tons Heavy Metal (MTHM) for Each Category of ................... DOE SNF at INEEL. Savannah River. and Hanford 3-21

................ Radionuclide Inventory for Each Category of DOE'SNF 3-22

Radionuclide Inventory for Commercial Spent Fuel and High-Level Waste ........................................................ 3-24

Number of Waste Packages of Each Type md Total Number of ........................... Packages Used for Disposal of DOE SNF 3-25

.................................. Physical Properties of DOE SNF 3-26

.................. Average MTHM of DOE SM: in Each Waste Package 4-39

......................... Diffusion Characteristics of Invert Materials 5-13

xiii

EXECUTWE SUMMARY

ES1 APPROACH '

A total system performance assessment (TSPA) for disposal of U. S. Department of Energy (DOE) v t nuclear fuel (SNF) in the potential Yucca Mountain repository was conducted. The approach used was similar to that of TSPA-1995 (Andrews et al., 1995) with site, waste package, and repository model parameters updated to the cunent understanding of the repository. The base case for these analyses is a repository that contains 70,000 metric tons heavy metal (MTHM) of waste, Composed of 63,000 MTHM of commercial spent fuel and 7,000 MTHM of HLW with a thermaI loading of 83 MTHM/acre of commercial spent fuel. The DOE SNF is grouped into 15 performance assessment categories, 13 of which were analyzed for this report. The 13 categories of DOE SNF analyzed are:

Category 1 Uranium metal spent fuel Category 2 Uranium-Zinium alloy spent fuel Category 3 Uranium-Molybdenum alloy spent fuel Category 4 Uranium oxide spent fuel Category 5 Uranium oxide-disrupted clad spent fuel Category 6 Uranium-Aluminum aIloy spent fuel Category 7 Uranium silicide spent fuel Category 8 High-integrity Uranium-Thorium carbide spent fuel Category 9 Low-integrity Uranium-Thorium carbide spent fuel Category 10 Uranium and Uranium-Plutonium carbide spent fuel Category 1 1 Mixed oxide spent fuel Category 12 Uranium-Thorium oxide spent fuel Category 13 ' Uranium-Zirconium hydride spent he1

. Category 14 (Sodium bonded spent fuel) and Category 15 (Navy spent fhel) were not analyzed. For each category of DOE SNF the radionuclide inventory was obtained fiom ORIGEN2 analysis of specific spent fuels that are typical of the category. The radionuclide inventory used for each category was based & disposal of the spent fuel in the year 2030.

For convenience, the analyses in this study were done in terms of dose history fiom drinking water at the accessible enviroment (dose to an individual drinking two liters of water per day fiom a well 5,004) meters down gradient fiom the repository). This approach was used to ease the burden of analysis of dose fiom all sources at 20 kilometers down gradient fiom the repository which is currently required by the DOE guidance. The simplified dose model and the shorter saturated-zone transport pathway are adequate for the comparison (in a relative sense) of the dose fiom DOE SNF to that fiom commercial spent fuel and/or HLW. Figure ES 1-1 shows a schematic of the transport

- pathway fiom the repository horizon to the accessible environment used in this study. This transport pathway for radionuclides, after the waste packages have failed, is through the engineered barrier system, through approximately 204) meters of the unsaturated zone, and laterally for 5,000 meters through the saturated zone to the assumed well (Figure ES 1-1).

For this study the percolation flux through the repository horizon was abstracted fiom the calibrated site-scale model for the unsaturated zone. For the six vertical columns used in the abstraction the flux ranges from 4 to 10 d y r with an average flux of 6.2 mmlyr. The repositow' drifts were assumed not b be backfilled, and no change of climate was assumed The reason for not including climate cycles is that the site-scale model has not yet been calibrated to examine the transient effects of a climate cycle.

For the analyses, 30% of the packages are assumed to be locad where drips occur and the remaining 70% of the waste packages were at locations where no drips occur. Where drips occur the conceptual engineered barrier release model "drips on the waste package" was used, and where no drips occur the conceptual release model for a capillary bamer was used. For both cases, release h m the fhiled waste package was assumed to be by diffusion, and for the capillary barrier there is an additional difkion pathway through the invert. Because the assumption that 30% of the waste packages are located where drips occur is conservative additional analyses were conducted assuming that 10% of the packages were located under drips (90% had no drips). Also, the application.of the conceptual waste package release model "drips on the waste package", over the long time period considered is not conservative because at some point in time the waste package will have degraded to a point where the percolation will be through it. In other words, at some point in time, the conceptual release model should switch from "drips on the waste package" to "drips on the waste form". The conceptual model "drips on the waste form" assumes that the percolation flux is through the waste (there is no diffusion pathway). For some of the sensitivity analyses, where the results are controlled by the diffusion pathway the conceptual model "drips on the waste form" was also used (i.e., dissolution rate and fuel surface area).

Transport parameters were also updated to the current understanding of the repository environment. In particular, the solubility of Neptunium was reduced by two orders of magnitude based on a reanalysis of existing data (Sassan. and Siegmann, 1997). This change reduced the calculated peak dose from Neptunium (form that presented in TSPA-1995) to below that fiom Iodine and Technetium. -This reduction allows for a shorter analysis period of 100,000 years because the peak dose from Iodine and Technetium occur at about 20,000 years.

The waste packages were assumed to have 75% galvanic protection for 90% of the packages when they are distributed evenly among the eight waste package failure groups. Waste package failure was assumed to occur when the waste package developed the first through-going pit, and the release from a failed package over time was dependent on the number of pits through the package. When all eight waste package failure groups are assumed, the first package failures begin to occur after about 600 years, and about 85% of the packages have failed in the &t 1,000,000 years. For waste package failure Group 1 the waste packages fail at between 600 and approximately 10,000 years. For this study the DOE SNF was assumed to be in packages in waste package failure Group 1 which is very conservative, and virtually neglects the effects of galvanic protection. This assumption was made to avoid the calculational difficulty of distribution of small quantities of DOE SNF evenly . among eight waste package failure groups.

The dissolution of the waste form in the failed waste package was analyzed using waste form

specific dissolution models. When more than one waste form was contained in the package the dissolution of each was calculated independently (i.e., spent fuel and HLW in a co-disposal package). No interaction bet&een waste form was assumed. .

The DOE SNF that was evaluated was assumed to be packaged for disposal in HLW and co-disposal packages with an assumed maximum fissile content of 14.4 kg per package. The co-disposal concept is to dispose of spent fuel and HLW in the same waste package in order to reduce the likelihood of criticality. Low enriched DOE SNF was assumed to be disposed in HLW packages that contain four canisters of spent fuel. Co-disposal of the DOE SNF required 8,745 MTHM of . HLW for the packaging assumptions made in this study. Thus, the entire repository analyzed in this study contains 63,000 MTHM of commercial spent fuel, 2,436 MTHM of DOE SNF, and 8,745 MTHM of HLW.

Each of the 13 categories of DOE SNF that was analyzed was assumed to be disposed of in the base case repository environment. The base case repository was assumed to contain 70,000 MTHM of waste (63,000 MTHM of commercial spent fuel and 7,000 MTHM of HLW). For ease in analysis, the amount of waste in each category of DOE spent fuel, in MTHM, was assumed to be added to the inventory of the base case repository. Each category of DOE SNF was analyzed with and without the co-disposed HLW, and the results of the latter case were compared to those from an equivalent amount of commercial spent fuel on an MTHM basis.

The entire repository containing the DOE SNF and co-disposed HLW (74,181 MTHM) was analyzed and the results were compared to those of the base case. The dose attributed to the DOE SNF and tbe co-disposed HLW was also scaled to examine the effects of a combined 7,000 MTHM (2,333 MTHM of DOE SNF for Categories 1 through 13 and 4,667 MTHM of HLW) and added to the dose attributed to 63,000 MTHM of commercial spent fuel to examine the effects of the larger repository.

Sensitivity analyses were conducted in order to gain a better understanding of the results and to examine the -effects of uncertain parameters on the peak dose at the accessible environment. Sensitivity analyses were conducted to examine fhe effects of the engineered barrier release model, cladding fidlure, fiee radionuclide inventory, dissolution model, dissolution rate, spent fuel surface area, and waste package fkilure group.

ES2 RESULTS

For the analyses of individual categories of DOE SNF and for the composite of the 13 categories the waste packages were assumed to be in waste package failure Group 1 (i.e., the most conservative waste package failure group). Packages in this group all fail between 600 and 10,000 years. This assumption was made to avoid distributing small quantities of DOE SNF evenly among the eight waste package failure groups. Assuming that the DOE SNF is in waste package failure Group 1 causes an increase in peak dose at the accessible environment of about an order of magnitude as compared to distribution of the waste evenly among all eight waste package failure groups (Figure ES 2-1). This figure is for Category 1 DOE SNF that is composed predominantly of N Reactor spent

fuel (88% on an MTHM basis). The difference in peak dose at the accessible environment between the two results is about an order of magnitude (with cli-dbution among the eight groups being lower).

The effects of the assumption of the percentage of waste packages that are located where drips occur is shown in Figure ES 2-2. The difference between assuming 30% and 10% drips is about a factor of five for Category 1 spent fuel.

A composite of the 13 categories of DOE SNF (2,436 MTHM) and the co-disposed HLW (8,745 '

MTHM) was analyzed. The dose attributed to the co-disposed high-level waste (HLW) was then removed and the results of the 2,436 MTHM of DOE SM; are shown in Figure ES 2-3. These results are for the assumption of disposal of all of the DOE SNF in waste package failure Group 1 and 30% of the waste packages located where drips occur.

The base case used in this study, 63,000 MTHM of commercial spent fuel and 7,000 MTHM of HLW distributed evenly among eight waste package fdure groups, is shown in Figure ES 2-4. The higher peak d o e seen in Figure ES 2-3 as compared to that of Figure ES 2-4 is caused by the conservative assumption that all of the DOE SNF is disposed in waste package failure Group 1.

When the conservatism of assuming the spent fuel is in waste package failure Group 1 is removed the peak dose will be reduced. Figure ES 2-5 shows results for distribution of the &e evenly

1 among the eight waste package failure groups for 30% of the waste packages located where drips occur. Th-. results are for the base case, for 2,436 MTHM of DOE SNF (composite of Categories 1 through 13), and for 8,745 MTHM of HLW (the HLW necessary to co-dispose the DOE SNF under the assumptions made in this study). A surrogate for the composite of the 13 categories of DOE SNF was used to distribute the spent he1 evenly among the eight failure categories. The surrogate was composed of Categories 1,4,5,6,8, and 1 1. These categories of DOE spent fuel are the significant contributors to the total dose fiom the DOE SNF at the accessible environment. The peak dose fiom the 13 Categories of DOE SNF is about two orders of magnitude lower than that of the base case, and is nearly the same as that fiom the HLW needed to co-dispose the 13 categories of spent fuel.

The results of the TSPA and the sensitivity analyses produced the following findings:

The dose at the accessible environment fiom the composite of 13 categories of DOE SNF . (2,436 MTHM) is somewhat less than that h m an equivalent amount of commercial spent .

fuel.

a The largest contributor to peak dose at the accessible environment from the composite of DOE SNF (2,436 MTHM) is Category 1 DOE SNF (2,136.8 MTHM) which is largely &mposed of N Reactor spent fuel. Compare the peak dose on the upper c w e of Figure ES 2-1 to the peak dose on Figure ES 2-3, they are approximately the same.

Distribution of the* composite (2,436 MTHM) of DOE SNF evenly among eight waste

package failure groups lowers the peak dose by about an order of magnitude. See Figure ES 2-1. The peak dose from the composite of DOE SNF, under this assumption, is about the same as that from the HI;W used for co-disposal(8,745 MTHM). See Figure ES 2-5. The change h m 30% of the waste packages being located where drips occw to 10% drips reduces the peak dose by about a factor of five Figure ES 2-2.

Six categories of DOE SNF contribute significantly to the dose history at the accessible environment fiom the composite of 13 categories of DOE SNF, Category 1,4,5,6,8, and 1 1. Of these six categories, Category 1 and Category 6 are the largest contributors to total dose at the accessible environment.

8 Dissolution rates of the waste based on assumed dissolution models range from approximately 102 to 105 dm?. The slowest rates occur after the repository temperature returns to near ambient (i.e., after about 10,000 years). The dissolution rates in descending order are: metallic fuel under wet oxic conditions, metallic fuel under humid oxic conditions, oxide spent fuel HLW glass, ceramic spent fuel, and carbide spent fuel. However, this five order of magnitude difference is muted by the diffusion rate out of the waste package. For example, the slow diffusion rate causes the release rate from packages containing different fbel types to be identical for different dissolution rates as long as there are more radionuclides dissolved than can be transported fiom the package by diffusion.

The assumption of different engineered barrier release models has a significant effect on the dose at the accessible environment. The peak dose assuming that 30?4 of the waste packages are- located where drips occur (70% have no drips) and assuming that all of the was'te package have no drips lowers the peak dose by more than an order of magnitude. When there are no drips the only mechanism for release fiom the engineered barrier system is by diffision out of the waste package and through the invert. The assumption that the drips occur on the waste form rather than the package causes the dose peak fiom the alteration controlled radionuclides (Iodine and Technetium) to occur earlier and to be a smaller value. This peak occurs earlier because the advetion is more rapid than d i i i o n , and it is lower because of depletion of alteration controlled radionuclides fbm the first waste packages that fail. For this case the Neptunium peak occurs at about 60,000 years and is somewhat higher than that from the alteration controlled radionuclides.

For the assumption of a capillary barrier (no drips) the dose at the accessible environment was found to change significantly based on the type of material used to construct the invert. .

Dose is lowered by nearly two orders of magnitude by using coarse sand rather than crushed tuff because the lower saturation of the sand produces a smaller coefficient of diffusion. When a capillary barrier effect is assumed to be produced using a drip shield, the lifetime of the drip shield becomes very important. If all of the drip shields are assumed to fail instantaneously at the end of their life, the peak dose is moved out in time an amount equal to the assumed lifetime of the drip shield (i.e., there is very little decay because of the relatively long half-life of the radionuclides contributing to the peak). Th'us, either the drip shields must last for a long period of time to reduce the peak, or they would have to fail over

a long period of time in order to allow for a significant decay of the Technetium. Technetium has a half-life of more than 200,000 years. Ifthe drip shields are assumed to last forever, then the effects can be simulated by assuming no drips (the capillary banier).

The assumed percentage cladding failure has a significant effect on peak dose at the accessible environment. However, the assumption of 100% cladding failure for DOE SNF is appropriate because of the c ~ t e r i s t i c s of the two categories of spent fbel that contribute much of total dose, Category 1 and Category 6. Category 1 is largely composed of N Reactor spent he1 which has a significant amount of cladding failure h m pool storage at Hanford, and Category 6 DOE spent fbel has Allnninum cladding that would be expected to degrade rapidly after waste package failure (Appendix A). Category 5 spent fbel also contributes significantly to the total dose and it has disrupted cladding. Also, for this study no credit was taken for the canister that contains the waste which is a simplifying and conservative assumption.

The effects of assumed dissolution rate and d a c e area are small for Category 1 DOE SNF. Increasing the assumed dissolution rate of Category 1 SNF does not change the peak dose because release h m the failed waste package is controlled by diffusion when the conceptual model "drips on the waste packagen is assumed Increasing the d a c e area causes lower broader dose peaks because of dilution in the increased volume of water (volume = surface area x film thickness) and depletion of the source term from waste packages that fail early. When the conceptual waste package release model "drips on the waste form" is assumed the dose at the accessible environment does not. change significantly for either a change of dissolution rate or of surface area This is because the alteration controlled radionuclides (Iodine and Technetium) are released into the flux through the waste package at the alteration rate and the solubiiity limited radionuclides (Neptunium) is released into the flux according the their solubility.

b The- dose at the accessible environment increases significantly (about an order of magnitude for the 3 3 categories of DOE SMF) when the conceptual waste package release model "drips on the waste form" is assumed in place of the conceptual model "drips on the waste package". The assumption of the model "drips on the waste package" does not give conservative results because at some stage of waste package failure flux will be though the package rather than around it. This eliminates the diffusion pathway. There may be a need to develop a switch from the conceptual release model of "drips on the waste package" to "drips on the waste form" at some point in time during the analyses.

2

Measure Cumulative Release (Ci)

and Peak Dose (rernlyr)

at

Unsaturated-

M Y' 4

Saturated-

Transport -

Figure ES 1-1 Schematic Diagram of the Ground-Water Transport Pathway fkom the Potential Repository to Humans.

1 0-8 I o - ~

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure ES 2-1 Sensitivity of Dose at the Accessible Environment to Waste Package Failure Gmup for Category 1 Spent Fuel.

0 20,000 40,000 60,000 80,000 100,000 Time (years) 1

Figure ES 2-2 Sensitivity of Dose at the Accessible Environment to Percentage of Waste Packages I 1

Located where Drips Occur for Category 1 Spent Fuel.

ES-8 1

. i

10s 0 20,000 40,000 60,000 80,000 100.000

Time (years)

Figure ES 2-3 Expected-Value Dose History at the Accessible Environment Over 100,000 Years ' fkom 2,436 MTHM of DOE SNF Disposed in Packages of Waste Package Failure Group 1.

Figure ES 2 4 Expected-Vahe Dose History at the Accessible Environment Ova 100,000 Years for a I&posito~~ Contaking 63,000 MTHM of Commercial Spent Fuel and 7,000 MTHM of HLW (70,000 MTHM Distributed Evenly Among Eight Waste Package Failure Groups).

Time (yrs)

Figure ES 2-5 Expected-Value Dose Histories at the Accessible Environment Over 100,000 Years for the Base Case, for 2,436 MTHM of DOE SNF, and for 8745 MTHM of HLW (Distributed Evenly Among Eight Waste Package Failure Groups).

The Yucca Mountain site is located in southern Nevada about 120 kilometers northwest of Las Vegas (Figure 1-1). The potential repository for disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) at Yucca Mountain would be located in the unsaturated zone about 200 meters above the water table. kigure 1-2 shows a schematic d i a w of the important components that contribute to waste isolation at Yucca Mountain. These components are a desert environment, unsaturated overburden, repository host rock, engineered barriers, waste package, waste form, unsanrrated rocks (along the transport pathway below the repository), and saturated rocks (along the transport pathway to the point where regulatory standards apply). Any aqueous release of radionuclides fiom the repository would have to be transported downward approximateIy 200 meters through the unsaturated zone ('2) to the water table and then laterally in the saturated zone (SZ) to locations where they might be accessed by humans. A schematic diagram of the repository and the potential transport pathway to humans is shown in Figure 1-3. The analysis of the potential long-tenn dose to humans fiom future repository releases is called a performance assessment and, if done for the entire engineered and natural system, is called a total system performance assessment (TSPA).

1.1 BACKGROUND

A TSPA requires, within a regulatory framework, the analysis of both the engineered and natural systems to determine the potential long-term release of ~ o n u c l i d e s from the repository to a location where a regulatory standard is applied. For convenience the sensitivity anal$es in A s report are based on dose to an individual fiom drinking water derived fiom a well at the accessible environment. The definition of the accessiile environment is established through regulations of the Nuclear Regulatory Commission (NRC) and the Environmental Protection Agency (EPA). It is considered to be bounded by the pund sdace above the repository and by a cylinder with a radius up to five kilometers fiom the repository in the lateral dimension (1 0 CFR 60 and 40 CFR 191, respectively). Thus, for a repository in the unsaturated zone, transport pathways to the accessible environment would be upward to the ground surface for gaseous radionuclides, and downward to the water table and laterally in the saturated zone for five kilometers for aqueous phase radionuclides. In the analyses conducted for this study only the dissolved aqueous phase radionuclides are considered.

. The EPA regulation, 40 CFR 19 1 (EPA, 1993) only applies to the Waste Isolation Pilot Plant (WJPP) in New Mexico, and to geologic repositories other than Yucca Mountain (Public Law 102- 486, The Energy Policy Act of 1992). The Energy Policy Act of 1992 initiated a process where the National Academy of Sciences WAS) recommended new environmental standards based on dose (NAS, 1995), and the EPA is currently in the process of promulgation of standards for Yucca Mountain based on the NAS recommendations. When the EPA standards are approved, the NRC will revise 10 CFR 60 (NRC, 1993) to conform. Until new standards for Yucca Mountain an available the Department of Energy (DOE) has issued interim guidance for prfobance assessment that requires the calculation of dose fiom all sources at a location 20 kilometers down gradient fiom the potential repository. In order to reduce the calculational burden of the longer transport pathway

and the analysis of all dose pathways, the sensitivity studies in this report are presented in terms of dose fiom drinking water at a location five kilometers fiom the repository.

The primary mechanism for movement of radionuclides from the repository to the accessible environment is expected to be by ground-water after the waste packages have failed. Disruptive events, such as human intrusion, have been analyzed and found to have only minor consequences (Wilson et al., 1994). The waste is assumed to be disposed of in metallic cylindrical packages emplaced on their side in drifts excavated in the rock of the unsaturated zone. In order for radionuclides to be released, the waste container and other engineered barriers (that may be included in the package or in the drift) must fail to either allow fonnation of a water film fiom moisture in the drift on the waste form through which release can occur by diffusion or to allow advection (flow) through the failed package. Radionuclides leached from the wastes (commercial spent hel, DOE spent fuel, and high-level waste (HLW) glass), may travel out of the failed barrier system either by -ion or advection, migrate to the water table h u g h the unsaturated zone, and be transported to the accessible environment in the saturated ground-water flow.

The flow bercolation flux) through the unsaturated repository horizon arises from the infiltration of precipitation at the ground surface minus the amount that is returned to the atmosphere through evapotranspiration. The small portion of the infiltration that remains percolates downward to the water table (top of the saturated zone) under the control of capillary and gravity forces. The heat fiom the radioactive decay of the waste influences this downward percolation and must be considered in performance assessment. The amount of heat present depends on the design of the repository (i.e., the spacing of repository drifts, the spacing and thermal loading of the waste packages, the type of waste, and the age of the waste). The migration of radionuclides &ough the unsaturated zone is a hct ion of the percolation flux, dispersion, and retardation of radionuclides by sorption to mineral d e s along the flow path, or by d i h i o n into slower percolating zones (diffusion fiom the frslcnues into the rock matrix where flow is slower). Sorption is a function of the chemistry of the specific radionuclide, the chemistry of the fluid, and the chemistry of the minerals along the flow path. Dispersion is the spreading of the plume caused by the tortuous paths that the radionuclides follow thro*& the openings (pores and fhctures) in the rock.

At.the water table, the concentration of radionuclides is diluted by the flow in the saturated zone. The radionuclides are then transported through the saturated zone to a defined point where their concentration can be compared to regulatory standards. Along the transport pathway, the concentration of radionuclides is reduced by dispersion (spreading of the plume), and deiayed by ietardation (lagging behind the flow due to sorption and matrix diffusion). The delay caused by retardation may provide time for reduction of concentration through radioactive decay, depending on the delay time in relationship to the half-life of the specific radionuclide.

At the accessible environment, the concentration of radionuclides over time is compared to performance measures such as: (1) cumulative release of specific radionuclides (the cumulative amount transported across the boundary of the accessible environment) over some time period, or (2) dose to an individual over some time period. Dose to an individual can be calculated based on a scenario that the individual is sustained by using ground water fiom a well at the accessible

1 -2

environment for production of all food, for household use, and for drinking water, or may be based on drinking water alone. The peak dose b m a specific radionuclide occurs at the time of peak concentration of that radionuidide in the ground water. Peak cdhc&ntration is controlled by all of the processes discussed above (is., peaks for different chemical p i e s occur at different times due to differences in solubility and sorption along the transport path). The concentration of radionuclides at the accessible environment as a function of time can then be converted to a performance measure (cumulative release across the accessible environment boundary over some time period or to dose history of an individual using the ground water over some time period). The results of recent assessments used both cumulative release and dose as alternative performance measures at the accessible environment (Andrew et al., 1994; Andrew et al., 1995; and Wilson et al., 1994). Because the new standard for Yucca Mountain is expected to be dose based the more recent d y s e s have reported dose history and peak dose at the accessible environment (Duguid et al., 19%; and INEEL Task Team Report, Appendix E, 1997). Future assessments such as the TSPA being conducted for the Viability Assessment of Yucca Mountain are expected to follow the current DOE guidance or the EPA standard for Yucca Mountain.

The method used in TSPA-1995 (Andrews et al., 1995) represents the latest complete set of analyses of a potential repository at Yucca Mountain. Since the TSPA-1995 analyses were completed a considerable amount of new information has been gained. The analyses in this report use the same general method as TSPA-1995, but the input data has been updated to account for the new understanding of the site, the repository, and the waste package. The bases for these analyses of DOE SNF are presented in Chapter 2.

The ~ati&d Spent Nuclear Fuel Program has categorized the DOE SNF into 15 categories for performance assessment. The categories of DOE SNF analyzed in this report are as follow;

Category 1 Uranium metal spent fuel Category 2 Uranium-Zirconium alloy spent fuel Category 3 Uranium-Molybdenum alloy spent fuel Category 4 Uranium oxide spent fuel Category 5 Uranium oxide-disrupted clad spent fuel Category 6 Uranium-Aluminum alloy spent fuel Category 7 Uranium silicide spent fuel Category 8 High-integrity Uranium-Thorium carbide spent fuel Category 9 Low-integrity Uranium-Thorium carbide spent fuel Category 10 Uranium and Uranium-Plutonium carbide spent fuel Category 1 1 Mixed oxide spent fuel Category 12 Uranium-Thorium oxide spent fuel Category 13 Uranium-Zinium hydride spent fuel

Category 14 (Sodium bonded spent fuel) and Category 1 S (Navy spent fuel) were not analyzed. This study focused on DOE SNF (Categories 1-1 3) and did not include Navy spent fuel (Category 15). Category 14 spent fuel is expected to be treated prior to disposal, and the characteristics of the

1-3

treated fuel were not avspilable for this assessment More detail on each of the 13 categories of DOE SNF is presented in Appendix A.

1.2 OBJECTIVES AND APPROACH

The objective of this study is to determine the effects on repository performance of disposal of DOE SNF in a potential geologic repository at Yucca Mountain. The repository analyzed is assumed to contain 63,000 MTHM of commercial spent fuel, the DOE SNF, and enough HLW to co-dispose all of the high-enriched and medium-eyiched DOE SW. This performance assessment evaluates the incremental change in dose to an individual at the accessible environment that can be attributed to the disposal of the DOE SNF using the calculational method used for TSPA-1995 with updated site and design assumptions. The performance measure used for these analyses is dose to an individual at the accessible environment based on drinking water.

The dose amibuted to each category of DOE SNF (Categories 1 through 13) is analyzed and the results are compared to an equivalent amount of commercial spent fuel. In addition, the dose attributed to the sum of all of the DOE SNF (the sum of Categories 1 through 13) is compared to an equivalent amount of commercial spent fhl, and its effect on the dose fkom the entire repository is analyzed. The geologic repository for TSPA-1995 was assumed to contain 70,000 metric tons heavy metal (MTHM) which was composed of 63,000 tons of spent fuel and 7,000 MTHM of HLW. For the analyses of DOE SNF the repository is assumed to contain 63,000 PvlTHM of commercial spent fhel, 2436 MTHM of DOE SNF(Categories 1-13), and 8,745 MTHM of HLW. The amount of HLW is based on co-disposal of high-enriched and medium-enriched DOE SNF assuming no more t h a 14.4 kilograms of fissile material in each co-disposal package. The co-disposal concept is for disposal of spent fuel and HLW in the same waste package in order to reduce the likelihood of criticality. This approach yields a repository that contains 74,181 MTHM. Because this exceeds the 70,000 MTHM limit for the first repository, the dose from the entire repository is presented for both a 74,181 MTHM repository and for a 70,000 MTHM repository. For the latter, the effects of the DOE SNE and HLW are scaled to 7,000 MTHM to obtain the results presented in Chapter 4.

1 3 TOTAL SYSTEM PERFORMANCE ASSESSMENT (TSPA)

A total system performance assessment brings together all relevant components of the waste containment and isolation system that affect the concentration of radionuclides that may reach the accessible environment and the corresponding radiation dose associated with that concentration The individual components that are modeled in a TSPA are shown in Figure 1.3-1. Although c l i i t e is shown in Figure 1.3-1, the effects of climate change were not included in this study (see Section 2.3). Each of the ovals in the figure corresponds to a process-level model or to an abstraction of the results of a process model, which is based on direct laboratory or field data that have been synthesized using either an empirical or physical relationship that describes the particular process. The simplified general flow of information between the different model domains is illustrated by the m w s in Figure 1.3-1. In this depiction, the boundary conditions of one domain are provided by the output h m the preceding domain. For example, the repository-scale percolation flux is derived from the results of the site-scale unsaturated-zone flow model. Similarly,

the source term for radionuclide transport in the unsaturated zone is provided by the calculated release h m the engineered m e r system (EBS). A detpiled description of the flow of information through the domains modeled foi analysis of the EBS is shown&~ Figure 1.3-2. The concentration of radionuclides released as a function of time fiom the EBS provides boundary conditions for the geosphere modules [unsaturated-zone (UZ) transport and saturated-zone (SZ) transport].

A performance assessment of a repository typically consihs of several levels of analyses, ranging fiom the detailed representation of individual processes to simplified analyses of the entire waste disposal system. This model hierarchy is shown in Figure 1.3-3. The base of the pyramid corresponds to conceptual models that are used as a basis for development of process models. The process models are detailed phenomenological models of processes acting within the engineered or the geologic components of the system. The subsystem models, which are less detailed than process models, are used to analyze subsystem components, such as the waste package or the engineered baniers. The apex of the pyramid corresponds to the abstracted (simplified) representations of the processes that are used to evaluate the performance of the total system. The total system is modeled using the RIP (Repository Integration Program).

For each detailed procesdconceptual model there exists a comqwnding, perhaps abstracted, version for the purposes of total system performance assessment. The need for abstracted models originates from the complexities inherent in total system assessments due to the coupling between various processes/sub-systems, parameter and model uncertainties, spatial and temporal vadabilities, and multiplicity of designs and future scenarios. In addition, abstraction is necessary because of computer-storage and mmhg time requirements. The use of probabilistic +rfoman~e assessments to evaluate regulatory compliance of complex systems within current computational capabilities also necessitates some degree of simplification within the abstracted models. The word abstraction is used to connote the development of a shplified.dealized model (with appropriately defined inputs) that reproduces/bounds the results of an underlying process model. The inputs for the abstracted model can be a subset of those required for the process model. Alternatively, intermediate results fiom the process model can be analyzed to develop "response functions" which can then be used as inputs to the abstracted model. In either case, it is necessary to demonstrate that predictions of both the process model and the abstracted model are reasonably similar.

The assessment of total system performance is based on results derived fiom process-level models which in turn are based on direct observations or interpretations of laboratory and/or field tests. The flow of information fiom "data" to process-level models (where applicable), to abstractions of . process-level models (where applicable), and finally to the total system performance assessment . itself, is illustrated in Figure I -3-4.

The flow of information for an assessment begins with field and laboratory test data and the interpretation of these data using process-level models to synthesize the data and other infonnation into a cbnsistent repmentation of the relevant processes that affect waste isolation. The relevant processes and infonnation are shown on the left-hand side of Figure 1.3-4, and they feed into the process-level models as indicated by the numbers 1 through 8. The abstractions of the results fiom the process-level models are iri the form of response surfaces, tables, or other functional

relationships that are used in the total system performance model. The information and results of the process-level abstractions are input to the total system performance assessment model. In many cases the information derived h m field and laboratory testing is used directly in the analyses rather than going through the process-level model and abstraction steps. Examples of information derived fiom laboratory experiments include alteration~dissolution rates of the waste form, solubility of individual radionuclides, and radionuclide sorption values. Examples of infoxmation derived directly or indirectly fiom field measurements include hydrogeologic unit thicknesses and rates of infiltration of precipitation. In many other instances, however, predictive models are required to provide results that can become input to the analyses. Examples of these include unsaturated-zone . and saturated-zone flow, drift-scale thermohydrology, and waste package degradation. In these instances, the results fkom the process-level model simulations are used to define the relationship between the "known" parameters, including their corresponding uncertainty and spatial variability, and the required results for use in the TSPA calculations.

In summary, for this analysis, site and design data and site and design information are used in four process level models that describe flow through the unsaturated zone, flow through the saturated zone, drift scale thenno-hydrology, and waste package degradation (Figure 1.3-4). The results fiom these process-level are used to develop abstractions of flux through the unsaturated and saturated zones, flux conditions in the vicinity of the waste package ( i.e., "dripsw), and near-field conditions of temperature, humidity, and liquid saturation (Figure 1.3-4). The near-field conditions are fed back into the process-level model for the waste package to determine waste package failure caused by the development of pits (Figure 13-4). The abstracted results fiom the process models are used in the to determine waste package degradation, radionuclide release h m the e n g i n e barrier system, transport through thi geosphere (unsaturated and saturated zones), and radionuclide concentration in ground water at the accessible environment. The flow of information h m the process and subsystem models into the RIP is shown in Figure 1.3-5. The process models used to develop input information for the RIP, shown in Figure 1.3-5 (TOUGH2, NUFT, WAPDEG, and FEHMN), are discussed in Chapter 2. Radionuclide concentration at the accessible environment is readily transformed into dose to an individual h m drinking water, assuming that the individual consumes two liters per day, using appropriate dose conversion factors @PA, 1988). These conversions are done within the RIP (Figure 1.3-5).

1.4 RIP (REPOSITORY INTEGRATION PROGRAM)

The performance assessment tool used in the analysis of DOE SNF is the computer program RIP (Repository Integration Program) which was the program used for TSPA-1995 in conjunction with detailed process-level models. The RIP was specifically developed by Golder Associates Inc. in order to evaluate the performance of a potential radioactive waste disposal facility at Yucca Mountain (Miiler et al., 1992; Golder Associates, 1993). It is firlly documented in a theory manual and user's guide (Golder Associates, 1994). The program has recently been formally verified. consistent ~ 5 t h ASME-NQA-1 and ISO-9000 standards (Golder Associates, 1995). Version 4.0% of RlP was used for these analyses.

The components of RIP are (1) waste package behavior and radionuclide release component model, (2) radionuclide transport pathways component model, (3) disruptive events model, and (4) biosphere dox/n'sk model. ' "he features and capabilities bf hese component models are summarized briefly below.

The waste package behavior and radionuclide release component model input requirements are descriptions of the radionuclide inventories in the waste packages, a description of near field environmental conditions (which may be defined as temporal and spatial variables), and subjective estimates of high-level parameters describing container failure, matrix alteration/dissolution, and radionuclide mass transfer. Waste package failure rates, along with matrix alteration/dissolution rates, are used to compute the rate at which radionuclides are exposed. Once waste is exposed, RIP computes the rate of mass transfer out of and away fiom the waste package. Parameters describing waste package failure and radionuclide exposure and mass transfer can be described as a function of near-field environmental conditions. The output fiom this component (for each system realization) consists of time histories of release of each radionuclide fiom the waste packages, and acts as the input for the transport pathways component.

The radionuclide trollspot? pathways component model simulates radionuclide transport through the near and far field in a probabilistic mode. The RIP model uses a phenomenological approach that attempts to d&be rather than explain the transport system. The resulting transport algorithm is based on a network of userdefinedparhways. The geosphere and biosphere pathways reflect the major features of the hydrologic system and the biosphere, and are conduits through which transport occurs. Q e pathways may be used for both flow balance and radionuclide transport purposes, and may account fm either gas or liquid phase transport. The purpose of a pathway is to repkent large- scale heterogeneity of the hydrologic system, such as geologic structures and formation-scale hydrostratigraphy.

Geosphere pathways may be subdivided intoflow modes, which address heterogeneity at the local scale (e.g., flow in rock matrix, flow in fractures). The flow modes are primarily distinguished fiom one mother on the basis of flow velocity in the mode, although retardation parameters may also differ between flow modes.

The transport of radionuclides along a geosphere pathway is based on a breakthrough curve, which is calculated & a cumulative probability distribution for radionuclide travel times along the pathway. The breakthrough curve combines the effects of all flow modes and retardation on the radionuclide travel time, and determines the expected proportion of mass that has traversed the pathway by any specified time. The breakthrough curve is computed, based on a Markov process algorithm for exchange between different flow modes.

The third component model in RIP represents disruptive events. Disruptive events are defined as discrete occurrences that have some quantifiable effect on the processes described by the other two component models. Examples of disruptive events include volcanism, faulting, and humaa intrusion. The user first identifies all significant events (i.e., events that are both credible and consequential). Having done so, each event is assigned a rate of occurrence, and if desired, one or

more descriptor parameters, which define the characteristics and magnitude of the event (e.g., length of a volcanic dike). Descriptor parameters may be described stochastically. Event occurrences are simulated as Poisson processes.

The fourth component model describes the fate and effect of radionuclides in the biosphere. The biosphere dosdrisk model allows the user a define dose receptors in the system. Receptors receive radiation doses from specified geosphere (e.g., a water supply aquifer) or biosphere (e.g., a pond, or flora and fauna) pathways. Concentrations in these pathways arc converted to radiation doses (or cancer risks) based on userdefined conversion factors. For the analyses in this study of DOE SNF only the dose h m the drinking water pathway was considered.

The RIP model has the capability of tracking the effefis caused by specified types and quantities of the total inventory (e.g., the commercial spent fuel the HLW, or the DOE SW). This feature was used to either calculate the dose fiom the total waste inventory, or to calculate 'the dose attributed to a specified component of the inventory, such as a specific category of DOE SNF.

Because the RIP software is not currently being operated under the control of the DOE Quality Assurance Program the data used in this pedonnance assessment are not being held to quality standards necessary for inclusion in the License Application for the potential repository.

Figure 1-1

Nye County

Location Map of the Potential Repository Site at Yucca Mountain, Nevada.

)Dry Desert Environment / Unsaturated Overburden A

/ / pRepooitory Host ~ o b k t---- //,'/E

7 Saturated Rocks

Figure 1-2. Schematic Diagram of Components that Contribute to Waste Isolation at Yucca Mountain.

Figure 1-3

Unsatunted Zone Percolation Flux

Repository Plan View

Unsaturated- Zone

Transport

Satumted-

tone

Measure Cumulative Release (Ci)

\

and Peak Dose (rem/yr)

at

Accassible Environment

( 5 k m )

Schematic Diagram of the Ground-Water Tramport Pathway born the Potential Repository to Humans.

. .

. .

Figure 1.3-1 Schematic Diagram of the Models Used in a.Total System Performance Assessment (TSPA).

q(t) 1 CO Associated RIP Cells a Rl External Process Model

Parameters as fftl q(t) conc of RN(i) C(t) chemical parameters Mi(t) mass flux q(t) percolation flux *

RH(t) relative humidity S,(t) liquid saturation T(t) temperature

Figure 1.3-2 Flow of Information and Results through the Domains Modeled for the Analyses of a Potential Repository at Yucca Mountain.

Figure 1.3-3 Hierarchy of Models Used in Performance Assessment.

~h~sical-chemical Processes ___)c Output

I=+ pii5-F I+

EBS D e s i i

Figure 1.3-4 Diagram of Iaformation Flow to the Total System Performance Assessment Model. (See Figure 1.3-2 for identification of p'ararneters transferred among models.)

Dose at Accessible

Environment

Figure 1.3-5 Flow of Information from Process and Subsystem Models into the Repository Integration Program (RIP).

Mountain-Scale Therrnohydrology

(NU FT) --+

Unsaturated-Zone Hydrology (TOU GH2)

Drift-Scale Repository

Therrnohydrology --+ Integration

(NUFT) Program

(RIP)

L

A

v

Waste Package Degradation @VAPDEG)

Geosphere Transport (FEHMN)

A

wmgCW3.ccir

Saturated-Zone Hydrology (TOUGH2)

2. PERFORMANCE ASSESSMENT OF DOE SPENT NUCLEAR FUEL (SNF)

The following assessment uses an approach similar to TSPA-1995 with model parameters updated to the current understanding of a repository at Yucca Mountain. It also differs from TSPA-1995, in the degree of coupling among the models used in the analyses. For example, in the DOE SNF analyses the flux through the unsaturated zone was taken from the calibrated three-dimensional site- scale model (TOUGW), and was used in the hydrothermal model (NUFT) to calculate parameters

'

that govern the waste package failure rate. The flux was also abstracted for use in the unsaturated zime transport model in RP. Because of the degree of coupling among the models it would not be appropriate to arbitrarily vary a parameter, such as percolation rate independent of the other related models. The degree of coupling among models yields more realistic results, but has the drawback of requiring additional process model analyses for key model parameters such as transient flux caused by climate change.

For the analyses that foIlow in this TSPA of DOE SNF two assumptiom are made that are more conservative than in other recent assessments. First, the amount of dripping water that contacts the waste was assumed to be greater (3WA of the packages are at locations where drips occur). Second, the DOE SNF was assumed to be placed in waste package failure Group 1 rather than being distributed among the eight waste package fiilure groups. Therefore, the results of these analyses should be viewed in a relative sense (i.e., used for comparison of one calculational scenario to another), and should not be viewed as absolute predictions.

The major components (domains) considered in a TSPA were discussed in Chapter 1 and are presented in Figure 1.3-1. The domains considered in this TSPA for DOE SNF include: the near- field environment for waste package and waste form degradation, radionuclide release, and

. engineered barrier transport; a thennohydrologic regime; ground-water flow and radionuclide transport for ttre msatm&d and saturated zones; and biosphere (Figwe 1.3-1). In this study, as in TSPA-1995, a simple biosphere model is used that considers the dose to an individual from drinking water. The near-field geochemical environment shown in Figure 1.3-1 was not considered in this performance assessment. For example, interaction among HLW glass and spent fuel in the same waste package or processes that cause mineralogical changes in the near field were not considered.

2 REGULATORY CONSIDERATIONS

The analyses in this report focused on only one measure of total system performance, the maximum radiation dose to an individual drinking g r o d water derived fiom a well drilled into the tuff aquifer at the accessible environment boundary. For consistency, the definition of the "accessible environment" is assumed to correspond to a location in the saturated zone, five kilometers down gradient h m the edge of the potential repository. It should be noted that the definition of the accessiMe environment could change as new environmental standards are promulgated. In addition, it is important to point out that the potentially exposed population defined by the "critical group1'

recommended by the NAS report (NAS, 1995) is based on existing land and water use patterns in the vicinity of the repository site. Currently, the closest members of the public, in the direction of ground-water flow from the site, are located near Amargosa Valley approximately 20 kilometers fiom the potential repository: Individuals located down gradient h m the site rely on water

I primarily derived b m tha alluvial aquifer which is down gradient h m the tuE aquifer (where dose is analyzed at the accessible environment). The DOE interim guidance requires analyses of dose from all sources at 20 kilometers downjpdient from the potential repository. However, as discussed earlier the analyses in this study are based on dose to an individual fiom drinking water at the accessible environment. The analyses that follow in this report should be construed to

i represent the dose associated with the average member of a "critical" population The NAS i I committee has recommended that the definition of the "critical" population be developed as part of i the EPA rulemaking process in promulgating the new environmental standards. I i

j - For the analysis of dose as a measure of total system performance, calculations are made over a , range of times following closure of the repository. The dose history plots in this report were 3

i 1

extended to 100,000 years in order to capture the peak dose, and a few plots were extended to 1,000,000 years. This is in contrast to the analysis time period required by 40 CFR Part 191 (EPA, 1985), which is for 10,000 years after waste emplacement A 1,000,000 year calculational time period was recommended by the NAS, that was based on the 1,000,000-year time period of earlier pedormance assessment calculations (Wilson et al., 1994; Andrews et al., 1994). These assessments showed that the peak dose occurred several hundreds of thousands of years following waste emplacement, and was caused by Neptunium. However, when the solubility of Neptunium is reduced by two orders of magnitude (Sassani and Siegmann, 1997) the peak dose occurs much earlier (at proximately 20,000 years), and is caused by a combination of Technetium add 1odine. A discussion of the reinterpretation of data that led to the modification of the solubility limits used in the assessment is presented in Section 2.5.

The likelihood of the new EPA standard being promulgated in terms of dose makes dose a logical choice as the. performance measure for this study. If the new EPA standard requires dose calculations at some greater distance h m the repository than the accessible environment (i-e., Amargosa Vdey), the results of this study could be converted to dose at the selected location using

I a dilution factor (e.g., one that accounts for the dispersion and mixing that occurs between the I I accessible environment and the selected location). The dilution factor could be determined by

conducting additional saturated-zone flow and transport modeling. Dose to an individual fiom all sources couId be calculated by application of a second conversion factor derived from an appropriate dose model.

The NRC subsystem requirements (i.e., the release fiom the engineered banier system) were not evaluated in this study because they may change. This change is likely to occur when NRC revises 10 CFR 60 after the EPA standard for Yucca Mountain Plas been promulgated.

23 SITE DESCR-ON

Yucca Mountain is located in the Southern Great Basin, about 120 km northwest of Las Vegas,

Nevada (Figure 1-1). The Oreat Basin is characterized topographically by north-trending mountain ranges separated by alluvium-filled valleys. Strucurally, Yucca Mountain is a complex of north- to northwest-trending fault-d&&ted ridges. The potential lepbiit6ry is to be constructed within Yucca Crest which is bounded to the west by the Solitario Canyon Fault and to the east by the imbricate fault zone and is transected by the Ghost Dance Fault (Figure.2.3-1).

Hydrologically, the Great Basin can be characterized as an arid to semi-arid region. Precipitation in the vicinity of Yucca Mountain is approximately 170 mm/yr, and the estimated potential evapotranspiration is about 1000 rnm/yr. Consequently, most of the precipitation is returned to the . atmosphere and only a small residual remains to infiltrate into the unsaturated zone. Net infiltration is believed to be extremely variable over Yucca Mountain due to variations in soil cover, topographic controls, and vegetation patterns (Fiint and Fiint, 1994).

Stratigraphically, the unsaturated zone beneath Yucca Crest consists of a layered sequence of tuff deposited from volcanic eruptions which occurred about 10 million years ago. The tuff units range fiom porous, nonwelded ash-flow, ash-fall and reworkdbedded tuff deposits to massive, welded ash-flow and ash-fall rocks (Figpre 2.3-2). The four major hydrogeologic units from the surface to the water table consist of the following:

a va Canvon welded (TCw) unit: consisting of moderately-welded to densely-welded tuff characterized by low matrix porosity, low matrix saturated hydraulic conductivity, and high fracture density.

paintbrush nonwelded (PTn) unit: consisting of partially-welded to nonwelded tuff characterized by high matrix porosity, high matrix saturated hydraulic conductivity, and low fracture density.

Tomaah Sming welded (TSw) unit: consisting of welded tuff characterized by low matrix porosity, low matrix saturated hydraulic conductivity, and high fracture density. The basal vitrophyre of the Topopah Spring member (TSv) is generally identified as a subunit because of its lower porosity compared to TSw. Portions of the lower Topopah Spring member are vitrified, and zeolitic alteration appears in the lower part.

Calico Hills nonwelded (CHn) unit: consisting of moderately-welded to nonwelded tuff of the Topopah Spring member underlying the basal vitrophyre and other partially-welded to nonwelded tuff located below the Calico Hills formation (i.e., Prow Pass, Bullfrog and Tram members of the Crater Flat Unit). The tufkeous beds of the Calico Hills contain both vitric and zeolitic beds leading to a further division of this unit into a vitric upper part (CHnv) and zeolitic lower part (CHnz). The hcture density is similar in both zones, and the porosity of the vitric tuff is marginally higher than that of the zeolitic tuff. However, the matrix saanatea hydraulic conductivity of the CHnv is roughly hvo orders of magnitude higher than . that of the CHnz.

The ground-water-flow regime through the partially-saturated tuffaceous rocks at Yucca Mountain

is controlled by the hydrologic characteristics, including the heterogeneity and spatial variability, of the hydrostratigraphic units. Because of the large disparity in capillary suction between fracture and matrix, pore water in the unsaturated tone is bound mostly in the matrix. However recent site characterization data indicates that higher infiltration of precipitation may occur, and that during intense storm events fracture llow may occur that moves water deep into the mountain in a relatively short time. The average annual precipitation at Yucca Mountain may only produce a small fraction of net infiltration. Infiltration appears to occur during a few intense storm events which may not occur every year. Surface runoff is infrequent and of short duration, ' a d no perennial streams exist in the area. Water infiltrates principally into the Tiva Canyon welded (TCw) unit, but also into the Paintbrush nonwelded (PTn) unit, and Topopah Spring welded (TSw) unit where they are exposed

i

at the land surface. The spatial distribution of infiltration is higher at the northern end of the Mountain and along the Yucca Mountain ridge (Flint, Hevesi, and Flint, 1996).

i i The description of the unsattrrated-zone flow dynamics in the site-scale model is based on the 3

1 conceptual hydrologic flow model of Montazer and Wilson (1 984). Ehstward lateral flow may occur within the PTn unit and above its upper contact. The lateral flow is intercepted by stiuctural features that transmit most of the infiltrated water vertically to the water table. Percolation through the matrix occurs principally vertically in the welded units and both laterally and vertically in the nonwelded units. Fracture flow is predominant in the TCw unit during intense pulses of infiltration and is insignificant in the TSw unit except near the upper contact and near structural features. The site-scale model of the unsaturated zone was calibrated using borehole saturation and moisture tension data, borehole temperatme data, and borehole pneumatic data (Bodvarsson et al., 1996). The spatial distribution of flux at the repository horizon and below the repository from the calibrated site-scale model was used as the basis for the msatwated zone flow in this TSPA of DOE SNF.

The ground-water flow in the saturated zone at Yucca Mountain is generally to the south southeast as indicated by the schematic regional ground-water flow system shown in Figure 2.3-3. A schematic cross section of the flow system between Yucca Mountain, Nevada and Eagle Mountain, California is presented in Figure 2.34. The approximate location of the cross section is shown in Figure 2.3-3. The flow system can be generalized to consist of recharge at the higher elevations just north of Yucca Mountain, flow through the hydrogeologic units to the south and southeast, and discharge through both evapotranspiration at Franklin Lake Playa and flow into other ground-water sub-basins (i.e., Death Valley). The average saturated-zone flux used in the analyses of DOE SM: is 0.31 mlyr and the porosity of the saturated-zone is assumed to be 0.2 (M&O, 1996). This is compared to a flux of 2.0 m/yr in the saturated zone that was used in TSPA-1995.

Given the long time frames of potential interest in total system performance, it is expected that the atmospheric conditions will change with a resulting change in climate, especially precipitation and net evapotranspiration. Therefore, the potential effects of climate change will be an important consideration. Although a range of estimates exist for the possible changes in precipitation in the Yucca Mountain region over the next 10,000 years, process model results do not cmently exist for the potential effects of precipitation changes on (1) percolation flux in the unsaturated zone or on (2) the elevation of the water table and advective flux in the saturated zone. It is reasonable to postulate that increased precipitation would result in an increase in percolation flux and a rise in the

water table, although the degree of correlation and the time lag between changes in surface processes and the subsurface eff- are uncertain. However, the siteds,de flow models, for both the unsaturated zone and the s a k t e d zone, have not yet been calibr&d for the potential change in infiltration. This perfonnan2e assessment does not addms climate change because of this, and because of the mount of coupling among the models used in the analyses.

2.4 REPOSITORY DESCRIPTION

A conceptual design of the potential repository at Yucca Mountain has been described in the Site Chamtakation Plan (DOE, 1988) and has been revised to take into account the possibility of alternative areaf mass loads as well as the decision to use a tunnel boring machine for the excavation of the emplacement drifts (M&O, 1994). Two alternative areal mass-load ranges have been proposed for the potential repository, a "low" thermal load of between 20 and 40 metric tons of uranium (MTHM) per acre and a "high" thermal load of between 80 and 100 MTHMIacre. Two areal mass loads were investigated in TSPA-1995,25 MTHM/acre and 83 MTHM/acre. For the d y s e s of DOE SNF, the 83 -acre thermal loading is used because the current design is for a hot repository. For this themd loading the waste can be disposed in the upper repository block as shown in Figure 2.4-1. The waste containers are disposed in the center of the disposal drift, and for the andyses that follow, the drifts are not backfilled (Figure 2.4-2). A discussion of the waste packages used in this study is presented in Section 3.3. The invert upon which the waste packages rest is assumed to be crushed tuff rather than that of the current design which is concrete. Diffusion properties of the invert were updated based upon studies by Conca and Wright (1992). In particular, the characteristic curves for crushed TSw2 have been used for the invert properties and yields a coefficient of diffusion of 6.94 x 1 O4 m2/yr.

In the current design concept of waste disposal containers for the potential repository at Yucca Mountain, two or three layers of different metals, depending on t h d load, have been proposed for the containment of spent nuclear fuel and vitrified high-level waste (HLW). For spent fuel in the high thermal had case, a corrosion-allowance material such as mild steel has been proposed as the outer containment barrier, and a corrosion-resistant material such as Inconel 625 (Alloy 625) has been proposed as the inner containment barrier. In the DOE SNF analyses all waste containers for spent fuel and high-level waste are assumed to have the same design, viz., a corrosion-resistant inner barrier of Alloy 625 and a corrosion-allowance outer barrier of carbon steel.

For a typical large waste package, containing 21 pressurized-water-reactor (PWR), or 40 boiling- . --reactor (BWR) fuel assemblies, the dimensions of the waste container are about 5.7 m long and . about 1.8 m in diameter. The thickness of the inner barrier for both the large spent fuel package and the HLW package is 20 mm; the thickness of the outer barrier for the spent fuel package is 100 mm, and for the HLW waste package is 50 mm. The HLW packages and the co-disposal packages are assumed to be of two different lengths b ~ccommodate the size of the DOE SNF canisters (3.1 m and 4.6 m, Appendix A).

2.5 NEAR FIELD ENVIRONMENT

Figure 2.5-1 indicates the major engineered barrier processes that occur in the near field leading to radionuclide release. The thennohydrologic drift-scale analyses conducted using NUFT provided waste-package d a c e temperature (T) and relative humidity (RH) which were used in the waste- package degradation modeling, and liquid saturation (S) of the gravel invert which was used in calculation of the diffusion coefficient for difkive release of radionuclides. The waste-package degradation modeling results provided the time to first pit penetration of the waste containery and subsequent degradation or pitting of the waste container (see Section 2.6). The waste form in the nominal case of TSPA-1995 and for the d y s e s of DOE spent fuel was assumed to be exposed upon first pit penetration, due to immediate cladding Mure. No cladding failure modeling was performed in TSPA-1995 (Andrews et al., 1995), and no credit is taken for cladding of the DOE SNF.

In TSPA-1995, two mechanisms for initiation of corrosion of the waste package were considered. One was based on relative humidity and the other was based on a combination of relative humidity and temperatme in the vicinity of the waste package. The latter is used in the analyses of DOE SNF in this report. In addition to consideration of pit formation, galvanic protection that is provided by the materials that make up the waste package (cornsion-allowance and corrosion-resistant materials) was shown to extend the lifetime of the package by thousands of years. The analyses where galvanic protection was considered in TSPA-1995 indicated that a significant number of the waste packages had not failed in the first million years. In these analyses of DOE SNF galvanic protection was assumed for 90 percent of the waste packages.

The near-field environmental conditions S e c t such processes as the waste-form dissolution, the solubility of the radionuclides in the aqueous phase in contact with the waste form, and the magnitude of both the advective and diffusive components of transport h m the waste-form d a c e through the degraded waste package and the indrift materials (the invert and backfill, if present) into the host rock. Waste-form dissolution rates for oxide fuels and HLW glass used in TSPA-1995 and in this study were derived fiom empirical fits to data obtained from laboratory experiments under a range of environmd conditions. The other dissolution models (for metallic, ceramic, and carbide kels), while judged to be appropriate, have not been subjected to this degree of comparison with test results because data are not available. The advective flux (drip rate) was taken to be equal to the average flux in the h t u r e s in the Topopah Spring. The diffusive flux component of radionuclide transport was derived from the hydrologic conditions in the drift materials as calculated by thermohydrologic modeling. These flow conditions led to the develop~pent of alternative conceptual models for release . of radionuclides based on the percolation flux and the percentage of waste packages that are in . l ~ o n s where fixture flow occurs (i.e., the amount of advective and/or diffusive release fiom the engineered barrier system).

The geochemical environment inside the waste package is assumed to be analogous to the ambient aqueous' geochemistry inferred to exist in the Topopah Spring Tuff host rock. This neglects any geochemical perturbation likely to occur as a result of the varying thermohydrologic regime, as well as the interaction of the pore fluids with the materids placed in the drift (mot the least of which is the thick iron-based corrosion-allowbce outer layer of the waste 'package). These simplifling

+ . assumptions were made in TS'P~-l995 because a detailed geochkmikl model of the waste package degradation over time was not available. The assumptions are thought to be consemtive since the presence of the iron in the w a h package would increase sorptidn o'f many of the radionuclides and delay their release. The geochemical effects of the DOE SNF and HLW on near-field model parameters have not been analyzed. These analyses should be done in the future to ensure that the near-field parameters used in evaluation of repository performance are conservative.

Radionuclide solubilities were derived h m empirical curve fitting to data obtained fiom laboratory experiments for a range of thermal and chemical conditions. Uncertainty in the fitting parameters reflects the uncertainty in the experimental data only and does not account for uncertainty in the stability of the controlling phase (i.e., conceptual uncertainty) or the variability associated with a range of local geochemical conditions. Radionuelide solubilities used for the analyses in this report are the same as those used in TSPA-1995 (Andrews et al., 1995) with the exception of the solubility ofNeptunium. The.solubility of Neptunium, based on d y s i s and reinteqxetation of existing data, has been decreased by two orders of magnitude.

Between 1991 and 1993 performance assessment analyses, incorporated new data on Neptunium that d t e d in revisions to both dissolution rates and solubility limits. These changes resulted in higher calculated Neptunium releases. The data upon which the revised solubility constraints were based in all likelihood represent m m b l e equilibrium between the aqueous fluids and metastable Neptunium phases (i.e., Na-neptunyl carbonate hydrates and NhO,, Nitsche et al., 1993 and 1994). At the relevant conditions the calculated stable Neptunium phase (Np02) corresponds to Neptunium concentrations that are many orders of magnitude lower than these metastable phases (Sassani and Siegmann, 1997). These investigators believe that there may be objections to using the calculated NpO, soIub0ities to constrain the solubility-limited Neptunium concentration may on grounds that kinetic baniers could prevent the formation of the NpO, phase. In their analysis Sassani and Siegrnann evaluated the applicability of the various phase constraints on aqueous Neptunium concentration through synthesis of data from the dissolution-rate studies, the solubility studies, and thermochemical calculations.

- - The highest value derived h m the spent-fuel dissolution studies was found to be only about three percent of the minimum value used as the basis for TSPA-1995. In addition, the highest value used (1.4 x 10" M) is an average. It includes the measured values h m the first test cycle that are much higher and values decreased in later test cycles to 9 x 10"' M. This value is about two orders of magnitude smaller than the average. This indicates that it is unlikely that the dissolution of spent he1 into water from the 5-13 well would produce concentrations of Neptunium that are high enough to approach satmation of the metastable phases upon which the TSPA-1995 solubility limits are based. The 3-1 3 well is located approximately eight kilometers down gradient fiom Yucca mountain, and water fiom it is characteristic of satmted-zone water along the tramport pathway from the repository to the accessible environment. Calculated Neptunium concentrations using NpO, solubility in 1-1 3 water were found to be even lower than the value derived from dissolution studies. The results of thermochemical calculkions indicated that, for water similar to that fiom well 1-13, NpO, solubility

'

is a more appropriate constraint because concentrations did not reach levels that are high enough to form the metastable phases observed by Nitsche et al. (1993, 1994). In addition, Sassani and

Siegmann found that the results of long-term (steady-state) dissolution studies are consistent with concentrations set by considering a NpO, solid solution in the UO ,fuel matrix. Based on these analyses, the conservative recommendation was made to lower the expected value for the Neptunium solubility by two orders of magnitude (Sassani and Siegmann, 1997).

Liquid fluxes and velocities (in both fractures and matrix) in the unsaturated-zone hydrogeologic units below the repository are based upon the most recently calibrated site-scale flow model. The estimated overall average percolation flux (average of fkctwes and matrix) at the repository horizon is 6.2 mdyr. This value is about three times that of the highest percolation flux (unaffected by climate change) that was considered in TSPA-1995. The updated site-scale flow model is based upon an isothermal ground-water flow simulation using a three-dimensional dual permeability model with the most recent hydrogeologic parameters obtained fbm inverse modeling using the model ITOUGH. For this calculation the range of the measured laboratory hydrology data was used to fit the field measured surface infiltration rate and the observed saturation profile data (Bodvarsson et al., 1996). The simulated liquid saturations and fluxes in the rock matrix and fractures were extracted for six columns within the footprint of the repository that have a range of flux h m about 4.0 to 10.0 d y r . The saturations and fluxes, together with geometric data, were fbrther processed to obtain liquid saturations and fluxes and layer thicknesses for all the layers within each of the six columns from the potential repository horizon to the water table. These processed data, together with the rock matrix permeabilities and porosities at the potential repository horizon, were used in the analyses of DOE SNF.

Updated drift-scale thermohydrologic calculations were performed that were based on the new calibrated-proPerty values and the unsaturated-zone site-scale flow model. This provides updated values for relative humidity and temperature at the waste-package swfke and liquid saturation and temperatme in the invert. These ternperatme% relative humidities (RH), and liquid saturations were calculated using the computer code NUFT @onisothermal wturated-Saturated Elow and ~mnsport) w&O, 1997). The NUFT code (PJitao, 1996) simulates the coupled transport of water, vapor, air,-and -heat in fractured porous media For the drift-scale analyses the k tu red tuff surrounding the repository drift was conceptualized as an equivalent continuum.

A two-dimensional grid was used for the thermohydrological simulations that reflects the center-indrift waste-package emplacement scenario described in the advanced conceptual design (M$O, 1996a). Symmetrical repository drift and waste package spacing was assumed, thus the simulations reflect the conditions of an average waste package that is relatively distant fiom the repository edge. The 83 -acre repository has a drift spacing of 22.5 m and an in-drift package spacing of 15.4 m. The t h d Ioad is assumed to be fiom fuel with an average age of 26 years. The boiling water reactor &el occupies 40% of the repository and pressurized water reactor fuel occupies the other 60%. The high-level waste is assumed to have a negligible impact on the t h d load (M&O, 1995).

2.6 WASTE P A W G E DEGRADATION

Given that the waste packages must "f%ll' (i.e., be breached to an extent that the mobile water present in the near-field enviro~lent can ingress into the package and dissolve radionuclides that can be

transported out of the package) before dissolution and transport of the waste form can occur, an important first -step in evaluation of system perfonngmce is the prediction of waste package degradation rate. The degradhbn rate of the waste package is dipendent on (1) the waste package design [in particular, the d a l ( s ) used in the waste package fabrication and the thickness of these matgial(s)], (2) the repository design (in particular, the thennal load, the presence and nature of baclc6ll (if any), and the size of the emplacement drifts), (3) the near-field thennohydrologic regime in the drifts adjacent to the waste package d a c e (in particular, the temperature and relative humidity), and (4) the degradation characteristics of the waste package materials'(inc1uding the criteria for corrosion initiation and the rate of corrosion as a function of the near-field thermohydrologic and chemical environment). Information on each of these topics is required as input to the waste package model to predict the time-rate of "failure" of the waste packages.

The waste package degradation model WAPDEG (Lee et al., 1996 and 1996a) was used to describe the failure history of waste packages & the repository environment. The new waste-package degradation simulations were performed based on the updated drift-scale thermohydrologic simulations. Since TSPA-1995, two major improvements have been made to the waste-package degradation model, WAPDEG. The corrosion time concept is used to incorporate the dependency of the corrosion rate on the current corrosion depth. This approach is more accurate than the approach used in TSPA-1995, in which the duration of corrosion was treated directly as a proxy for the corrosion depth. The "patches" approach has also been developed since TSPA-1995 to give a more accurate representation of spatial wrkhdity on individual waste packages. This approach gives a strong correlation between the general comsion depths at neighboring locations of a specific waste

i

package @f&O, 1997).

In the repository en-ent, depending on the exposure conditions (temperature, relative humidity, dripping water, chemistry of dripping water, mineral deposits, microbiologidly influenced corrosion, rockfall, and integraty of welds), a waste package could undergo degradation by several different corrosion modes (humid-air corrosion, aqueous corrosion, general corrosion, localized pitting and/or- crevice corrosion, and stress corrosion cracking), and these modes could occur simultaneously It is likely that diffecent corrosion modes would be operative at different locations on a waste package (i.e., top, bottom, and sides), and that the corrosion rate for a given corrosion mode would be different depending on the location on the waste package. The patches model takes different locations ("patches") into consideration. The model divides the waste package surfhce into "patches" and assignes different corrosion modes and rates to each patch in order to model the waste package corrosion/degmdation explicitly for different corrosion modes and different locations on the wage package.

With the "patchesw approach, corrosion modes and their rates c8n be correlated among neighboring patches. For example, patches surrounding a given patch are likely to undergo degradation by the same corrosion mode as the patch in the center. The surrounding patches would have somewhat different corrosion r e than the central patch, but the difference would not be expected to be large unless the corrosion created defects or the patch was under abnonnal mechanical stresses. A patch that is located several patches away fiom the central patch could undergo corrosion at substantially different rates, or could experienck corrosion of a different corrosion mode. In the current waste

package simulations, corrosion rates and modes for neighboring patches are not directly correleated, and are sampled randomly.

The assumed degree of galvanic protection has been modified since TSPA-1995, but is based upon the same type of model. In particular, a threshold is specified to determine the duration of galvanic protection. This threshold is a function of the mass of the outer barrier that has undergone general corrosion. For the base case, it is assumed that 90% of the 'packages have 75% galvanic protection, while the remaining 10% of the packages have no galvanic protection. The galvanic protection . '

model assumed is based on an expert's assessment (McCright,' 1995) and has not yet been substantiated. Experiments are cmently underway at Lawrence Livermore National Laboratory where galvanically coupled samples are Wig tested to develop data and information for galvanic processes such as, galvanic protection threshold, gdvauic corrosion of the outer barrier (carbon steel), and galvanic protection of the inner barrier (Alloy 625). In addition, new information on galvanic protection ofthe h e r barrier was developed from the recent expert elicitation on waste package degradation (M&O, 1997a).

Based on these assumptions and the updated thermai history for a waste loading of 83 MTHM / acre, WAPDEG was used to calculate the *tion of waste packages that failed over time (M&O, 1997). Failure is considered to occur when the outer barrier develops the first through-going pit (Figure 2.6- 1). However, the release fiom a Med package over h e depends on the number of pits through the package. The first package failures begin to occur after about 600 years, and about 85% of the packages have failed in the first 1,000,000 years (Figure 2.6-1).

Waste failure groups were developed based on the approach used in TSPA-1995. In this approach the degradation of 400 representative waste packages was analyzed using WAPDEG. Eight groups were selected to represent the heterogeneity in the waste package failure, and the pitting history of each group was abstracted for use in the RIP. For each of the eight waste package failure groups an average time to first pit penetration and the pitting profile was developed (Figure 2.6-2). Group one contains the packages that develop the first pit between 600 and about 10,000 years, and the average number of pits through the package is shown by the upper curve in Figure 2.6-2. The dip in the failure curve.is an d a c t of the modeling. It is caused by the temporal averaging of the number of pits on all of the waste packages with the number of packages that are failing. At some point in time, the temporal average is lower than the number of packages that are failing. Figm 2.6- 2 shows the failure history of packages in the eight failure groups, and it can be seen that the dip also appears in failure group 2 (the second highest curve on Figure 2.6-2).

The first packages to fhil are those in waste package failm Group 1, and are those with no galvanic protection. For the analyses of DOE spent h l s in Chapter 4, the packages are assumed to be placed in Waste Package Group 1. This was done because of the difficulty of distributing the waste packages in each spent fuel category evenly among the eight waste package failure groups [i.e., for Category 4 DOE SNF-there are seven waste package types (Table 3.3-1) and distribution of these package types among eight failure groups yields 56 failure groups that would have to be tracked in the W, which is not possible in the version of RIP used for these analyses]. This assumption is conservative because it yields peak doses at the accessible environment that are about an order of

I / ' <,-. : magnitude higher than if the &packages were distributed evehly among the eight waste package failure groups (see Section 5.7). The assumption that hat of the waste packages are in waste package failure Group 1 essentially eliminates the effects of galvanic protection because it contains all of the waste packages that fail early (ir., the 10% of the waste packages that do not have galvanic . protection).

2.7 RADIONUCLIDE TRANSPORT PROCESSES

2.7.1 Engineered Bamer Release Conceptual Models

T h e conceptual models of engineered banier system release were evaluated in TSPA-1995 (Figure 2.7-1). First, for the conceptual model of advective and diffusive release from both the waste package and engineered barriers (the "dripssn-waste-form model), after a waste package has "failed" (i.e., the initial pit has penetrated the inner cornsion-resistant layer), it is assumed that the near-field conditions (i.e., the temperature, humidity, liquid saturation, and the presence of drips) occurring outside of the package are immediately transferred to the inside of the waste package. These environmental conditions, combined with information on the behavior of the waste form and other engineered barriers under these environmental conditions, are required in the prediction af radionuclide releases from the engineered barriers to the host rock. In this model, advective release occurs at a rate proportional to the flow of dripping water in the drift, and diffusive release occurs at a rate proportional to the number of pits penetrating the waste container [Figure 2.7-1 (a)].

A second qnceptual model of release (the "dripson-waste-package'' model) is presumed to be more realistic than the first model, and takes more credit for a partially intact waste container [Figure 2.7-1 @)I. This model assumes only di-ive releases through the waste container, because of corrosion products filling the corrosion pits and blocking advective flow into the waste container. Near-field environmental conditions (except for dripping flow) were assumed present inside the waste container immediately after the first pit. The model still assumes both advective and diffusive release from the engineered W e r system (e.g., where drips occur, the release through the invert is by advection, and where drips do-not occur, release through these components is by diffusion).

A third conceptual model of release was developed to evaluate the potential benefits associated with the emplacement of a so-called Richards'.or capillary barrier, in which the backfill is designed to conduct any advective flux (i.e., drips) away h m the waste package and underlying invert materials due to the capillary-pressure differences across unconsolidated materials of different grain size. Only diffusive releases &om both the waste package and the barriers were allowed to occur in this model at all package locations [Figure 2.7-1 (c) 1.

For the analyses of the performance of DOE SNF the conceptual model of "drips on the waste package" was assumed for 30% of the waste packages that are located where dripping water occurs. For the other 70% of Be waste packages, only diffusive release occurred through the engineered barrier system.

For the conceptual model "drips on the waste package" diffusive transport of radionuclides through

perforations in breached waste package is assumed. This assumption is based on limited data and observations of potential "plugging" of the perforations by corrosion products and mineral precipitates (Jones and Wilde, 1987; Massari, 1996; and Raman and Nasrazadani, 1990). The movement of water through perforations andlor cracks that are filled with corrosion products and mineral precipitates is expected to be slow because of the presence of the porous corrosion products that are cemented by the precipitated mineials (Schwertmann and Taylor, 1989). Under extremely slow advective water flux the tramport can be approximated by diffusion Experimental studies & required to substantiate andor validate this conceptual model and the underlying assumptions. Because of the uncertainty in the assumption of diffusion h m the failed waste package the . conceptual model "drips on the waste fonn" is assumed for some of the sensitivity analyses in Chapter 5.

Once the waste package has failed, other baniers such as fuel cladding are subject to corrosion and failure. For the analyses presented in this report no credit is taken for the fuel cladding (i.e., it is assumed to fail instantaneously with the waste package reteasing any radionuclides contained in the. gap between the fuel and the cladding if a gap exists). h e to the method of manufacture of some of the DOE spent fuels a gap does not exist.

The dissolution of the waste form is affected by temperature, pH, and concentration of dissolved constituents in the water. The radionuclides are released h m the waste form based on its alteration rate, and are transported &om the package at a rate dependant on whether. they are alteration controlled or solubility limited mdionuclides. The alteration controlled radionuclides are '%e, '%b, v c , 12$ -and 13'Cs, and the remaining radionuclides in the inventory are solubility limited. The alteration controlled radionuclides arc highly soluble and can be tramported fiom the waste package in a small amount of water at the rate that they become available for transport, the alteration rate. The solubility limited radionuclides are dissolved into the volume of water contacting the waste according to their solubility and are removed fiom the waste package by either diffusion or advection. The difference between the two mechanisms for release has a strong influence on dose history at the accessible environment as the quantity of spent b l is reduced. As will be seen later (Chapter 4) the alteration controlled radionuclides may be scaled directly using the ratio of reduction in the MTHM of the waste form [i.e., '/z if the quantity of spent fixel (MTHM) is reduced by a factor of two]. This is not the case for solubility limited radionuclides because they are released in proportion to the amount of water flux through the package (or diffusion) and the release depends on their solubility and the amount of water contacting the waste.

The dissolution of the waste form was aualyzed using dissolution models that are discussed'in Section 3.5. Where two waste fonns are contained in the same waste package, as in the case of codisposal where spent fuel and HLW are in the same package, the dissolution of the spent fbel and the HLW glass were analyzed separately (i.e., there was assumed to be no interaction between them).

2.7.2 Geosphere Transport

The radionuclides released fiom the engineered barrier system are available for transport through the geosphere to the accessible environment. The geosphere provides for both a physical and chemical

delay of radionuclides reaching the accessible environment. This delay is a function of (1) the percolation flux distribution the unsaturated zone, (2) the advective flux distribution in the saturated zone, and (3) the -rt parameters in the hydrogeolb~k layers along the likely ground- water travel path between the repository and the accessible environment. The percolation flux distribution within the Topopah Spring hydrogeologic unit is a function of the infiltration rate and the conceptual model for ground-water flow through the maturated zone. The infiltration rate is a complex function of many near-SUrfhce hydrologic factors including (1) precipitation timing, intensity, and duration, (2) surface slope orientation and angle, (3) surface geology, and (4)

'

vegetation.

Geosphere transport is also affected by the potential for radionuclide sorption on the mineral grains in the rock matrix. Whole rock distribution coefficient &'s) were used based on laboratory-derived data and the "minimum k, concept" (Meijer, 1992) for the highly-sorbed radionuclides. These distribution coefficients are related to the individual hydrogeologic unit. Although the actual retardation within any particular unit is expected to be spatially variable due to mineralogic heterogeneity and perhaps local geochemical variability, this stochastic effect is not considered in the current total system performance assessment. The use of the "minimum" k, value (i.e., most conservative h m a release or peak dose perspective) obviates the need to account for the spatial variability explicitly. The distribution coefficient (kd) for Neptunium has been modified to reflect the most recent infommtion that is being used in FEHMN, the site-scale transport model (Dash et al., 1995). In particular, the maximum Neptunium is set equal to 2.5 cc/g in zeolitic units and 0.0 cc/g in nonzeolitic uuits (Robinson et d.,1996), whereas in TSPA-1995 the values were 0.5 cclg and 1.0 cclg, reqxgtively.

The entire advective fIux distribution incorporates the effects of the large-scale spatial heterogeneity of aquifer properties. Small-scale heterogeneity is included through the use of dispersion in the solution of the one-dimensional advective-dispersive equation. Because of the one-dimensional nature of the solution algorithm, only longitudinal dispersion is simulated, i.e., there is no transverse

. dispersion. This is conservative when considering predictions of peak conceritration or peak dose.

2.8 BIOSPHERE AND RADIATION DOSE

Although only engineered barrier and natural barrier (i.e., geosphere) models and parameters are required in the prediction of cumulative releases of radionuclides at the accessible environment, the calculation of radiation dose requires the defmition of the potentially exposed population(s) and the . potential biosphere pathways by which individuals are exposed to any radionuclides released. In the study, as in TSPA-1995, it was assumed that an individual receives the peak dose by taking all his or her drinking water from the tuff aquifer. It was also assumed that this individual is located at the point on the accessible environment boundary which corresponds to the peak of the radionuclide concentration within the tuff aquifer. The individual is assumed to consume two liters of drinking water per day, 365 *days per year. Dose conversion faetors, which convert radionuclide concentrations to radiation doses from drinkkg water, were derived h m published values of the U.S. Environmental Protection Agency @PA, 1988).

2.9 BASE CASE CALCULATIONAL SCENARIO

The performauce assessment that is presented in this report uses the same general appmach as is used in TSPA- 1995 with parame& that have been updated to the current understanding of a repository in Yucca Mountain. The analyses assume that the repository loading is for 83 MTHM/acre, the waste packages are centered in the drift on a crushed tuff invert, and the drifts are not backfilled. The repository is assumed to contain 63,000 MTHM of commercial spent fuel, 2436 MTHM of DOE SNF, and enough HLW glass to co-dispose the high-enriched DOE spent he1 assuming a fissile content of 14.4 kilograms per package (8745 MTHM of HLW). This is a total repository loading of 74,18 1 MTHM. Because this loading is above the 70,000 MTHM S i t set by the Nuclear Waste Policy Act, results will be reported for both the total amount (74,18 1 MTHM), and for a repository where the DOE spent bl and HLW are scaled to 7,000 MTHM (i.e.,a 70,000 MTHM repository).

The release model used is for drips on 30% of the waste packages while the remaining waste have no drips and release from their engineered bamer is by diffusion through the crushed tuff invert. The waste packages are assumed to have galvanic protection fonned by the two package metals with 900? of the packages having 75 % protection. Waste package Mure is by pit formation with cornsion rate being a function of corrosion depth and a strong correlation with corrosion depth at neighboring locations of the package (patches of corrosion). Co~osion is initiated by a combination of temperatwe and relative humidity as was done in TSPA-1995. The flux in the unsaturated zone was taken from the calibrated site-scale model and abstracted into six vertical columns from the repository footprint to the water table. The flux in these columns range h m 4 to 10 mm&r with an average of 6.2 mm/yr_. The three-dimensional flux distribution from the s i te-de umaturated-zone flow model served as input to the thennohydrologic model that was used to calculate tern-, e o n , k d relative humidity needed for the waste package failure model. Because of the coupling among models it would not be correct to arbitrarily vary a parameter such as percolation flux. For this reason climate cycles are not included. Based on the regional modeling the average flux in the saturated zone has been reduced from 2.0 m/yr to 0.3 1 m/yr (M&O, 1996).

Based on recent work the distribution coefficient for Neptunium has been modified somewhat (Robinson et al., 1996). The solubility of Neptunium has been reduced by two orders of magnitude (Sassani and Siegmann, 1997), a reduction that has a significant effect on dose history at the accessible environment (i.e., the peak dose at the accessible environment is fiom the combination of q c and 'q and occurs at tens of thousands of years rather than being h m a'Np and occurring at hundreds-of-thousands of years as in previous perfomance assessments). Because the peak dose . occurs earlier dose history plots only need to be generated for time periods of 100,000 years.

East West

Soiitarjo Canyon fault

Figure 2.3-1 Schematic Cross Section Through Y U ~ ~ountain Showing the Potential Rcposito~y . in Relatonship to Faults, Stratigraphic Units, and the Ground-Water Table (A&r Oah et al., 1985). Units symbols are; TCW (Tiva Canyon welded unit), PTn (Paintbrush nonwelded d), TSw (Topopah Spring welded unit), CHn (Calico Hills nonwelded unit), PPw (Prow Pass welded unit), and BFw (B&og welded unit).

Figure 2.3-2 Major Geologic Units, Lithology, and ~ ~ d r o ~ e o l o ~ i c a l Units at Yucca Mountain, Nevada (Modified from Montazer and Wilson, 1934 using Sawyer et al., 1995).

2-16

Hydrogeological Unit

Tiva Canyon TCw

Paintbrush PTn

Topopah Spring TSw

Calico Hills CHn

Crater Flat

Lithology

densely welded tuff moderately welded tuff

partially welded tuff nonwelded tuff bedded tuff

nonwelded to moderately welded tuff

bedded tuff

nonwelded to moderately welded tuff

bedded tuff

nonwelded to partially welded tuff

moderately to densely welded tuff

densely welded tuff basal vitrophyre

moderately to partially welded tuff

bedded tuff nonwelded to partially

welded tuff

nonwelded to partidy welded tuff

bedded tuff

nonwelded to densely welded tuff

bedded tuff

nonwelded, moderately welded, to welded tuff

Geological Unit

5

d 2 2 -3

Tiva Canyon Tuff

Yucca Mountain Tuff

Pah.Canyon Tuff

Topopah Spring Tuff

Calico Hills Formation

i

2 8 , E b Q) Y

E U

Prow Pass Tuff PPw

BullfrogTuff BFw

Tram Tuff

Figure 2.3-3 Regional Ground-Water Flow System in the Vicinity of Yucca Mountain Showing Three Ground-Water Subbasins (Modified fiom DOE, 1988).

Figure 2.3-4 Idealized Geohydrologic Cross Section from Yucca Mountain to Eagle Mountain (After Czarnecki, 1989).

Development Access Ramp I

Figure 2.4-1 Schematic Plan View of the Repository Layout.

Pier W'aste Package Support

Figure 2.4-2 Schematic Diagram of Waste Emplacement.

I

Thermohydrologic Results (NUFT)+Temp., Rel. Humid.,

and Sat. .

I % of Waste Packages . with Drips

Waste Package Degradation History (WAPDEG)+History

of Pit Development

Cladding Failure Waste Form Exposure

D'iusion Coefficient

Waste Form Dissolution

+ 'I 1

Engineered Barrier Release Model Radionuclide

I Solubility I

Radionuclide Release from

I Engineered Barrier System Diffusion from Waste Package

Diffusion or Advection Through Invert

Figure 2.5-1 Engineered Barrier System Processes.

'loo 1000 10000 100000 Time (years)

Figure 2.6-1 History of First Pit Development Through Waste Packages.

10 100 1000 10000 100000 1000000 Time (years)

Figure 2.6-2 Average Numba of Pits Through Waste Packages in Eight Repository Failure Groups.

Uniformly Distributed Solid Waste (spent fuel or HLW glass monolith)

Radionuclide Rekase

Figure 2.7- l(a) Conceptual Model for Diffusive and Advective Release fiom Both the Wast Package and Other Components of the Engineered Barrier System ("Drips on the Waste Form").

Wall

'it

I Unifarmly bistri huwd Sdid Waste (spent fuel or HLW glass monolith)

Container iVa

Radldnuclide Release

Figure 2.7-l(b) Conceptual Model for Diffusive Release from the Waste Package and Both Diffissive and Advective Release from Other Components of the Engineered Barrier System ("Drips on the Waste Package").

11 Uniformly Distributed

11 Solid W s t e [spent fuel ar HLiV glass monolith)

I I

I 8 I I

I I Container Wall

I I

I I #I

I I

8 I

I I

Figure 2.7- l(c) Conceptual Model for Diffusive Release fiom Both the Waste Package and Other Components of the Engineered Barrier System ("Capillary Barrier Effect").

Table 2.9-1 Major Parameters Used in Assessing the Performance of DOE SNF in a Potential Repository at Yucca Mountain, Nevada

Thermal Load .

Diffusion Coefficient in Waste Package

Corrosion Initiation

Galvanic protection .

Cladding

Radionuclide Solubility

EBS Release Model

Crushed Tuff hvert Diffusion Coefficient

Invert Diffusion Path Length

Percolation Flux

- -

Saturated-Zone Flux - -

Saturated-Zone Porosity

Dispersion Coefficient

Climate Change

VALUE

83 MTHM / acre

Controlled by relative humidity and temperature

90% of the waste packages have 75% galvanic protection

No credit taken for spent fuel cladding

Neptunium solubility reduced two orders of magnitude fiom that used in TSPA-1995 (Sassani, 1997) - 30% ofwaste packages have drips and the remaining 70% have diffusive reIease

None

4.0 to 10.0 mm / yr spatially distributed with an average of 6.2 mm / yr (Bodvarsson et al., 1996)

Average value of 0.3 1 m /yr

Unsaturated Zone - Marcovian Saturated zone - 10% of Path Length - 500 m

None

3. SOURCE TERM

The source texm is a description of the rate and duration of radionuclide release b m the engineered barrier system. It is a hct ion of the total inventory of mdionuclides, the number of waste packages, the waste package failure rate, the alteratioddissolution (leaching) rate of the waste form in the near- field environment,the surface area exposed, the radionuclide availability to leaching (i.e., fiee radionuclide inventory), the solubility of a specific radionuclide, and the mode of migration through the failed package and engineered barrier system (i.e., by advection or diffusion). The rate of release over time is influenced by temperature, the mount of water present, the rate of movement of the water, and the chemistry of the water in the vicinity of the waste. This chapter describes the primary components of the SO- term used for each waste form. It includes the radionuclide inventory, the waste packaging, the physical characteristics of the waste, and the dissolution/alteration model.

The information on inventories presented here was received from several DOE sites, including the . idaho National Engineering and Environmental Laboratory (INEEL), Hanford, and Savannah River and then merged into a single data base. The data given in this chapter are based on the National Spent Fuel Database, which was used as the authority to resolve any discrepancies in data provided by the DOE sites. The approach of attributing a spent fuel to a particular DOE site is based on the Record of Decision for the Final Environmental Impact Statement for management of DOE spent fuels (DOE, 1995) as modified by the 1995 settlement agreement among the State of Idaho, the Department of the Navy, and the DOE.

3.1 W-MTE FORM CATEGORIES

The DOE has recently categorized all of its spent fuel into categories that were developed based on fuel composition and characteristics (Stroupe, 1997). The primary concern in grouping the spent fuels was assigning every fuel to a category and making certain that all of the spent fuels fit into a category. The grouping of spent fuels into categories is dependent on the analyses that will be performed using a representative spent fuel fiom that category. For example, the spent fbels are grouped differently for conducting a TSPA than they are for conducting criticality analyses because different attributes of the fuel are required for performing the different analyses.. The DOE SNF categories for TSPA me shown in Figure 3.1-1. Only the first 13 TSPA categories are analyzed in this performance assessment because the Sodium bonded spent fitel (Category 14) will likely be treated prior to disposal, and the final form of the treated fuel has not yet been determined. Also, the Navy spent fuel (Category 15) was not analyzed.

The physical properties of a spent fuel category are assumed to be bounded by the properties of a specific spend fuel or a combination of several spent fuels in that category depending on the type of analyses that are being performed using these data (i.e., performance assessment or criticality). Table 3.1-1 'shows the 13 categories of spent fuel that were analyzed in this performance assessment

' and the typical fuel that was used to bound the physical properties of that category.

The total amount of DOE spent fuel in the 13 categories is approximately 2436 MTHM. The

distribution of these spent fuels among INEEL, Savannah River, and Hanford is shown in Table 3.1 - 2. This table indicates that INEEL has spent fuel in all categories except 6 and 7; Savannah River has fuel in Categories 5,6, and 7; and Hanford has fuel in Categories 1,4,5,10,11, and 1 3. By far the largest amount of spent he1 is in Category 1 with 2136.8 MTHM of spent he1 that is represented by N Reactor fuel (see Tables 3.1-1 and 3.1-2). Categories 2,3,8,10, and 13 contain the smallest amount of spent fuel, and range in amount fiom only 0.04 to 3.93 MTHM.

3.2 RADIONUCLIDE INVENTORIES

The analyses of perfomance of disposal of DOE SNF requires the determination of the radionuclide inventory for each category of fuel that will be disposed. The computer code ORIGEN2 (Oak Ridge Isotope Generation) is widely accepted for calculating spent fuel radionuclide inventories (ORNL, no date). An ORTGEN;! run requires detailed input of data for the fuel core composition, the power history of the reactor, and operating conditions of the reactor. In particular, the nuclear cross section libraries for each fuel type are required for the reactor in question, and this library development is a lengthy and involved task. With rigorous library development and comparison of calculational results with measured values, the ORIGEN2 model is capable of accuracy sufficient for nuclear =actor operational support. Such carefully developed nuclear cross section libraries were not available for all fuels evaluated in this performance assessment, and some of the input data for the ORIGEN2 had to be estimated.

A conservative approach for the use of ORIGEN2 was adopted to estimate the radionuclide inventory for each spent fuel category. For categories that have large amounts of spent fuel (i.e., Category -1 represented by N Reactor fuel) an effort was made to minimize the degree of conservatism. For small inventory categories, minimizing the degree of conservatism is less important, and it is easier to accept the "penalty" of an overestimate of the amount of radionuclides when the dose they produce is relatively small compared to the dose resulting from larger amounts of spent fuel in other categories.

-.

The radionuctides included in the performance assessment are based on screening that was conducted for TSPA-1993 (Andrew et al., 1994). The screening was done in two ways. First, radionuclides were screened using the ratio of their inventory to EPA Table 1 release limits (EPA, 1985). The ratio of the weighted average inventories of specific radionuclides were determined for spent commercial fuel to corresponding the EPA Table 1 values for 1,000, 10,000, 1 00,000, and 1,000,000 years. The fractional contribution of each, isotope to release at a time of 1,000, 10,000, 100,000, and 1,000,000 years was calculated assuming a combiition of delay due to waste package - lifetime and retarded transport of 1,000 to 1,000,000 years. Isotopes which contributed at least a fiaction of the EPA release limit at any of the selected times passed this screening. The entire decay chain for daughters which contributed greater than lQ5 of the EPA release limit at anytime were also included (Andrew et al., 1994).

Second, the screening was also conducted using dose for the same time periods. The waste form was assumed to be altered at a rate of 10" of the total inventory per year. The isotopes were assumed to dissolve, as they were made available by the assumed waste fonn alteration rate, at the

maximum solubilities according to NAS (1983), EPRI (1992), and Barnard et d. (1992). The advective, downward flw in ground water moving through the .unsaturated zone was assumed to occur at 0.1 d y r over the footprint of the repository. On arrivd at the saturated zone, the isotopes were assumed to mix in the 'saturated zone with a flow rate of 10,000 m3/yr. Ingestion of 700 litadyear by a person using this ground water was assumed. The ingested dose was calculated using the maximum effective (whole body) dose conversion factor from DOE (1988a), NRC (1981), or EPA (1988). The M o n a I contribution of each isotope to total dose at times of 1,000, 10,000, 100,000, and 1,000,000 years was determined. For radionuclides with two or more isotopes present in the waste,.the solubility l i t was set for the element (i.e., all isotopes) and then proportioned between the individual isotopes by the mass hction present at the comsponding time. All isotopes contributing less than 1 PS of total dose at any time period were eliminated from the inventory unless they were in the decay chain for daughters which contributed lCs of total dose at any time. This process yields 39 radionuclides. The process of using the maximum dose conversion factor from the three alternatives is sufficiently conservative that the change in flux (i.e., 0.1 to 10.0 mmlyr) would not yield additional radionuclides to those being considered. For the analyses in this study the EPA dose conversion kctors were used bec8use the EPA will set the standard for the potential repository.

In order to estimate the 39 radionuclides needed as input to the performance assessment, ORIGEN2 was used to model the fuel types in a given category of spent fuel (in the National Spent Fuel Data Base). Spreadsheets were prepared to merge the results (radionuclide concentrations) from the ORIGEN;! runs into a composite radionuclide inventory for each category. For each he1 entry selected in the category, a representative ORIGEN2 m was made. The inventory was then calculatedas a scaled amount based on Uranium content of each he1 type in the category. Tods were then calculated for each of the 39 radionuclides.

Fuels for which ORIGEN2 runs were used directly are as follow: Category 1

- N Reactor Category 3

FERMI (Enrico Fenni Reactor) Category 4

Commercial Pressurized Water Reactor (PWR) Pathfinder Power Burst Facility (PBF) Transient Reactor Test (TREAT)

Category 5 Pulstar Buffalo Three Mile Island (TMI)

Category 6 Advanced Test Reactor (ATR)

Category 8 Fort St. Vrain .

General Atomics-High Temperature Gas Cooled Reactor (GA-HTGR)

Category 9 Peach Bottom

Category 10 Fast Flux Test Facility (FFTF) Carbide

Category 11 Fast Flux Test Facility (FFTF) Oxide

Category 12 shippingport

Category 13 Training Research Isotopes-General Atomics (TRIGA) '

Results from these spent fuels were also used as templates for fuels for which no direct runs were available (i.e., Category 2).

The packaging of the fuel was calculated separately, as described in Section 3.3. The cumulative curie count for each category was divided by the total packages in the category to arrive at the mean inventory for each package. The fuel inventories used in this study were based on the National Spent Fuel Database (DOE, 1996), which was used as the authority to resolve any discrepancies in. other data After the curie inventories were calculated for each category, the number of MTHM was totded and minor errors were noted between the calculated amount and that h m the National Spent Fuel Data Base. For categories 1,4, and 5 the total curie inventory categories were adjusted by factors of 1 . O M , 1.0163, and 0.9884, respectively. This had the effect of estimating the inventory for the fuels not specifically included in the analyses as being similar to the rest of the category.

Composite spreadsheet totals of the curies in a particular category were apportioned equally across the total number of canisters in that category. This results in a mean value per canister which does not consider that canisters within a category are of different size. For instance, no distinction in curie inventory is made between 10,17, and 24 inch diameter fuel canisters, or for canister length.

. - The analyses resulted in the radionuclide inventories that are presented in Appendix A for the 13 categories of DOE spent fuel that are analyzed in this study. The radionuclide inventories for each of the 13 categories were converted into the units of CifMTHM and are presented in Table 3.2-1. The total MTHM for each category was discussed earlier (see Table 3.1-2).

The repository modeled in TSPA-1995 was assumed to contain 63,000 MTHM of commercial spent fie1 in 6,468 packages, with an average radionuclide inventory which is presented in Table 3.2-2. The radionuclide inventory for the HLW glass is also presented it Table 3.2.2 and is the same as that used in TSPA-1995.

3 3 WASTE PACKAGE TYPES AND LOADING

The fuel loading configurations for these analyses were based on the co-disposal approach for high- and medium-enriched fuels prevalent in the DOE spent fuel inventory. This was done to reduce the potential for criticality by limiting the fissile mass content of each package based on its enrichment.

High-level waste glass canisters are considered to be emplaced in the same package with the spent fuel. The co-disposed spent fuel, for this pedonnance assessment, is assumed to be packaged in four canister HLW packages and co-disposal packages of two lengths, %87 meter (12.8 A), and 5.25 meter (1 5.2 ft).

The DOE SNF is assumed to be contained in 266 nun (10 in), 450 mm (1 7 in), and 640 mrn (24 in) canisters. Co-disposal in a HLW package can be achieved in two ways; substitution of a 640 mrn canister of spent f k l for one of the canisters of HLW glass (a 3 x 1 package), or placing a 266 mm canister of spent fuel in the central cavity which is left by the four canisters of glass (a 4 x 1 package). A 3 x 1 package would be Iike the 4 x 1 package shown in Figures 3.3-1 except the central canister would not be present and one of the canisters of glass would be replaced by a canister of spent fuel. In the package nomenclature used in this report the first number refers to the number of glass canisters and the second number refers to the number of spent fuel canisters in the package (i.e., a 4 x 1 package would contain fbur canisters of glass and one canister of spent he1 as shown in Figure 3.3-1). The co-disposal package contains five canisters of glass and a central 450 mm canister of spent fuel (Figure 3.3-2). Hence, using the waste package nomenclature it is a 5 x I package.

Low-enriched spent fuel, in this report, is assumed to be disposed in four canister HLW package with no canisters of HLW (four 650 mm canisters of low-enriched spent fbel). These are referred to as 0 x 4 packages.

Packaging decisions for this performance assessment of DOE SNF were made using two measures, physical she and fissile content. In the absence of complete, definitive criticality analyses for each fie1 type, the amount of a5U for high-enriched fie1 (i.e., enrichment greater than 20 percent a5U) was restricted to approximately 14.4 kg per package (Research Reactor Spent Nuclear Fuel Task Team, 1996). The Hanford spent fuel is being treated and packaged in a Multi-Canister Overpack (MCO) for interim storage. The MCO has an external diameter of 25.3 1 in. (about 640 mm), and is assumed to-fit into a HLW canister position. The Savannah River spent fbel is being placed in 450 mm (17 in) canisters and it is assumed to be disposed in co-disposal packages. The INEEL spent fuel'dimensions and fissile content allows most of the fuels to be contained in 450 mm (1 7 in) canisters. For the INEEL spent fuels the 266 mm (1 0 in) canister was used when high fissile loading resulted in a small amount of fuel, and 640 mrn (24 in) canister was used for fuel of large dimensions. Details of the packaging of DOE spent fuels are provided in Appendix A.

Table 3.3-1 shows the waste package configurations for each category of DOE SNF. On the basis of the number of HLW canisters needed for co-disposal (Table 3.3-1 or Appendix A), the total size of the repository becomes more than the current 70,000 MTHM limit for the first repository (i.e., 74,181 MTHM). The amount of HLW needed for co-disposal will likely go down as detailed criticality analyses are performed (i.e., 14.4 kg of fissile material is conservative), and different packaging configurations may change the amount of HL,W needed. 'Ihe results reported for the entire repository are for a 74,18 1 MTHM. For comparison the amount of DOE SNF and HLW are

'

scaled to 7,000 MTHM which yields a 70,000 MTHM repository.

The commercial spent fuel assemblies will be disposed in the 21 -PWR assembly package shown in Figure 3.3-3 or in the 44-BWR assembIy package shown in Figure 3.3-4. For calculational purposes the 63,000 MTHM of commercial spent fuel was assumed to be disposed in 6,468 PWR waste packages containing the average radionuclide inventory shown in Table 3.2-2.

3.4 PHYSICAL PROPERTIES OF SPENT FUEL

Fuel properties were determined by the judgement of those personnel most familiar with the fuel '

inventory, and a consensus on the values was gained by review among the local spent fuel programs '

at the DOE sites (Cresap, 1997). The grouping of fuel into categories, based on chemical fuel form, aided in assigning common properties to each of the categories. The physical properties that are used in the performance assessment are presented in TabIe 3.4-1 for each category of DOE spent fiel.

The prior performance assessments (i-e., TSPA-1995) have compiled the physical properties for commercial spent fuel, and these values were adopted or used as a beginning point for estimating values for the DOE spent fie1 categories. For instance, some fuel categories use the same value for dissolution rate as commercial fuel, others used the commercial rate scaled by an appropriate factor, and still others used entirely different models. Using existing models wherever possible simplified the approach for both data generation by the National Spent Nuclear Fuel Program and modeIing done in this performance assessment. The effect of a range of physical properties of the fuels on performance is examined in Chapter 5, SensitivityNncertainty Analyses.

Table 3.4-1 provides estimates of fuel dissolution rate, surface area, cladding failure, fiee inventory, and gap inventory. The dissolution under wet oxic conditions is taken fiom the dissolution models that are described in Section 3.5 or as some fraction of the results of these dissolution models. For dry oxic conditions the dissolution rate is assumed to be zero as in the case for commercial spent fuel. Surface area for DOE fuels was based on the area and weight of the fuel meat (unclad fuel) itself. Calculations were simplified by the fact that the chemical form of the fie1 meat within each category was the same. Where different geometries or dimensions contributed to the same category,

I a dominant type was selected or average values were used. Conservative estimates were made of i I

the fiaction of fbel 'cladding that has failed. These numbers are typically high, reflecting the I approach that little, if any, credit is claimed for fbel cladding as a barrier to releases. The value

given is an estimate of the initial condition at the time of waste package failure, and normal corrosion processes would proceed after that time. In the analyses reported in this document no credit is taken in the base case for the fuel cladding.

The last two columns of Table 3.4-1 provide the fiee and gap radionuclide inventories for each fuel category. The fiee inventory is the W o n of the radionuclide inventory that has been released fiom the fuel, but still contained in the disposal package until the time the package is breached. It is then available for immediate release. Since the fbel in most cases has been stored a long time prior to repository package emplacement, most of the inventory available for immediate release muld already be gone prior to sealed containment in a canister. After the canister is sealed for repository disposal, conditions in it are benign, and not likely to facilitate degradation of the &I. For these

reasons, the fiee fraction of the inventory is low. The gap referred to in the gap inventory is between the fuel meat and the cladding, The inventory is the M . o n of radionuclides that has migrated fiom the fuel meat to the gap and is available for immediate =lease when the cladding is penetrated. This inventory may be specified separately for diffmnt isotopes (i.e., 14C). Some of the W E spent fuels are physically constructed so as to eliminate a gap that could accumulate radionuclides. For example, the N Reactor fuel meat is co-extruded with the cladding which eliminates the presence of a gap. Thus it has no gap inventory (Table 3.4-1).

3.5 WASTE FORM DISSOLUTION MODELS

Dissolution models for metallic, wbide, ceramic, and oxide (commercial) spent fuel, and for HLW glass are presented below. The metallic and carbide models were taken from the analysis of DOE SNF conducted by Sandia National Laboratories (Rechard, 1995), the models for commercial spent fuel and HLW were used in TSPA-1995 (Andrews et al., 1995), and the ceramic model was used in the evaluation of plutonium waste forms (Duguid et al., 1996). All of these dissolution models were also used in the assessment of the performance of INEEL spent fuels (INEEL Task Team on Spent Nuclear Fuel, 1997). A comparison of the glass and spent fuel dissolution models that are described below is presented in Figure 3.5-1. This figure indicates that the largest dissolution rate over time is for metallic spent fuel and the slowest is for carbide spent fuel. The dissolution rates in order fiom fastest to slowest are metallic spent fie1 under wet oxic conditions, metallic spent fuel under humid oxic conditions, oxide spent fuel, HLW glass, ceramic, and carbide spent fuel. The change in rate over time is caused by the buildup of corrosion products on the surface of the waste form which slows the rate of dissolution. Initially there are about seven orders of magnitude difference between the fastest and slowest waste form dissolution rate (i.e., metallic and carbide), and after 100,000 years there are about five orders of magnitude difference in dissolution rate between these models.

35.1 Metallic Spent Fuel -.

The model formetallic spent fuel (Uranium metal and Uranium metal alloy) used in this study was taken fiom the 1994 performance assessment of DOE SNF (Rechard, 1995). The model is based on the equation that follows:

Where:

M = mass of layer corroded in a time step (kg), A = Arrhenius-type pre-exponential term 0cg/m2s), B = Arrhenius-type advation energy term (K), t, and t1 = time at the beginning and end of the time step in seconds, C = time dependence term,

D = saturation dependence tern, which is O,1, or 1 -e(-L('-m)

where a = ln(.01)/(Sg9 - TS) Sat = fkacture water saturation, TS = threshold fracture saturation b'elow which wet corrosion does not occur, and S, = fracture saturation where the wet corrosion rate is 99% of the corrosion rate at 100%

saturation. E = oxygen concentration dependence tenn, and SA = surface area of the layer (m2).

Wet oxic conditions are assumed when the temperature in the repository is below 100 O C and humid oxic conditions are assumed at all other times. The parameter values used for metallic spent fuel analyses are:

A = 9.4 x 1 O3 kg/m2s for wet oxic conditions, A = 1.35 x Id kg/m2s for humid oxic conditions, B = 7970 K for wet oxic conditions, B = 7240 K for humid oxic conditions, C = 1 for wet and humid oxic conditions, D = 1 which is assumed to be consemative, and E = 0.2, the oxygm concentration tnm has been approximated by x,, which is the mass fraction of air within-the gas phase.

3.5.2 Carbide Spent Fuel

The model for carbide spent he1 used in this study was taken h m the 1994 performance assessment of DOE SNF (Rechard, 1995). The 1994 assessment used a model based on the equation that follows: - -. -

For silicon carbide coatings,

A= 3 x 1 0'12 Is, B-OK, C = 1 sec, D = 1 which is assumed to be a conservative assumption, E = 0.2, and M,, = the mass of the layer at time zero.

The equation above can be used to calculate the mass corroded at any tim step. In order to obtain

the dissolution rate, the mass of the layer at time zero and the surface area must be known. For carbide spent fuel, the surface area of the silicon carbide is assumed $0 be 1325 m2 and the mass of the layer is assumed to be Mw = 126 kg. This model describes tlie dissolution of the silicon carbide coatings on the fuel particles, and does not reprent the fbel itselt Once the coating has failed the he1 reacts rapidly with water releasing the inventory of the fbel particle. Thus, the model only applies to Category 8 SNF, but because of their small inventory (Table 1 .l-2) it was applied to Category 9 and 10.

3.53 Ceramic Spent Fuel

The dissolution model for ceramic was extracted from Lappa (1995). The composition of the ceramic is assumed by Lappa (1995) to be similar to that of Synroc-C, a titanate ceramic. Reeve et a1 (1 989) propose the cumulative release from Synroc to be:

Where:

Q is the release per unit surfhce area (g/m2), Q, is the instantaneous release fiom grain boundaries and metastable phases (g/m2), 0 is a complex kinetic function that accounts for ionic diffusion, selective matrix attack, etc

(g/m2), S is me solubility of matrix (glm3), F is the ground-water flow rate (m3/day), A is the surface area (m2), and t is time (days).

Lappa (1995) states that it is likely that the long-term release fiom the Synroc is controlled by the third term in the cumulative release equation. Existing data indicate that S is less than or equal to 0.007 dm3 based on a long-term leaching rate of less than or equal to 104g/m2/day when SNC is 10 m'' at 70 degrees C in deionized water (lappa, 1995). The leaching rate also increases with increasing temperature (Ringwood et al, 1988). This temperature effect is described by:

R = a 1 O*'OOOm

Where:

R is the leaching rate (g/m2/day), T is temperature (degrees K), and a$ are constants.

Based on the Synroc ddta in Ringwood et a1 (1 988), = 1.0 and a = 0.082 g/rn2/day if the long-tern leaching rate is assumed to be 104 g/m2/day at 70 degrees C. These are the same conditions as assumed in TSPA-1995 for the HLW glass. The radionuclides are released fiom the waste based

on its alteration rate, and are then transported at a rate dependent on whether they are alteration- controlled or solubility-limited radionuclides.

The alteration/dissolution rate used for ceramic is for a monolithic form, &d when cracked along grain boundaries may be substantially different and should be m h in the ongoing research and development program to provide a more realistic rate for the repository environment.

3.5.4 Commercial Spent Fuel

The radionuclide release fiom spent commercial fuel used in TSPA-1995 is presented below. The dissolution~alteration rate is aEected by temperature, water satmation, and water flux in the vicinity of the fuel. The radionuclides are released fiom the spent fuel based on its alteration rate, and are transported from the package according to the radionuclide solubility which leads to some radionuclides being alteration controlled and others being solubility limited. The alteration controlled radionuclides are I4C, '9Se, T c , IsI, and '35Cs, and the remaining radionuclides are solubility limited.

For radionuclide release from the spent fire1 waste form (firel pellets), two distinct release modes are considered: (1) instantaneous release, and (2) matrix release. The instantaneous release mode consists of species in the gap between fuel pellets and cladding, and species on fuel grain boundaries. These species arc referred to here as "gapinventory species" and include 14C, "'Cs, I3'Cs, '2q, T c , and '%e. The species are characteristically mobile and highly soluble in water. Typically, 1 to 2 percent of their inventories are present in these regions of instantaneous release (Apted et al., 1989). All other radionuclides are assumed to be located in the spent fuel matrix. h e gap fractions of gap-inventory species are assumed to be available for immediate release as soon as both waste package container and cladding fail.

The distribution of the I4C gap inventory represents the uncertainty of 14C inventory in the gap and the oxidationlayer on the cladding surface, and is the same as used in TSPA-1993 (Andrews et al., 1994). Van Kanynenburg et al., (1986) reported that about 65 percent of the total "C inventory is present in cladding, crud and other fucl assembly hardware. However, only the I4C in the gap and the grain born* of the spent fixel matrix (about 1 percent), and in an oxidation layer on the surface of the cladding is available for instantaneous release. The release rate for the rest of the I4C is considerably slower ( B a d et al., 1992). The release of I4C is calculated assuming that all the I4C available at a given simulation period migrates out of the EBS as a gas and then dissolves in the .

. aqueous phase, i.e. no gaseous transport in the geosphere is considered.

The second release mode is matrix release in which the release rates of radionuclides are proportional to fuel matrix alterationldissolution rate. Radionuclides released in this mode are referred to here as "matrix-release species", and may be grouped by their solubility limits into alteration/dissolution-limited species for highly soluble species and solubility-limited species for relatively inso1uble s*es.

A semi-empirical model for intrinsic dissolution (alteration) rate of the spent fuel matrix was developed fbm the experimend &a reported by Gray et d., (19b2) M Steward and Gray (1 994). For this analysis, the post-closure environment inside the potential repository is assumed to maintain the atmospheric oxygen partial pressure of 0.2 atm. The model is expressed as a function of temperature, total carbonate concentration, and pH of contacting water as follows:

=1 logks = uo + - + 3 10g[Co3] + u3pH + E T \

where k, is the intrinsic dissolution rate of spmt fuel (mg/m2*day), T is temperature (K), [CO,] is the total carbonate concentration of the contacting ground water (in molarity units), and c is a term representing uncertainty not included in the model. The parameter values determined for the hctional form are: a, = 7.323 * 0.957, a, t -1585.2 * 303.3, a, = 0.2621 0.0743, and a, = -0.1 140 k 0.0679. The dissolution rate strongly depends on temperature and total carbonate concentration, and is less influenced by pH.

The model prediction and the associated uncertainty of the dissolution rate as a function of temperature at total carbonate concentrations of 0.002 and 0.02 M were evaluated and compared to experimental data in TSPA-1995. The spent fuel dissolution rate increases with temperature and total carbonate concentration in the water contacting the waste enhances the dissolution rate, but to a smaller extent than temperature (Andrews et al., 1995). The enhanced dissolution rate is due mostly to increased complexation of uranium and other actinides with carbonate ions. The values of dissolution rate predicted by this equation are somewhat higher (about 2 to 4 times depending on temperature and the total carbonate concentration) than those used in TSPA-1993 (Andrew et al., 1994). This difference is caused by the replacement of the TSPA-1993 temperature dependance tm. (exp (T)) with a more physically reasonable Anhenius temperature dependence term (exp (1 /T)) in TSPA-1995.

I The actual spent f k l alteration rate is detemined by multiplying the intrinsic dissolution rate by the i available surface area that is exposed (or wetted). To calculate the actual fuel matrix alteration rate

in TSPA-1993 (Andrew et al., 1994), the surface area of spent fbel available for dissolution was a function of the fiaction of fuel that was wet. In TSPA-1995, assuming cladding fails at the same time as the waste container, the entire waste form d a c e area is assumed to be exposed to the near- field environment and covered with a "thin" water film once the waste container is pitted through its wall thickness. This conceptualization is based on the assumptions that, as soon as the waste container fails, the inert erivironment inside is quickly replaced with the near-field environment, and

I the moisture fieely moves through the space inside the "failed" waste container to readily wet the ! waste form. This new approach provides higher degrees of dissolution than considered in earlier

studies (TSPA-1993). . I I

A recent report by Gray and Wilson (1995) provides an estimate of the specific suTface area of non- oxidized spent fbel. From the measured spent fbel pellet surface area (1 50 mm2 per millimeter of

fuel rod length-Bamer, 1985), a geometric specific surface area was calculated to be 2 . 2 ~ lo-" m2/g (Gray and Wilson, 1 995). A factor of 1 8 was suggested for taking into consideration the surface area contributed fiom cracks that have formed during reactor operation and some penetktion of the grain boundaries by water. The resulting specific ntrfacc area of 3.96~10" m2/g was recommended as a reasonable minimum d a c e area for spent fbel (Gray and Wilson, 1995). A suggested maximum Surface area derived from the Bnmauer-Emmet-Teller (B.E.T.-Brunauer, 196 1) surface area measurements is about 0.1 m2/g (Gray and Wilson, 1995). A specific surfice area of 3 8.15 mm2 per millimeter of he1 rod length was selected h m Bamer (1985). This is equivalent to 5.57 x 10" m2/g (using a fuel rod diameter of 9.3 mm and a he1 density of 10.08 g/cm3), and was used in TSPA- 1995 and for the nominal case in this analysis.

3.5.5 High-Level Waste Glass

The dissolution model for HLW glass, described in this section, is only used for this vitrified waste form. The dissolution is afZected by temperature, pH, and concentration of dissolved silica in the water. The radionuclides are released fiom the waste fom based on its alteration rate, and are transported from the package depending on whether they are alteration controlled or solubility limited radionuclides. The alteration controlled radionuclides are "%e, %, v c , I ? , and 135Cs, and the remaining radionuclides in the inventory are solubility limited.

As in the spent fuel alteration/dissolution modeling discussed above, the entire d a c e area of the glass waste form is assumed to be exposed to the near-field environment as soon as the first pit penetrates through the waste container. The waste form is then assumed to be covered by a "thin" water film; and the altemtioddissolution processes are initiated when the temperature drops below 100" C.

Since no new information has been presented for models for glass waste form dissolution, the same conservative rate equation used in both TSPA-1993 and TSPA-1995 (Andrews et d., 1994 and 1995) is used. The dissolution rate equation is written as follows (Knauss et al., 1990; Bourcier, 1993):

where Rm is the dissolution rate of glass waste form &/day), S is the surface area of the glass exposed to alteration solution (m3, LW is the intrinsic dissolution rate constant for the glass (g/rn2-day) which is primarily a function of temperature and pH, Q is the concentration of dissolved silica in the contacting solution (M), and K is the equilibrium constant for amorphous silica dissolution (M).

It has been suggested that glass dissolution rates increase with temperature (Knauss n al., 1990; '

Bourcier, 1993; Bourcier et al., 1994). Using the same glass dissolution rate data provided by Bourcier (1 993), a new empirical equation was derived as:

log k,, = a, + alT + q p H + a f l ~ 2 + a 4 p ~ ' ~ + c * . . -

where bw has the same units as in the previous equation, and T is in "C. The parameter values determined are: -0.442 A 0.290, a, = 0,0307 4.58~10-~, a, = -1.17 * 7.02~ 10 -*, a, = 0.0793 A 6.38~10'~, and a, = 9.68~10'~ 695~10 '~ . The E term represents additional uncertainty not included in the model. This equation is valid from 10 to 100 "C and pH values from 1 to 12. The ratio QK in the equation above is estimated with the same empirical equation used in TSPA-1993 and TSPA-1995 which was derived h m Bourcier's temperature dependent estimates of Q and K and is given as:

Where T is in "C.

The dissolution conceptualization presented here embodies several assumptions and limitations. The radionuclides are assumed to be released as fast as the glass structure breaks down, which is conservative because it does not account for solubility-limited radionuclides. No credit is taken for the fgct that "experiments have shown that the actinides more commonly are included in alteration phases at the surtace of the glass either as minor components of other phases or as phases composed predominantly of actinides" (Bourcier, 1993). The model does not include any solution chemistry other than pH and dissolved silica concentration. However, a variety of experiments show that species such as dissolved Mg and Fe can change glass dissolution rates by up to several orders of magnitude, with Mg decreasing the rate and Fe increasing the rate (Bourcier, 1993). The model also does not include vapor-phase &eration of the glass. Glass has been observed to undergo hydration in a humid-air environment and, upon subsequent contact with water, radionuclide releases fiom a hydrated glass layer were several orders of magnitude higher than those from an unhydrated (fiesh) glass waste fogn (Bates et al., 1990; Bourcier, 1993; Ebert and Bates, 1995; Bates et al., 1995).

Nominal surface area of a HLW glass monolith (or log) is 5 m2 per canister, and e e surface area is increased by a &tor of 10 to 30 per canister to account for the cracks that form during cooling following the glass pouring, leading to a glass surface area of 50 to 150 m2 per canister (Bourcier, 1993). The current design concept calls for a HLW disposal container with four-pour canisters; thus, the total sudkce area of HLW glass waste form per waste disposal container is h m 200 to 600 m2. The surface area range is assumed to be distributed uniformly.

. TSPA Categories DOE SNF Groups

I ---4 1. U Metal Zr Clad, W o f e d , (LEU). N Reactar 1

( 1. U Metal Fuels 4 2. U Metal. AIClad, (LEU). Single Pass Reactcr 1

4 I 3. U-ZC (FEU), CP56HVKTR , 4. U-Mo, Zr Clad, (HEU). Fetmi 1

-. 25. urn oxldcr, ~r clad, (HEU FOE), LWBR 1 ( 12. Uflh Oxk!eFuels I*

26. WTh Oxde, SST C M , (HEU FGE), Ikesden 1

--I 5 U Ox&, 3 Clad, Intact (HEU). Shlppmgport PWR 1 r 6. U Oxde, Zr Clad, Intact (MEU), Saxton 1 I 7. u OX& zr clad, ~ntact (LEU), Commarwal 1

(4. Ox& Intact Fuels 8. u OrIde, SSrclad, Intact (HEU), ML-1 1

I 9. U Oxide, SST Clad, Intact, (MEU), PBF 1 i --A 10. U Orde, SST Ctad, Intact, (LEU). FFTFTFA 1

I 27. U-Zr-M, SSTAncdoy Clad, (HEU), TRlGA Flip 1

1 5 U Oxde FarleoXWfad Fmls

1 28. U-ZI-Hx. SST.ncoloy C&d, (MEUJ TRlGA Sfd 1 1 13. U-Zr-Hx Fuels

I 29. U-Zr-HX. A1 Clad. (MEU). TRlGA Alum 1

1 12. U Oxide, Fa&d or W a d , (MEU) O W L SST & Zrl

I 13. U Orw'e. FarM or Declad. (LEU), TMF2 1 I 14. U O m , Al Clad, (HEU), HFIR 1

I 30. U-Zr-Hr, Declad, (HEU), SNAP 1

- 1 15. U Ox&. A/ Clad, (MEU). FFRR MTR

16. U-A1 a U-A& Al Clad, (HEU). ATR 1 17. U-A/ or U-Aix. A1 Clad. (MEU), FRR MTR 1 18. USI. A1 Chd, (HEU, MEU). FRR MTR 1

f9. U/T~ ~ & e , ~raphrte. ~clntegnty, ~ E U ) , ~ t . st ~rarn 1 e. Graphm Low-IntegnM (HEU), Peach Bottom 1

rl0. uo. up-- 21. u o r u r n Cartvtie, NonOfaphut. (MUE FGEI SRE. FFTF carb (

I 22. MOX ZrC&d, (HEU FGE), GE Test 1 I 11. MOX Fuels If I 23. MOX, SST; (HEU FGE). FFTF-DFA 1

24. MOX Misc Clad, (MEU 6 LEU FGE). FFTF-7FAdCO 1

4-1 37. Na-Bmded, S S T M C mu. MEU 6 LEU). FERMI /Blanket 1 4 32. Class#&, (HEU), Navy I

Figure 3.1-1 DOE Spent Fuel ~gegories for Total System Performance Assessnicm (TSPA) (Modified form Stroupe, 1997).

4 POUR CANISTERS .,- .. . \

OUTER BARRIER LID INNER BARRIER LID

(ALLOY 625) \ \

UTER BARRIER LID Ad (A 51 6)

OUT 'ER BARRIER (A 516)

\ SHIPPINGPORT DISPOSAL CANISTER (304L)

INNER BARRIER LID (ALLOY 625) a

F i m 3.3-1 Waste Package Used for Disposal of High-Level Waste and Co-Disposal of DOE SNF.

OUTE

INNER BARRIER (ALlOY 625)

\

OUTER BARRIER LID

INNER BARRIER LID \

't OUTER BARRIER (A5161

L

\ INNER BARRIER

DOE SNF CONTAINER (ALLOY 625)

DOE SNF BASKET (304L)

Figure 3.3-2 Waste Package Used for Co-Disposal of DOE SNF.

OUl

INNER BARRIER (ALLOY 625)

OUTER BARRIER LID \

SIDE GUIDE (A516) \ A \ INTERLOCKING PLATES

INNER (A1

OUTER BARRl (A516)

INNER BARRIER LID \ ~ Y L

BAR UOY

IER

RlER 625)

. (A516)

TUBE (A51 6)

Figure 3.3-3 Waste Package Used for Disposal of 21 Uncanistered Spent Fuel Assemblies h m a Pressurized Water Reactor.

LID

OUTER BARRIER LID

INNER BARRIER (ALLOY 625)

OUTER BARRIER (A 516) \

\ INTERLOCKING PLATEB ( STAINLES6 STEEL BORON)

I , , . .. INNER BARRIER

(ALLOY 625) LID

f L ;:: Ib l iEk BARRER LID

(ALLOY 626)

Figure 3 .34 Waste Package Used for Disposal of 44 Uncanistered Spent Fuel Assemblies from a Boiling Water Reactor.

1 0-3 - 0 20000 40000 60000

Time (yrs)

Figure 3.5-1 Comparison of Dissolution Rates for High-Level Waste Glass and Commercial, Metallic, Carbide, and Ceramic Spent Fuels.

Table 3.1 - 1 Categories and Typical Members of DOE SNF

Typical Spent Fuel

N Reactor

Heavy Water Component Test Reactor ( H W r n ) -

FERMI (Enrico Fermi Reactor)

Commercial Pressurized Water Reactor (PWR)

Three Mile Island (TMI) core debris

Advanced Test Reactor (ATR).

Foreign Research Reactor-Materials Test Reactor (FRR MTR)

Fort St. Vrain

Peach Bottom

Fast Flux Test Facility (FFI'F) Carbide

Fast Flux Test Facility @FIT) Oxide

Shippingport Light Water Breeder Reactor (LWBR)

Training Research Isotopes-General Atomics (nut A)

r

Category

1. Uranium Metal

2. Uranium-Zirconium Alloy

3. Uranium-Molybdenum AUoy

4. Uranium Oxide, intact cladding

5. Uranium Oxide, failed cladding

6. Uranium-Aluminum Moy

7. Uranium Silicide

8. Uranium-Thorium Carbide, high integrity

9. Uranium-Thorium Carbide, low integrity

10. Uranium or Uranium-Plutonium Carbide, non-&aPhite

I 1. Mixed Oxide

12. Uranium-Thorium Oxide

13. UraniunZirconium Hydride

Table 3.1-2 Summary of Metric Tons Heavy Metal (MTHM) for Each Category of DOE SNF at INEEL, Savannah River, and I an ford'

~ o r m a t i o d - ~ ~ ~ l i e d by the National Spent Nuclear Fuel Program at INEEL (see Appendix A). - No spent fuel in this category.

Total

2,136.8

0.04

3.93

97.07

87.93

8.96

11.4

24.67

1.66

-0.22

1 1.75

49.66

1.99

2,436.08

Hanford

2,102.2

a

-- 17.8

0.15

-- -- - -- 0.17

10.2

-- 0.03

2,130.55

Savannah River

-- -- -- -- 3.67

8.96

11.4

-- -- -- - - - 24.03 .

Category

1

2

3

4

5

6

7

8

9

10

11

12

13

Total

m L

34.6

0.04

3.93

79.27

84.1 1

-- -- 24.67

1.66

0.05

1.55

49.66

1.96

28 1.5

Table 3.2-1 Radionuclide Inventory for Each Category of DOE SNF' '

Table 3.2-1 (Continued) Radionuclide Inventory for Each Category of DOE SNF'

'Radiunuclide inventory in the year 2030.

I

Iwope

'"Pu

I"PII

'%I

"'I%

'"Pu

mR.

"Ss

'"Sm

'"Sn

"Tc

'%I

"WI

"%

I ~ U

II ~ ,

IYSU

'%I

mu "Ir

CatepyI (CUMTHM)

4.73&+01

I.07Wm

&24E+OI

6.85E42

3.01E42

9.96E.07 .

8.OSEll

3.9OEOZ

638@+01

6.811342

lllE+00

8.08E-09

8.88@-07

1.0SE-I0

4.97E-06

4.22841

1.61E-02

6.11E-02

3.32E-01

1.83E-01

C a t e r n 6 (CVMTHM)

3 . 6 1 W

2 . 1 5 ~

1.09er02

29DE#l

1.22e-01

3.04E.10

6.13E-08

4.00EUlO

3 . 7 6 W -

3.58E40

1.35EiO2

3.@4@-08

1.14E.06

l.8lE-07

2.04e04

1.13E01

I.20@+00

4.33E*u)

1.28E-01

2.03E+01

Clcepy2 (CiIMTHM)

8.6%Ol

214E102

l.22E+02

4.ZOEIOI

1.83E-01

4.38E-10

2.48E-I I

S.SEE+(IO

5.30E403

5.26E*a)

1.9uE102

2 12E-08

1.71E-06

4.W-I0

1.82E-M

t.7IE-01

1.63840

6.71EUIO

3.OOE-02

3.02E+OI

Category3 (CVMTHM)

3.02UOI

3.60e43

I.IIE+OI

l.lOE40

&%E-(W

l.llE-06

7.40EW

3.03E*a)

I.27E*M

6.%E*Oo

8.26EtOI

4.338-06

211E-w

7.92E-06

I . 8 I W

9.WE-01

5.86E+OI

2.29E-

2.68E+Ol

t.ZE+OI

Cmebay7 (CiMTHM)

l.lle*02

2 . 1 6 i m

8.41Wl

5.398403 b

I .OSE-02

6.(106-11

I .73EOf

5.55E-01

9.S(#*02

3.16MI

l.88E+Ol

4.71E-M

1.0913.07 _ I _ _ - - - - -

5.1 1 E m

2.45E-04 _I_-----------

6.39E-03

4.05E-01 .

5.49E-01

3.068-0I

3.34E*00

Category8 (CVMTHM)

I.6E403

4 . 3 ~ ~ ~ 0

7.71E*a)

O.(IoFAIO

0.OOlMO

9.90E-OS

I.3X-01

5.99E-Ol

9.I8EtO2

282E-01

I.SIE+OI

5.6sE-01

3.U7Em

1.03E-01

1.4OE+O2

l.l4E+OI

4.4X-02

4.33841

1 .WE43

2.198+0l

C a t e m 9 (CVMTHM)

6.63E42

1.4a8+01

1.16e+Ol

l.OmO3

1.33842

3.37E65

C60Un

4.86E-01

9.1-

4.49E-01

1.46EtOl

247E-01

4.76E.03

9.06E-02

9.49901

1.37@+01

2:448-01

6.168-01

255E.03

2.31E40

C a t e m 4 (CVMTHM)

3.51E403

3.117~42

3.78Wm

3.mE*o4

1.91E100

3.IlE-07

2.70E-84

3.94E-01

1.13E+03

4.98E-01

1 . 3 7 ~ 1

1.03E-M

3.5SE-05

2.8Se-04

3.IIE-01

2.98E-01

3.37842

2.82E-01

2.64E-01

1.8SE40

C I ~ & S (CilMTHM)

C I I W

1.24-

4.36@+01

4 . 4 7 W

3.69802

1 .76~-07

7.7ltEoS

3.01 8 6 1

3.67E-

298E-01

1.01E+O1

2.24804

259e-M

8.2OE5OS

8.36802

1.0(#.01

1.18€-01

3.33E-01

3.15E-01

1.33E*u)

Cdegory10 (CiTHM)

2328*03

1 . 2 6 W

l.69E*01

9.mE+CM

2 . 9 1 W

IJ2E-07

3.l7Gll

0.-

9.18E+O3

O.a?E*a)

O.OOE*a)

3.87E-m

3.34E-03

5.UE-I I

7.I5E4M

2.jOE-01

3.ME-04

6.838-02

7.I5E-13

O.OOE*a)

Category12 (CVMTHM)

3.94EMXI

2.55E-01

l.lGO1.

3.47E+OI

3 . 2 7 ~ ~ 4

6.43E-0)

I.OIE-0I

3.51E-01

1.29@+02 .

3.94E-01

3.27EUlO

26oE-01

9.83E43

1.20E-01

1.69E#2

1.39EKQ

5.64E.W

1.16E.M

1.DEM

8.21E-01

Category II (CVMTHM)

2 3 4 W

I . l 4 W

9.~t~8+03

A R W

2 3 W 1

I.53@-07

6.6lE10 '

4.49e-02

' 8.&?@43

5.DE-02

1 .%E+00

227E-08

3.40E-03

1 .WE49

1.600-(H

2.~e-01

3.718-03

9.16E-02

3.338-02

2.1 18-01

+

C*epy13 (CVMTHM)

1 . M W

2.286+02

8.s7a+Oyt.;.::

6.5se+m*

I . P E M

1.138-I0

1.76Fl-07

6.46E-01

1.20@403

5.9uE-01

2.11m0l

5.74E.08

1.92E-07 .'

s . I ~ E ~ &

2 . 8 1 ~ 4 n S

I.03842

4.94E-01

6.63E-01

2.918-01

3.89E40

Table 3.2-2 Radionuclide Inventory Used for Commercial Spent Fuel and High-Level Waste

Assumes 40,747 MTHM PWR with a burnup of 42,300 A4Wd/MTHM, and 22,253 MTHM BWR with a burnup of 32,250 MWd/MTHM and an age of 30 years at the time of disposal.

Chlorine inventory assumed to be non-gaseous release. * Doeinot exin or is present in insignificant amounts.

HLW Glass Inventory at 30 yrs

(Ci/MTHM)

2.59E00

1.8 1E00

8.10E41

2.75E-03

5.13E-08

*

5.02E-02 * *

1.81EOO

8.26B-06

6.78B-06

5.74E-05

3.19B-04

2.74E-02

4.34E-05

2.38E-04

2.07B-03

3.83E-01

Isotope

23PpU

2%

241Pu

24%

2 2 6 ~ a

= ~ a

79Se

'51Sm

I2Sn

T c

23"Th

232~h

233U

mu 235U

"6v

"*U

9 3 ~

HLW Glass Inventory at 30 yrs

(Ci/MTHM)

3.29E-04

4.73E41

1.13E-02

2.01E-02

* *

6.24EOO

3.09E-05

3.5OE-06

6.29E-02

1 .ME-%

3.00E-01

1.65E-05

1.48E-02

*

1.55E-02

5.33E-04

1.49E-08

* 2.19E42

Isotope

m~~

%'Am

2 4 W ~ m

2 4 3 ~ m

14C

3 6 ~ 1 2

W r n

245Cm

246Cm

lXCs

'q 93MNb

94Nb

5%i .

63Ni

"'NP

"lpa

21"Pti

'Ofpd

"*h

Commercial Fuel Inventory

at 30 years (CYMTHh4)

3.66Ei-02

5.40E42

3.48E+04

2.06EOO

2.57E-06

3.18E-10

4.53E-01

3.62E+02

8.73E-01

1.44Ei-O 1

3.63E-07

3.69E-04

4.47E- 10

7.2OE-05

1.38E00

1.73E-02

2.79E-01

3.1 5E-0 1

2.44EOO

Commercial Fuel Inventory

at 30 years ( C i '

1 .WE-05

3.83E+03

2.22EMl

2.53E41

1.42EOO

1.14E-02

1.19E+03

3.45E-01

7.14E-02

5.27E-01

3.52E-02

1.87E00

8.46E-01

- - 2.42EOO -

3.18E+02

4.47E-0 1

3.39E-05

6.93E-07

1.29E-0 1

3. f 3E+03

Table 3.3-1 Number of Waste Packages of Each Type and Total Number of Packages Used for Disposal of DOE SNF'

- I Information supplied by the National Spent Nuclear Fuel Program at INEEL (see Appendix A). A 3 x 1 package contains thee canisters of HLW and one canister of spent fuel (Figure 3.3-1

without the central canister). A short package has an external length of 3.87 meters. A long package has an external length of 5.25 meters. ' A 4 x 1 package contains four canisters of HLW and one canister of spent fuel (Figurt 3.3-1). A 5 x 1 package contains five anisten of HLW and one canister of spent fuel (Figure 3.3-2). ' A 0 x 4 package contains no HLW and four canisters of spent fuel (Figure 3.3-1 without the central &ster). 'The number of waste packages for this study is based on the assumption of 14.4 kg of fissile materid per package. It is likely to change based on planned criticality analyses of DOE SNF.

Table 3.4- 1 Physical PmKrties of DOE SNF'

-

9 ' Information supplied by the National Spent Nuclear Fuel Program at INEEL (see Appendix A). * Dry oxic dissolution rate for all spent he1 categories is assumed to be the same as commercial spent fuel. No credit is taken for'cladding as a barrier to release in the analyses of DOE SNF.

+

Cl-g Failure3

(%I

100

10

E 0

Commercial

100

100

100

1

35

10

10

30

0

Free Inventory

(%I

0.1

0.001

0.001

Commercial.

0.0 1

0.0 1

0.01

Gap Inventory

(%I 0

0

0

Commercial

0

0

0

Surface Area

(m2/pksI

264.41

0.26

5.4

28.54

861.33

------- 40.6

------- 171.31

153.9

0.95

1.9

2.0

79.2

37.3

Surface Area

(m2/g)

1 . 4 6 ~ l e5

5 .18~10 '~

6.80 x

5.1 8 x 105

5 .10~10 '~

1.30 x 1W3

1 . 3 0 ~ 10"

3.40 x 1 W3

5.18 x 10"

5.1 8 x 10"

5.1 8 x 10"

1 .10~ 1V

1.90 x 1 C3

Category

1

. 2

3

4

5

6

7 -

8

9

10

11

12

13

Wet Oxic iss solution^

Metallic fuel

Metallic fuel

Metallic fuel x 10

Commercial fbel

Commercial fuel

Metallic fbel

Metallic fuelxO.1

Carbide fuel

Carbide fbel

Carbide fuel

dommercial fuel .

Ceramic fuel

Commercial fuel x 0.1

0.001

0.004

0.001

0.001

0

0.003

0

0

Commercial

Commercial

Commercial

Commercial

i

4. DOSE AT THE ACCESSUBLE ENVIROh'MENT

The EPA regulation, 40 CFR 191, requires the calculation of both dose and release at the five kilometer boundary called the accessible environment. However, the EPA is currently developing a new standard based on the recommendations of the National Academy of Sciences (NAS, 1995) which will likely be based on dose, and could require calculation of dose to an average member of a critical group at locations beyond the current definition of the accessible environment. Dose to a critical group is less consexvative than dose to the maximally exposed individual because members .

of the critical group may have different lifestyles and locations. The calculations that follow are for annual dose to an individual at the accessible environment (5,000 meters from the edge of the repository) based on drinking two liters of ground water per day, and are consistent with the assumptions used in the latest potential Yucca Mountain repository performance assessment (TSPA- 1995, Andrews et al., 1995). The results could be extended beyond the accessible environment (as defrned currently) by decreasing the dose by an appropriate dilution fxtor. This dilution factor would account' for the dilution caused by dispersion and mixing that occurs between the accessible environment and the desired location.

For the most part, results of the analyses of DOE SNF ate presented as dose histories for an individual (drinking two liters of water per day) at the accessible environment over a period of 100,000 years. As presented earlier, the reduction in the solubility of Neptunium cauies the peaks to occur earlier because Neptunium is no longer the radionuclide that produces the peak dose from the potential repository. The peak dose is fiom 'q and % at about 20,000 years. The change in the solubility of Neptunium causes its peak to occur before 100,000 years. For some spent fuel categories, 1,000,000 year analyses are presented for comparison with earlier work on INEEL spent fuels (Duguid et al., 1997). The results presented in this Chapter were obtained assuming that the DOE SNF and co-disposed HLW are disposed in packages that are all in waste package failure Group 1 (Figure 2.6-2). This was done because of the difficulty of distributing small numbers of packages evenly among the eight waste package failure groups. The effect of distributing the packages across all eight waste package failure groups is presented in the sensitivity analyses of Chapter 5. However, for the analyses of commercial spent fuel it is assumed that the packages are distributed evenly among the eight waste package fhilure groups. Section 4: 1 contains the analyses for each spent fuel category, Section 4.2 describes the analyses of the repository and the composite of all DOE SNF categories, and Section 4.3 present a comparison of the results.

4.1 DOSE ATTRIBUTED TO I N D I V I D U ~ WASTE CATEGORIES

The dose history at the accessible environment largely reflects the 63,000 MTHM of commercial spent fuel in the repository (i.e., the dose fiom the small amount of waste in each category of DOE SNF is masked by the large amount of commercial spent fuel). For this reason the dose attributed to each'category of spent fuel and HLW that is co-disposed with it is evaluated separately to determine the effects of the specific category of spent fuel. As stated earlier the spent fuel and the HLW are tracked separately in the model. This allows the effect of the HLW glass to be removed so that the spent fuel in the categoPjl can be compared to an equivalent amount of commercial spent

4.1.1 Uranium Metal Spent Fuel (Category 1) .

The dose history at the accessible environment over 100,000 years attributed to Category 1 DOE SNF and co-disposed HLW is presented in Figure 4.1-1. Category 1 contains 2,136.8 MTHM of metallic fuel (Table 3.1-2) and N Reactor is the typical fuel (Table 3.1-1). It is assumed to be d i i s e d in 11 8 packages of which 13 are co-disposal packages (Table 3.3-1). Figure 4.1-2 shows the same analysis in which the dose attributed to HLW has been removed. Figures 4.1-1 and 4.1-2 are nearly identical because of the small amount of HLW used for co-disposal of this fuel category. However, there is a small difference in the dose caused by the actinides (Thorium and Neptunium) on the two figures that is caused by the presence of these actinides in the HLW. Figure 4.1-3 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 1 spent fuel (2,136.8 MTHM). The dose from the equivalent amount of commercial spent I k l is somewhat higher than b m Category 1 (Compare Figure 4.1-3 and 4.1-2). The reason for this is that the DOE spent fuel in Category 1 is of lower burnup than average commercial spent fuel resulting in a smaller inventory of Technetium and Iodine. It is aIso interesting to note that 36C1 contributes to the dose fiom commercial spent fuel and not fiom the metallic spent fuel. The radionuclide inventory of Category 1 fuel (Table 3.2-1) shows that there is not a significant amount of 36C1 in the metallic spent fuel.

4.13 Uranium-Zirconium Alloy Spent Fuel (Category 2)

The dose history at the accessible environment over 100,000 years attributed to Category 2 DOE SNF and co-disposed HLW is presented in Figure 4.1-4. Category 2 contains only 0.04 MTHM of Uranium Zirconium Alloy he1 (Table 3.1-2) and Heavy Water Component Test Reactor (HWCTR) is the typical fuel (Table 3.1-1). It is assumed to be disposed in nine packages ail of which are co- disposal packages (Table 3.3-1). Figure 4.1-5 shows the dose fiom same packages with the effects of the HLW removed. Here it can be observed that the dose from ?c changes while the '*q does not. This is because there is very little ''9 contained in the HLW, but a significant amount of v c . Also, the amount of 23'Np is reduced due to the reduction of the amount of "'Am and 2 4 ' P ~ (parents of 237Np) that are contained in the HLW.

Figure 4.1-6 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 2 spent fuel (0.04 MTHM). The peak dose fiom the equivalent amount of commercial spent fuel is about the same as that fiom Category 2 (Compare Figure 4.1-5 and 4.1-6) though the amval of the peak for commercial spent fuel is later than that for the metallic spent fuel.

Another comparison that should be made at this point is between F i p s 4.1-3 and 4.1-6, These figures are for different amounts of commercial spent hel, 2136.8 and 0.04 MTHM, respectively (Table 3.1-2). By using a scaling factor of approximately 53,000 (2136.8 / 0.04) the values of peak dose for v c and '?I can be scaled fiom Figure 4.1-3 to obtain their approximate value on Figure 4.1-6. The peak dose for 2%Jp between the two figures does not follow this scaling procedure. This

is because 99Tc and '3 are alteration controlled and is solubility controlled. Neptunium is released in proportion to the amount of water present to dissolve it. To test this, simulations were conducted in which the solubility of Neptunium was increased so that it behaved like an alteration controlled radionuclide, and it was observed that the Neptunium peak dose could then be scaled based on total MTHM fiom one result to another (results not shown).

4.13 Uranium-Molybdenum Alloy Spent Fuel (Category 3)

The dose history at the accessible environment over 100,000 years attributed to Category 3 DOE SNF and co-disposed HLW is presented in Figure 4.1-7. Category 3 contains 3.93 MTHM of Uranium-Molybdenum Alloy spent fuel (Table 3.1-2) and spent fuel from the FERMI reactor is the typical (Table 3.1-1). It is disposed in 55 packages all of which are co -d i sa l packages (Table 3.3- 1). Figure 4.1-8 shows the same analysis in .which the dose fiom the HLW has been removed. Here, as in the case of the Uranium-Zirconium spent fuel analysis, the amount of T c changes while the '291 does not. This is because there is little Iodine contained in the HLW. Also, the amount of Neptunium is reduced which is due to the amount of 24'Am and 24'Pu (parents of Neptunium) that are contained in the HLW. Figure 4.1-9 shows the dose history at the accessible environment for an amount of commercial spent he1 equivalent to Category 3 spent fuel (3.93 MTHM). The dose fiom the equivalent amount of commercial spent fuel is less than a factor of five lower than that fiom Category 3 spent fuel (Compare Figure 4.1-8 and 4.1-9).

4.1.4 Uranium Oxide Spent Fuel (Category 4)

The dose history at the accessible environment over 100,000 years attributed to Category 4 DOE SNF and co-disposed HLW is presented in Figure 4.1-10. Category 4 contab 97.7 MTHM of Uranium oxide fuel (Table 3.1-2) and Pressurized Water Reactor fuel is typical (Table 3.1-1, i.e., it is assumed to be the same as commercial spent fuel). It is disposed in 203 packages of which 175 are co-disposal packages (Table 3.3-1). Figure 4.1-1 1 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of previous categories of spent fuel, the amount of % changes while the l2Q does not. This is because there is very little I2q contained in the HLW. Also, the amount of 23'Np is reduced which is caused by 2 4 ' ~ r n and 2 4 1 ~ (parents of ='Np) that are contained in the HLW. Figure 4.1-12 shows the dose history at the accessible environment for an amount of commercial spent fuel quivalent to Category 4 spent fuel (97.7 MTHM). The dose at the accessible environment for Category 4 spent he1 is somewhat higher than that for-an equivalent amo* of commercial spent fuel primarily due .to the smaller surface area .assumed for Category 4. See Section 5.6 for analyses of the effect of surface area

Comparison of Figures 4.1-1 1 and 4.1 -12 shows a slightly higher peak dose on Figure 4.1 -1 1 for the same amount of commercial spent fuel and the sazne dissolution model. This was caused by placing the spent fuel in Category 4 in seven package groups in RIP to obtain the results for Figure 4.1 - 1 1 (i.e., the seven waste package types, Table 3.3-1) and by placing the spent fuel packages of the comercia1 equivalent'in one group (within the RIP) to obtain the results for Figure 4.1-12. Thus, it is believed that the difference in the dose peaks is an &act of the way the two results were produced.

4.1.5 Uranium oxide-~is 'm~ted Clad Spent Fuel (Category 5)

The dose history at the accessible environment over 100,000 years attributed to Category 5 DOE SNF and co-disposed HLW is presented in Figure 4.1-13. Category 5 contains 87.93 MTHM of Uranium oxide fuel with failed and disrupted clad (Table 3.1-2), and Three Mile Island (TMI) spent fuel is typical for this category (Table 3.1-1). It is disposed in 595 packages of which 509 are co- disposal packages (Table 3.3-1). Figure 4.1-14 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of previous categories of spent fuel, the amount of % changes while the I2q does not. This is because there is very little I2?I contained in the HLW. .Also, the amount of Neptunium is reduced. This reduction is caused by 241~rn and "'Pu (parents of Neptunium) contained in the HLW. Figure 4.1-15 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 5 spent fuel (85.93 MTHM). The dose at the accessible environment for Category 5 spent he1 is somewhat lower than an equivalent amount of commercial spent fuel. This is due to the lower amount of V c in the Category 5 spent he1 as compared to that of commercial spent he1 (Table 3.2-1 and 3.2-2). Also, note that the Iz9I dose for commercial spent fuel is somewhat higher than Category 5 spent fuel which would be anticipated fiom comparison of the two inventories (Table 3.2-1 and 3.2-2).

4.1.6 Uranium-Aluminum Alloy Spent Fuel (Category 6)

The dose history at the accessible environment over 100,000 years attributed to Category 6 DOE SNF and co-disposed HLW is presented in Figure 4.1-16. Category 6 contains 8.96 MTHM of Uranium-Aluminum alloy fuel (Table 3.1-2) and Advanced Test Reactor (ATR) fuel is typical (Table 3.1-1). It is disposed in 750 packages all of which are co-disposal packages (Table 3.3-1). Figure 4.1-17 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of previous categories of spent fixel, the amount of T c changes while the 12? does not. This is because there is very little I2q contained in tho HLW. Also, the amount of ='Np is reduced which is due to the amount of "'Am and 241Pu (parents of W p ) that are contained in the HLW. The peak dose is reduced somewhat when the HLW is removed. This is due to the large amount of glass used in the codisposal of this category of spent fuel (i.e., all packages are 5 x 1 packages, Table 3.3- 1). Figure 4.1 - 18 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 6 spent fuel (8.96 MTHM) is about the same (compare Figures 4.1-1 8 and 4.1-1 7). The difference in the shape of the early portion of the dose history c u m for Category 6 spent fuel (Figure 4.1.17) as compared to commercial fuel (Figure 4.1-1 8) is due to the fact that the Category 6 fuel contains considerably more ?c and '7 than commercial spent fucl (Tables 3.2-1 and 3.2-2), but the peak dose is essentially the same for 100,000 years.

4.1.7 Uranium Silicide Spent Fuel (Category 7)

The dose history at the accessible environment over 100,000 years attributed to Category 7 DOE SNF and co-disposed HLW is presented in Figure 4.1 -1 9. Category 7 contains 1 1.4 MTHM of Uranium silicide fuel'(Table 3.1-2) and Foreign Research Reactor-Materials Test Reactor (FRR MTR) fuel is typical (Table 3.1-1). It is disposed in 225 packages all of which are co-diiposal packages (Table 3.3-1). Figure 4.1-20 shows the same analysis in which the effects of the HLW

have been removed. Here the amount of ?c changes by approximately a factor of four while the '3 changes very little. This is because 86% of the T c is contributed by the HLW, and there is very little '?I contained in the HLW. Also, the amount of 23'Np is reduced which is due to the amount of 241~m and 24'Pu (parents of '?Np) that are contained in the HLW. Figure 4.1-2 1 shows the dose history at the accessible environment for an amount of commerrcial spent fuel equivalent to Category 7 spent fuel (1 1.4 MTHh4) is about the same (compare Figures 4.1-20 and 4.1-21). Peak doses fiom both T c and '9 are lower for commercial spent fbel, but 23wp is higher. The total dose peak from the commercial spent fuel, although about the sarne as that fiom Category 7 spent fuel, occurs much later (i.e., after 60,000 years) because it is fiom Neptunium and the peak is delayed because of retardation.

4.1.8 High-Integrity Uranium-Thorium Carbide Spent Fuel (Category 8)

The dose history at the accessible environment over 100,OQO years attributed to Category 8 DOE SNF and co-disposed HLW is presented in Figure 4.1-22. Category 8 contains 24.67 MTHM of Uranium-Thorium carbide spent fuel (Table 3.1-2) and Fort St. Vrain fuel is typical (Table 3.1-1). It is disposed in 545 packages all of which are co-disposal packages (Table 3.3-1). Figure 4.1-23 shows the sarne analysis in which the effects of the HLW have been removed. Here, as in the case of previous categories of spent fuel, the amount of % changes while the ' 2 9 ~ does not. This is because there is very little 12')1 contained in the HLW. Also, the amount of =?4p is reduced which is due to the amount of 241Am and 24 'P~ (parents of 23?4p) that are contained in the HLW. Even though there is about 300 times more 24'pU and about 50 times more 24'Am in the HLW than in the fuel, the Y'Np changes very little because its release is solubility limited. Figure 4.1-24 shows that the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 8 spent fitel (24.67 MTHM) is somewhat higher (compare Figures 4.1-23 and 4.1-24). Although there is only about 5% mox ?'c in commercial spent fuel than in the carbide fuel the dose is higher by more than an order of magnitude. This is because the dissolution rate of commercial fuel is more than three orders of magnitude higher than that of the carbide fuel (Figure 3.5-1).

-.

Prior analysesof INEEL spent fuel @uguid et al., 1997) indicated that other actinides contributed to the total dose, i.e., 229Th, a3U, ''Vb). For comparison to this prior work a dose history plot for a longer time period (1,000,000 years) is provided in Figure 4.1-25. The actinides reach the accessible environment at a later time because of retardation. For comparison with the prior radionuclide inventory used for Category 8 spent fie1 (called PA-5 in Duguid et al., 1997) a dose history plot is provided using the previous inventory (Figure 4.1-26). The diierence between Figures 4.1-25 md 4.1-26 is due to a difference in the number of waste packages, while the difference between Figure 4.1-26 and prior plots (Duguid et al., 1997) is in the model used (i.e., higher flux through the mountain, cathodic protection, and waste package failure rate). The Category 8 spent fixel is contained in 545 waste packages (Figure 4.1-25) and the spent fuel for Figure 4.1-26 is contained in 63 waste packages. All of the radionuclides shown in both figures are solubility limited, and the more packages diffksing these radionuclides at their solubility limit the higher their dose peaks (i.e., the approximate order of magnitude increase in waste packages translates into m approximate order of magnitude increase in dose).

4.1.9 Low-Integrity Uranium-Thorium Carbide Spent Fuel (Category 9)

The dose history at the accessible environment over 100,000 years attributed to Category 9 DOE SNF and co-disposed HLW is presented in Figure 4.1-27. Category 9 contains 1.66 MTHM of low integrity Uranium-Thorium carbide spent fire1 (Table 3.1-2), and Peach Bottom fuel is typical (Table 3.1-1). It is disposed in 103 packages all of which are co-disposal packages (Table 3.3:l). Figure 4.1-28 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of previous categories of spent fuel, the amount of v c changes by more than two orders of magnitude while the '9 changes very little. The change in ?c is due to removal of the inventory . in the glass, and the lack of change in Iodine is because there is very little '?I contained in the HLW. Also, the amount of Neptunium is reduced which is due to the amount of 2 4 ' ~ r n and 241Pu (parents of 23'Np) that are contained in the HLW. Category 9 spent fuel is co-disposed in 5 x 1 packages which accounts for the large reduction in Neptunium when the effect of the HLW glass is removed. Figure 4.1-29 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 9 spent fuel(1.66 MTHM) is more than two orders of magnitude . higher (compare Figures 4.1-28 and 4.1 -29). The difference is attributed to the slower dissolution rate of the carbide spent fuel (Figure 3.5-1) and to the smaller surface area of the spent fuel (Table 3.4-1). See Section 5.6 for an analysis of the effect of surface area. Comparing among the carbide fuels of Categories 8 and 9 (Figures 4.1.23,4.124,4.1.26, and 4.1 27) it is apparent that surface area is having a considerable effect on dissolution (both categories use the carbide dissolution model). The sensitivity analyses in Chapter 5 confirm this.

4.1.10 Uranium and Uranium-Plutonium Carbide Spent Fuel (Category 10)

The dose history at the accessible environment over 100,000 years attributed to Category 10 DOE SNF and codisposed HLW is presented in Figure 4.1-30. Category 10 contains 0.22 MTHM of Uranium and Uranium-Plutonium spent he1 (Table 3.1-2), and Fast Flux Test Facility (FFTF) carbide fuel is typical (Table 3.1-1). It is disposed in 5 packages all of which are co-disposal packages (Table 3.3-1). Figure 4.1-3 1 shows the same analysis in which the effects of the HLW have been removed. Here T c and 1291 disappear because there is none in the spent fuel, and the peak dose is from 23'Np. Category 10 spent fuel is co-disposed in 4 x 1 packages, and the removal of the dose attributed to the glass lowers the dose h m Neptunium. This occurs because of two factors, the removal of inventory in the glass and the faster dissolution rate of the glass as compared to the carbide spent fuel (Figure 3.5-1). Figure 4.1-32 shows that the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 10 spent fuel (0.22 MTHM) is approximately a factor of five higher (compare Figures 4.1-3 1 and 4.1-32). The difference is attributed to the slower dissolution rate of the carbide spent fuel (Figure 3.5-1) and to the smaller surface area of the carbide spent fuel (Table 3.4-1). Sensitivity of dose to dissolution model is presented in Section 5.4, and the sensitivity of dose to surface area is presented in Section 5.6.

. 4.1.11 Mixed Oxide spent Fuel (Category 11)

The dose history at the accessible environment over 100,000 years attributed to Category 1 1 DOE

SNF and co-disposed HLW is presented in Figure 4.1-33. Category 1 1 contains 1 1.75 MTHM of mixed oxide (MOX) spent fie1 (Table 3.102)~ and Fast Flux Test Facility (FFTF) oxide he1 is typical (Table 3.1-1). It is disposed in 352 packages all of which are co-disposal packages (Table 3.3-1). Figure 4.1-34 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of many of th previous categories of spent &I, the peak dose of *Tc changes by approximately a fsctor of 1.5 while the 9 changes very little. The change in *Tc dose is due to removal of the inventory in the glass, and the lack of change in Iodine dose is because there is very little '9 contained in the HLW glass that was removed. Also, the amount of z'Np is reduced which is due to the amount of 241~m and "'Pu (parents of ?Np) that are contained in the HLW. Category 11 spent fiiel is codisposed in 4 x 1 and 5 x 1 packages which accounts for the reduction in Neptunium. Figure 4.1-35 shows that the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 1 1 spent fuel (1 1.75 MTHM). The dose for commercial spent fuel is about the same as that for the MOX spent fuel because the same dissolution rate was used for both fuels (Table 3.4-1) and the radionuclide inventories are quite similar.

4.1.12 Uranium-Thorium Oxide Spent Fuel (Category 12)

The dose history at the accessible environment over 100,000 years attributed to Category 12 DOE SNF and co-disposed HLW is presented in Figure 4.1-36. Category 12 contains 49.66 MTHM of Uranium-Thorium oxide spent fuel (Table 3.1 -2), and the Shippingport light water breeder reactor GWBR) fuel is typical (Table 3.1-1). It is disposed in 69 packages all of whkh are co-disposal packages (Table 3.3-1). Figure 4.1-37 shows the same analysis in which the effects of the HLW have been removed. Ha, as in the case of many of the previous categories of spent fuel, the peak dose of q c changes by more than an order of xnagnitude while the 'q changes very little. The change in ?c dose is due to removal of the inventory in the glass, and the lack of change in Iodine dose is because there is very little '3 contained in the HLW. Also, the amount of a'Np is reduced which is due to the amount of " '~rn and 24'Pu (parents of ='Np) that are contained in the HLW. Category 12 spent fuel is co-disposed in 3 x 1 packages which accounts for the reduction in Neptunium. Figure 4.1-38 shows the dose history at the accessible environment for an mount of commercial spent fuel equivalent to Category 12 spent fuel (49.66 MTHM). The dose for commercial spent fuel is more than rn order of magnitude higher than Category 12 spent fuel because the ceramic dissolution rate (Table 3.4-1) is more than two orders of magnitude less'than that of commercial spent fuel and the radionuclide inventory of the Category 12 spent he1 is lower (Tables 3.2-1 and 3.2-2).

4.1.13 Uranium-Zirconium Hydride Spent Fuel (Category 13)

The dose history at the accessible environment over 100,000 years attributed to Category 13 DOE SNF and co-disposed HLW is presented in Figure 4.1-39. Category 13 contains 1.99 MTHM of. Uranium-Zirconium spent fuel (Table 3.1-2), and the fuel fiom the Training Research Isotopes- General Atomics (TRIGA) reactor is typical (Table 3.1-1). It is disposed in 102 packages of which 101 are co-disposal packages (Table 3.3-1). Figure 4.1-40 shows the same analysis in which the effects of the HLW have been removed. Here, as in the case of many of the previous categories of

spent fuel, the % dose decreases while the 12? changes very little. The decrease in Technetium is due to removal of the inventory in the glass and the lack of change in Iodine is because there is very little '9 contained in the HLW. Also, the a'Np dose is reduced which is due to the inventory of 24'Am and 241Pu (parents of 23'Np) that are contained in the HLW. Category 13 spent fuel is co- disposed in 4 x 1 and 5 x 1 packages (Table 3.3-1) which accounts for the reduction in Neptunium. Figure 4.1-41 shows the dose history at the accessible environment for an amount of commercial spent fuel equivalent to Category 13 spent fuel (1.99 MTHM). The dose for commercial spent fuel .

is somewhat higher than for the hydride spent he1 because of the lower dissolution rate for the hydride (0.1 x commercial spent fuel, Table 3.4-1) and because it contains less ='Np.

4 3 DOSE ATTRIBUTED TO COMBINED WASTES

The expected-value dose history attributed the amount of co-disposed HLW (8,745 MTHM) used for disposal of the high-enriched and medium-enriched DOE SNF is shown in Figure 4.2- 1. These doses may be compared to the amount of HLW assumed for the base case which is 7,000 MTHM (Figure 4.2-2). In both analyses (8,745 and 7,000 MTHM) the waste was distributed evenly among eight waste package failure groups (Figure 2.6-2). The base case for this study is for a 70,000 MTHM repository that contains 63,000 MTHM of commercial spent fbel and 7,000 MTHM of HLW. The peak dose for the two cases is not significantly different.

The expected-value dose history at the accessible environment over 100,000 years attributed to 63,000 MTHM of commercial spent fuel is presented in Figure 4.2-3. In this case the commercial spent fuel is distributed evenly among eight waste package failure groups (Figure 2.6-2). The peak dose shown in Figure 4.2-3 would be expected to be nearly identical to that fiom the base c&e repository because the 7,000 MTHM of HLW (Figure 4.2-2) produces peak dose that is approximately two orders of magnitude lower than that fiom commercial spent fuel (Figure 4.2-2).

The expected-value dose history for all 13 categories of DOE SNF (2,436 MTHM) is shown in Figure 4.2-4 .where it is assumed that all of the DOE SNF is in waste package failure Group 1 (Figure 2.6-2). -This is compared to the dose history fiom an equivalent amount (2,436 MTHM) of commercial spent fuel in the same waste package failure group (Group 1) which is shown in Figure 4.2-5. Comparison of the two figures indicates that the peak dose for commercial spent b l is slightly higher than that fiom the combined DOE SNF. Additionally, differences are noted in the top six dose producing radionuclides for the commercial and DOE spent fuel. For the DOE spent fuel 233U is among the top six radionuclides while for commqrcial spent he1 3 6 ~ 1 is among the top .six dose producing radionuclides. Figure 4.2-6 shows the dose history for a repository containing 63,000 MTHM of commercial spent fbel, 2,436 MTHM of DOE SNF, and 8,745 MTHM of HLW with the commercial spent fuel distributed equally among the eight waste package failure groups, and the DOE he1 h d co-disposed HLW assumed to be in waste package failure Group 1. The sensitivity of dose to the assumption of waste package failure group is presented in Section 5.7.

Figure 4.2-7 shows the dose history at the accessible environment for the base case repository (63,000 MTHM of commercial and 7,000 MTHM of HLW). The peak dose &om the repository containing a composite of the 13 categories of DOE SNF is higher than that of the base case. This

is caused by placing all of the DOE SNF in waste package failure Group 1. This w8s done because of the difficulty of distributing small numbers of packages evenly among eight waste package failure groups (i.e., Categories 2 and 10).

Figure 4.2-8 shows the dose history from a repository containing 63,000 MTHM of commercial spent fuel, 2,333 MTHM of DOE SNF, and 4,667 MTHM of HLW (i.e., a 70,000 MTHM repository). For comparison, Figure 4.2-9 shows a repository that contains 60,667 MTHM of commercial spent fuel distributed among eight waste package failure groups, 2,333 MTHM of commercial spent fuel in failure Group 1, and 7,000 MTHM of HLW in failure Group 1 (i.e. the base case with 2,333 MTHM of commercial spent fuel in waste package failure Group 1). The assumption of failure group for the HLW has no significant influence on the dose h m ,the repository because of its relatively small radionuclide inventory (see Figures 4.2-1 and 4.2-2). Comparison of Figures 4.2-8 and 4.2-9 indicates that the dose from both repository loadings are nearly the same, and that as mentioned above the difference between the base case and Figure 4.2-9 (or 4.2-6) is due to placing the 2,433 MTHM of commercial spent fuel in the failure Group 1 (or in the case of Figure 4.2-6 the 2,436 M M of DOE SNF in failure Group 1).

Two additional analyses were conducted to examine the long-term peak dose h r n t29Th at the accessible environment. These analyses consisted of 1,000,000 year dose histories for the base case and for the 74,181 MTHh4 repository. The Thorium peak for both analyses appeared to be identical, consequently the results are not shown.

4 3 COMPARISON OF RESULTS

The average amount of DOE SNF per package (MTHM/package) is shown in Table 4.3-1 for the 13 categories of DOE spent fuel. The values shown in this table are irrespective of waste package type and do not include any of the codisposed HLW. For comparison a commercial spent fuel package is assumed to contain 9.74 MTHM. A comparison of dose attributed to one package of Categories 1 through 3 of DOE SNF and one package of commercial spent fuel is shown in Figure 4.3-1. The do* at the accessible environment fiom one package of metallic spent fuel is less than that fiom commercial spent fuel even though a package of Category 1 spent he1 contains nearly twice as much spent fuel on an MTHM basis. The lower dose from Category 1 spent firel is attibuted to the large amom? of N Reactor spent fuel that is low bumup (2,102 MTHM as compared to 2,136 MTHM in the Category). The dose histories at the acmssi'ble environment for one package of Category 4,5,6, and 7 are shown in Figure 2.3-2. Category 6 spent fuel is packaged with the

. second smallest amount of spent fuel in a package (0.01 MTHM, Table 4.3-1). It produces higher dose at the accessible environment per MTHM than some of the other DOE SNF categories because it is higher bumup. A comparison of the dose histories from one package of Category 8,9; and 10 DOE SNF is shown in Figure 4.3-3, and dose histories for Categories 11,12, and 13 are shown in Figure 4.3-4. On Figure 4.3-3 and 4.3-4 it can be seen that spent fuel Categories 8, 10, and 11 produce dose curves that are characteristic of Neptunium (i.e., curves that peak after 60,000 years). The peak dose at the accessible environment for Category 13 spent fuel is also fiom Neptunium (Figure 4.3-4). This peak occurs earlier (before 40,000 years) and is due to &e reIatively smaIl amount of Neptunium in the spent fuel inventory (Table 32-1) that is becoming depleted after about

40,000 years. Other DOE spent fuels (Categories 1,2,3,4,5,6,7,9, 12, and 13) have an earlier dose peak which is atiributed to Technetium (before 20,000 years), and for some of these spent fuels (Category 2,5,6,7, and 13) the highest dose peak occurs later and is fiom ~eptunium.

A comparison of the dose history at the accessible environment attributed to each Category (1 through 13) is shown in Figures 4.3-5 and 4.3.6. It is relatively easy to visualize that the dose history attributed to the composite of the categories (2,436 MTHM) would nearly follow the dose history for Category 1 to approximately 30,000 years and then neirly follow the dose history for Category 6 fiom 30,000 to 100,000 years (Figure 4.3-5 and 4.3.6). This may be compared to the dose history for the composite of DOE SWF. shown in Figure 4.2-4. The reason that these two categories (1 and 6) nearly control the dose from the composite of DOE SNF is that Category 1 contains 2,136.8 MTHM, and Category 6, even though a relatively small amount (8.96 MTHM) is a high burnup which produces more transuranics in its radionuclide inventory (i.e., Neptunium and its parents).

Comparison of Figures 4.1 - 12 and 4.1 - 15 that are for nearly the same amount of commercial spent he1 (97.07 and 87.93 MTHM, respectively) shows the effects of an increase in the number of waste packages (203 and 595, respectively). The higher Neptunium peak on Figure 4.1-1 5 is due to the larger number of failed waste packages releasing Neptunium at its solubility limit. The more rapid decline of the alteration controlled radionuclides (Iodine and Technetium) on Figure 4.1 - 1 5 is caused by depletion of the smaller source of these radio~uclides in failed waste pac&ges.

Another comparison that can be made is of the shape of dose peaks for the analyses of Category 1 spent fhel-(Figures 4.1-2 and 4.1-3). Here for both Category 1 DOE SNF and the equivalent amount of commercial spent fbel the peak dose fiom Iodine and Technetium is considerably higher than the Neptunium influenced part of the total dose curve. This is because of the release of these radionuclides at the he1 alteration rate fbm the larger amount (MTHM) of spent fuel. Also, a comparison these figures to Figure 4.2-3 (for 63,000 MTHM of commercial spent fuel) shows that they are similar in shape.

Some of the plots for commercial spent fuel in this Chapter show 36 C1 as a dose producing radionuclide while others show 229Th (compare Figures 4.1-3 and 4.1-6). In this study only the top six dose producing radionuclides were plotted on the figures and this causes the difference seen among the figures. When the release scenario yields a larger amount of 23'Np in relatioilship to other radionuclides (as in Figure! 4.1-6) "9Th and U3U (daughters products of Neptunium) are among the top six radionuclides. ConverseIy, when the Neptunium release is smaller 36C1 is among the top six dose producing radionuclides (Figure 4.1-3).

0 . 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.1-1 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Category 1 Spent Fuel and HLW.

I I

Figure 4.1-2 Expected-Value Dose History at the Accessible Environment Over 100,000 Yean I from Category 1 Spent Fuel.

Figure 4.1-3 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Commercial Spent Fuel Equivalent to Category 1.

0 20,000 40,000 60,000 80,000 100.000 Time (yrs)


Figure 4.1-5 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Category 2 Spent Fuel.

I 0'

10-8 0 20,000 40,000 60,000 80,000 100,000

. Time (yrs)

Figure 4.1 -6 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Commercial Spent EueI Equivalent to Category 2.

4-13

Figure 4.1-7 Expected-Value Dose History at the Accessible Enviromt Over 100,000 Years from Category 3 Spent Fuel and HLW.

I o-' 104

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

F i p 4.1-8 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 3 Spent Fuel.

106 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)


108 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

I

Figure 4.1 - 1 0 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 4 Spent Fuel and HLW.

Figure 4.1-1 1 Expected-Value Dose History at the Accessible Environment Over 100,000 Years h m Category 4 Spent Fuel.

109 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figure 4.1 - 12 Expected-Value Dose History at thc Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 4.

4-16

Figure 4.1 - 13 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Category 5 Spent Fuel and HLW.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.1-14 Expected-Value Dose History at the Accessible Environment Over 100,000 ' Years from Category 5 Spent Fuel.

Dose (remlyr) Dose (remlyr)

g lob n

10-6

- Total

0 20,000 40,000 60,000 80,000 100,000 Time (yrs) .

Figure 4.1-1 7 Expected-Valuc Dose History at the Accessible Environment Over 1 00,000 Years fiom Category 6 Spent Fuel.

Figure 4.1 -1 8 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom Commercial Spent Fuel Equivalent to Category 6.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.1 - 1 9 Expected-Value Dose History at the Accessible Environment Over 1 00,000 Years h m Category 7 Spent Fuel and HLW.

Figure 4.1-20 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 7 Spent Fuel.

10-7

1Oa 0 20,000 40,000 60,000 80,000 100,000

lime (yrs)

- Total

Figure 4.1-21 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 7.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

I I Figure 4.1-22 Expected-Value Dose Histoxy at the Accessible Environment Over 100,000

I Years from Category 8 Spent Fuel and HLW. I


Figure 4.1-24 Expected-Value Dose History & the Accessible Environment Over 100,000 Years from Cornmercid Spent Fuel Equivalent to Category 8.

Figure 4.1-25 Expected-Value Dose History at the Accessible Environment Over 1,000,000 Years fiom Cadegory 8 Spent Fuel.

10s Oe+O 2e+5 4e+5 6e+5 8e+5 le+6

Time (yrs)

Figure 4.1-26 Expected-Value Dose History at the Accessible Environment Over 1,000,000 Years fiom Category 8 Spent Fuel Using the Radionuclide Inventory fiom Duguid et al., 1997.

. Figure 4.1-27 Expected-Value Dose, History at the Accessible Environment Over 100,000 Years h m Category 9 Spent Fuel and HLW.

Figure 4.1-28 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 9 Spent Fuel.

4-24

g lab n

106

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)


10s 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figure 4.1-30 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Category 10 Spent Fuel and HLW.

8 lo. n

106

Figure 4.1-3 1 Expected-Value Dose History at the Accessible Environment Over 100,000 Years h m Category 10 Spent Fuel.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.132 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 10.


- Total

Figure 4.1-35 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 1 1.

h L 10-2

p lo. g 104

0 20,000 40.000 60.000 80,000 100.000 Time (yrs)



- Total

0 20,000 40,000 60,000 80,000 100,000 ~ i m e (yrs)

Figure 4.1-38 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom commercial Spent Fuel Equivalent to Category 12.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.1-39 Expected-Value Dose History at the Accessible Environment Over 100,000 Years h m Category 13 Spent Fuel and HLW.

'

10-8 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figure 4.140 Expected-Valuc Dose History at the Accessible Environment Over 100,000 Years from Category 13 Spent Fuel.

F i p 4.1-4 1 Expected-Value Dose History at the Accessible Environmemt Over 100,000 Years from Commercial Spent Fuel Equivalent to Category 13.

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure 4.2-1 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom 8,745 MTHM of HLW.

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure 4.2-2 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from 7,000 MTHM of HLW.

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure 4.2-3 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from 63,000 MTHM of Commercial Spent Fuel.

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure 4.2-4 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from 2,436 MTHM of DOE SNF.

0 20,000 40,000 60,000 80,000 100,000 Time (years)

Figure 4i2-5 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from 2436 MTHM of Commercial Spent Fuel.

0 20.000 40.000 60.000 80,000 100.000 Time (yrs)

Figure 4.2-6 Expected-Value Dose History at the Accessible Environment Over 100,000 Years for a Repository Containing 63,000 MTHM of Commercial Spent Fuel, 2436 MTHM of DOE SNF, and 8,745 MTHM of HLW (74,181 MTHM).

lo-'

0 20.000 40,000 60,000 80.000 100.000 Time (yrs)

Figure 4.2-7 Expected-Value Dose Hiory at the Accessible Envimnmat Over 100,000 Years for a Repository Containing 63,000 MTHM of Commercial Spent Fuel and 7,000 MTHM of HLW (70,000 MTHhd).

4-34

Figure 4.2-8 Expected-Value Dose History at thc Accessible Environment Over 100,000 Years for a Repository Containing 63,000 MTHM of Commercial Spent Fuel and a Combined 7,000 MTHM of DOE SNF and HLW.

0 20,000 40,000 60,000 80,000 100,000 Time (yrs)

Figure 4.2-9 Expected-Value Dose History at the Accessible Environment Over 100,000 Yean for a Repository Containing 70,000 MTHM of Commmial Spent Fuel and HLW with 2,333 MTHM of Spent Fuel in Waste Package Failure Group 1 and 60,667 MTHM of Spent Fuel and 7,000 MTHM of HLW Distributed Evedy Among Eight Waste Package Failure Groups.

10-8 0 20,000 40,000 60,000 80,000 100,000

Time (yrs) .

Figure 4.3-1 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom One Package of Commercial Spent Fuel as Compared to One Package of Category 1,2, and 3 DOE SNF.

10-7

lo-" 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figure 4.3-2 Expected-Value Dose Histofy at the Accessible ~nviroment &er 100,000 Years from One Package of Category 4,5,6, and 7 DOE SNF;

Figure 4.3-3 Expected-Value Dose History at the Accessible Environment Over 100,000 Years fiom One Package of Category 8,9, and 10 DOE SNF.

g 10d lob

Figure 4.3-4 Expected-Value Dose History at the Accessible Environment Over 100,000 Years from One Package of Category 1 I , 12, and 1 3 DOE SNF.

h 10-2.

$ lo. g 10-4

Figure 4.3-5 Expected-Value Dose History at thc Accessible Environment Over 100,000 Yein - from Category 1,2,3,4,5,6, and 7 DOE SNF.

-. . A

1 w2 $ lo.

104 Y

10 DOE Sf; 0.22 MTHM

Figure 4.3-6 Expected-Value Dose History at the Accessible Environment Over 100,000 Yean h m Category 8,9,10,11,12, and 13 DOE SNF.

4-3 8

Table 4.3-1 Average MTHM of DOE SNF in Each Waste Package

Category

1

2

3

4

5

6

7

8

9

10

11

12

13

MTHMlPackage

18.1

0.004

0.07

0.5

0.1

0.01

0.05

\ 0.05

0.02

0.04

0.03

0.7

0.02

5. EVALUATION OF SENSITIVITYAJNCERTAiNTY

The sensitivity analyses presented in this Chapter were conducted to aid in the understanding of the results presented in Chapter 4, to provide a justification for some of the interpretations presented in Chapter 4, and to examine the effect of data uncertainty on selected categories of DOE SNF. The analyses include sensitivity of dose at the accessible environment to engineered banier release model, type of waste package, cladding failure, dissolution model, dissolution rate, waste form surface area, and waste package failure group.

For the sensitivity analyses that follow, the flux through the repository horizon was kept constant (i.e., the effects of climate change were not analyzed). The flux that was used in this performance assessment was abstracted fiom the results of the calibrated site-scale model for the unsaturated zone. The site-scale model has not yet been calibrated for the transient effects of climate change, and because of this the effects of a change in flux could not be analyzed. Likewise, changing the flux arbitrarily would not be appropriate because the coupled nature of these analyses of DOE SNF could yield erroneous results.

5.1 ENGINEERED BARRIER RELEASE MODEL

There are three engineered barrier release models discussed in Section 2.7.1 [see Figures 2.7-l(a), 2.7-l(b), and 2.7-l(c) that represent "drips on the waste fonn", "drips on the waste package", and "no drips" (the capilhy barrier), respectively]. The results presented in Chapter 4 are based on the assumption that 30% of the packages experience drips and 70% of the packages have no drips. or the release model "drips on the waste fom" it is assumed that once the package has failed at locations where drips occur, the advective flux of the drip is through the fiailed package. This is a very conservative assumption. For the conceptual model "drips on the waste package", at locations where drips occur, the release pathway is by diffusion out of the package (after the package fails) into the advective flux of the drip. In the "no drip" model the radionuclide release fiom the barrier system occurs due to diffusion out of the failed waste package and diffusion through the crushed tuff invert.

Figure 5.1-1 shows the dose history at the accessible environment fiom Category 1 DOE SNF for the three different engineered barrier release models. For these analyses, all of the waste packages were assumed to be in waste package failure Group 1, and drips were assumed to occur at 30% of . the package locations. For the results of the model "drips on the waste form" shown in Figure 5.1-1, '

drips contact the waste directly at 30% of the waste package locations while the remaining 70% of the waste packages have no drips (e.g., capillary barrier). Under this release assumption, the dose fiom Technetium and Iodine occurs earlier in time because there is no diffusion banier to delay the release. The peak is lower because of depletion of Technetium and Iodine from the first packages that fail in fkilure Group 1. Also, it can be seen that the dose from Neptunium is significantly higher because there is more water contacting the waste and dissolving this solubility limited radionuclide (the later portion of the curve on Figure 5.1-1). For drips on the waste package the peak anives later because of the delay of diffusion out of the failed waste package, and is somewhat higher because

there are more diffUsion sources as the packages fil. The first packages to f d are not depleted and are releasing Technetium and Iodine at the same rate as the last packages to fail under the Group 1 waste package failure assumption. The capillary barrier is assumed to be created by having a drip shield at all package locations (including the 30% of locations where drips occur). The dose curve . for this release model is a typical difbion release curve (Figure 5.1-1). Technetium, Iodine, and Neptunium are all diffusing out of the package and through the crushed tuff invert. This results in a peak dose after about 40,000 years that is equal to the dose fiom the release model "drips on the waste package". The total dose fiom a combination of Technetium, Iodine, and Neptunium contributing to the peak after 40,000 years is presented in Figure 4.1-2.

An observation that can be made is the results of the two release models drips on the waste package and the capillary barrier shown in Figure 5.1-1. This observation is that the two curves represent end points between 30% drips and no drips. Thus, as the percentage of drips on the waste package is reduced the dose peak should move downward and to the right (i.e., becomes lower and farther out in time) until it reaches the results for the capillary barrier.

Another factor that controls dose at the accessible environment in the capillary barrier release model is diffusion coefficient in the invert material. The diffusion coefficient in the invert is a function of the material used to construct the invert. Figure 5.1-2 shows the effects of choice of invert material on dose at the accessible environment for Category 1 IX)E SNF. Category 1 spent fie1 was selected for these analyses because it contains the majority of the DOE SNF. These results show that as the diffusion coefficient is reduced the dose at the accessible environment is reduced accordingly (Figure 5:l-2). The diffusion coefficient for crushed tuff, basalt gravel, and sand (in descending order) are 6.94 x 104 m2/yr, 9.87 x 1Qh2/yr, and 7.1 7 x lo*' m2/yr, respectively.

A second controlling factor for release from a capillary barrier is lifetime of the drip shield. The sensitivity of dose at the accessible environment to assumed lifetime of the drip shield is shown in Figure 5; 1-3. For these analyses all of the drip shields were assumed to fail instantaneously at the end of their lifetime, for lifetimes. ranging fiom 1 0,000 to 100,000 years. This assumption moves the peak caused by Technetium and Iodine out in time, an amount equal to the assumed lifetime of the drip shield. The peak is lowered slightly due to a small amount of decay of Technetium as it moves out in time. Had the failure of the drip shields been assumed to occur over some period of time the peak dose would be reduced and the width of the peak would be increased.

5 3 TYPE OF WASTE PACKAGE

The assumption of the type of waste package in which the spent fixel is disposed should not change the dose at the accessible environment because of the assumptions made in this assessment of DOE SNF unless an input error was made for a particular package type (i.e., all of the package types are assumed to have the same wall thickness and are assumed to contain an average amount of spent fuel). In order to check this, Category 4 DOE SNF was selected. This Category was selected because it has packages of all seven types (Table 3.3-1). The result assuming that all of the Category 4 spent fie1 is contained in two package types [(0 x 4) and (5 x I)] is compared to the Category contained in seven package types (Figure 5.2-1). All of the waste package types were

assumed to have the same thicknesses of corrosion barriers and should have the same failure rate if they are assumed to be in the same waste package failure group. For these analyses the effects of the HLW were removed. The two dose histories are identical, as expected. Had different types been compared on the basis of the actual MTHM per package, the results would differ by the amount of inventory in each package type, but the dose from the entire amount of spent fuel in the category would remain unchanged.

5 3 CLADDING FAILURE AND FREE l[NVENTORl'

The rate of cladding fhilure was not analyzed because an appropriate cladding corrosion model was not abstracted for use in RIP for the analyses of the performance DOE SNF. Thus, only the effect on dose at the accessible environment of assuming different percentages of cladding failure was analyzed. Category 1 DOE SNF was selected for these analyses, because it is the category that contains the largest amount of spent fuel (2,136.8 MTHh4). Figure 5.3-1 shows the effects of assuming loo%, SO%, lo%, and 1% cladding failure. The percentage of cladding failure was assumed to exist at the time of waste package fbilure, and did not change over the time period of the calculation. The results indicate that cladding could have a significant effect if it were intact and corroded slowly. However, the two DOE spent fuels that contribute significantly to the total dose at the accessible environment, Categories 1 and 6, can reasonably be assumed to have 100% cladding failure at the time of waste package failure. Category 1 is primarily N Reactor spent fixel which currently has a significant amount of cladding failure from being stored in the pools at Hanford. Category 6 fuel has Aluminm cladding (Appendix A) which would be expected to corrode rapidly after waste package failure. Because of this incorporation of a cladding corrosion model into the assessment would not be expected to significantly change the dose at the accessible environment fiorn the composite of DOE SNF evaluated in this study.

The effects of the assumed amount of fk radionuclide inventory was also examined. Analyses were done for O.W, 0.01%, 0,1%, and 1.0% fiee inventory for Category 1 spent hel, and the resulting dose histories at the accessible environment appear identical. This is likely due to the effects of the-change in fiee inventory being masked by solubility of the radionuclides. The . dissolution of alteration controlled radionuclides is rapid independent of their being in the h e inventory and the release of solubility Iimited radionuclides is controlled by solubility independent of their being in the fiee inventory.

5.4 DISSOLUTION MODEL

The dissolution model for some categories of DOE SNF is uncertain. Category 9 spent fuel was selected to examine the effect of different dissolution models on dose at the accessible environment. This category is the low-integrity Uranium-Thorium carbide spent fuel which is represented by Peachbottom spent fuel. Peach Bottom fuel does not have the silicon carbide coatings on the fuel particles that are used ip the manufactme of the fuel. Because of this the carbide dissolution model, which was used for the analyses, may not apply. The effect on dose at the accessible environment of assumed dissolution model is shown in Figure 5.4-1. The results indicate that the oxide and metallic dissolution models produce identical dose histories (the plots are on the same line). There

is approximately one order ifmagnitude diffncnce between assuming the oxide or metallic model and the ceramic model, and two orders of magnitude difference between assuming either of these two models and the carbide model. Here it should be observed that Category 9 spent fuel would not have a significant effect on the dose from the composite of DOE SNF irrespective of the dissolution model used (Compare Figures 5.4-1 and 4.2-1).

The dissolution rate from the four spent fuel dissolution models is different by more than four orders of magnitude (Figure 3.5-1). This difference is reduced to two orders of magnitude at the accessible environment because of the combined effects of diffusion and solubility. For example, the results fiom the oxide and metallic model are identical because more radionuclides are dissolved then can diffuse out of the package. The result is controlled by diffusion out of the waste package rather than the dissolution rate withim the waste package.

5.5 WASTE FORM DISSOLUTION RATE

Within a given dissolution model the rate of dissolution is uncertain because of the limited amount of test data available to verify the model. To examine the effects of dissolution rate on dose at the accessible environment, Category 1 DOE SNF was selected. This Category was selected because it is the significant contributor to dose at the accessible environment, and because of the relatively rapid rate of fiel failure observed in N Reactor storage pools. Figure 5.5-1 shows the effects of a range of dissolution rate on dose at the accessible environment for Category 1 spent fiel. Increasing the dissolution rate by three orders of magnitude produces no change in the dose because the release from the package is controlled by diffusion. There is more dissolved Technetium q d Iodine available at the assumed dissolution rate (or at 1,000 times the dissolution rate) than can be released from the package by *ion. Decreasing the dissolution rate by three orders of magnitude reduces the peak somewhat because the dissolution is slower than the mpoa rate out of the package by diffusion. The slower rate of removal fiom the waste package causes the peak to have a slightly longer duration. This causes the dashed curve on Figure 5.5-1 to cross the curves for the other results. -.

Because the d i h i o n pathway controls the release fiom the package for the conceptual waste package release model "drips on the waste package" the sensitivity to dissolution rate of Category 1 spent fuel was also examined for the conceptual release model "drips on the waste form" (Figure 5.5-2). The change in dissolution rate has little effect on the dose history at the accessible environment because the release is controlled by the flux through the waste package (compare Figure 5.5-2 and 5.1-1). Iodine and Technetium are released into the flux at the alteration rate and Neptunium is dissolved into the flux according to its solubility (i.e., the only significant effect is from the change in conceptual release model).

5.6 WASTE FORM SURFACE AREA

The dissolution rate ii also a function of surface area. To examine the effects of surface area on dose at the accessibIe environment, two categories of DOE SNF were selected. Categories 1 and Category 5 are represented by N Reactor spent fie1 and Three Mile Island core debris, respectively.

For both of these categories of DOE SNF the surface area is uncertain.

The volume of water in the RIP that dissolves the spent Gel is equal to the water film thickness on the fuel times the fuel sdace area. Thus, a change in the volume of water available for leaching radionuclides is directly related to the change in surface area. Decreasing the surface area increases the concentration of the peak so that the first packages to fail under waste package Group 1 are releasing at higher radionuclide concentrations. This produces the earlier higher peak shown on Figure 5.6-1 (compare the 0 . 1 ~ and x dose peaks). As surface area is increased peaks are lowered by dilution in more water volume and because the first packages that fail are becoming depleted.

'

The depletion of packages that fail early lowers the peak and delays the peak under waste package failure Group 1. As the peaks are lowered and are moved out in time, because of conservation of mass, they become broader which accounts for the curves crossing on,Figure 5.6-1. A similar ,but less dramatic result was found for Category 5 spent fbel (Figure 5.6-2).

Because the diffusion pathway controls the release from the package for the conceptual waste package release model "drips on the waste package" the sensitivity to change in surface area of Category 5 spent fuel was also examined for the conceptual release model "drips on the waste form" (Figure 5.6-3). The change in surface area has little effect on the dose history at the accessible environment because the release is controlled by the flux through the waste package. Iodine and Technetium are released into the flux at the alteration rate and Neptunium is dissolved into the flux according to its solubility (i.e., the only significant effect is fiom the change in conceptual release model).

5.7 WASTE PACKAGE FAILURE GROUP

For most of the preceding analyses the spent fbel was assumed to be contained in waste packages in waste package failure Group 1. This is a very conservative assumption and was made to avoid distributing packages in 13 categories of DOE spent fbel evenly across the eight waste package failure groups. To investigate the effects of waste package failure group on dose at the accessible environment Category 1 spent fixel was selected. This category was selected because it alone is the significant contributor to peak dose at the accessible environment from a composite of the 13 categories of DOE W. Figure 5.7-1 shows a comparison of results assuming that all of the packages are in waste package f$ilure Group 1 and assuming that the packages are evenly distributed among the eight waste package failure groups shown in Figure 2.6-2. The peak dose is lowered by about an order of magnitude for the latter assumption, Another observation that can be made is that the HLW needed to dispose of all of the DOE W (Figure 4.2-1) produces about the same peak . do& as that from the composite of all DOE SNF (which is approximately represented by the lower curve of Figure 5.7-1 because Category 1 DOE SNF is the significant contributor to peak dose). This may be seen by comparing Figure 5.7.1 and Figure 4.2-1.

5.8 SUMMARY OF RESULTS

The sensitivity analyses show that increases in dissolution rate and surface area would not increase the peak dose fiom the composite of the 13 categories of DOE SNF. The peak dose is due to the

peak concentrations of Technetium and Iodine from Category 1 spent he1 which consists primarily of N Reactor spent fuel. Lowering the percentage of drips on the waste package h m 30% to zero yields more than an order of magnitude reduction in peak dose. The lifetime of drip shields is an important factor in assuring that the percentage of drips is reduced. Drip shield lifetime has little effect on peak dose when they are all assumed to fail instantaneously at the end of their lifetime. A reduction in peak dose of more than an order of magnitude could be achieved if drip shields last indefinitely (the capillary barrier release model). Reduction in peak dose h m the capillary barrier could be achieved by using materials with a lower coefficient of diffusion in the construction of the invert. The amount of cladding failure assumed has a considerable effect on peak dose. However, . the assumption of 100% cladding failure for Categories 1 and 6 is appropriate because of the amount of cladding failure that already exists for N Reactor spent he1 (the majority of Category 1) and the fact that Category 6 SNF has Aluminum cladding that would be expected to degrade rapidly after waste package failure. Category 1 and 6 DOE SNF are the most significant contributors to dose at the accessible environment h m the composite of 13 categories of DOE SNF.

The assumption that all of the DOE SNF is disposed in packages that are in waste package failure Group 1 is very conservative. This assumption yields peak doses that are about an order of magnitude higher than if the DOE SNF was assumed to be distributed evenly over all eight waste package failure groups. When eight waste package failure groups are assumed the peak dose h m the composite of 13 categories of DOE SNF is nearly equal to the peak dose from the HLW used for co-disposal. The peak dose from 13 categories of DOE SNF disposed in eight waste package failure groups, which can be assumed to be approximately represented by the lower curve on Figure 5.7.1 is nearly two orders of magnitude below that f?om 63,000 MTHM of commercial spent fiel. This can be seen in Figure ES 2-5 that shows the DOE SNF and HLW distributed evenly among the eight waste package failure groups.

Figure 5.1-1 Comparison of Peak Dose at the Accessible Environment for Engineered Barrier Release Model "Drips on the Waste Form", "Drips on the Waste Package", and "No Drips" (Capilary Barrier) for Category 1 Spent Fuel.

Figure 5.1-2 Sensitivity of Peak Dose at the Accessible Environment to Diffusion Through an Invert with Assumed Properties of Crushed Tuff, Gravel, and Sand for Category 1 Spent Fuel.


, Figure 5.3-1 Sensitivity of Dose at the Accessible Environment to Percent of Fuel Cladding Failure for Category 1 Spent Fuel.

Figure 5.4-1 Sensitivity of Dose at the Accessible Environment to Assumed Dissolution Model (Metal, &bide, Ceramic, Oxide) for Category 9 Spent Fuel.

1 0-' 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figure 5.5-1 Sensitivity of Dose at the Accessible Environment to Assumed Dissolution Rate for Category 1 Spent Fuel Assuming the Waste Package Relepse Model "Drips on the Was& Package".

Figun 5.5-2 Sensitivity of Dose at the Accessible Environment to Assumed Dissolution kite for Category 1 Spent Fuel Assuming the Waste Package Release Model "Drips on the Waste Fond'.

1 0-7

10-e 0 20,000 40,000 60,000 80,000 100,000

Time (yrs)

Figurr 5.6-1 Sensitivity of Dose at the Accessible Environment to Assumed Surface Area of Category 1 Spent Fuel Assuming the Waste Package Release Model "Drips on the Waste Package".

Figure 5.6-2 Sensitivity of Dose at the Accessible Environment to Assumed Surface Area for Category 5 Spent Fuel Assuming the Waste Package Release Model "Drips on the Waste Package".

Figure 5.6-3 Sensitivity of Dose at the Accessible Environment to Assumed Surface Area for Category 5 Spent Fuel Assuming the Waste Package Rdease Model "Drips on the Waste Fom".

Time (years)

Figure 5.7-1 Sensitivity of Dose at the Accessible Environment to Waste ~&kage Failure Group for Category 1 Spent Fuel.

t ! - ti

Table 5.1 - 1 Diffbion Characteristics of Invert Materials

*Data &om Conca and Wright, 1992. **Saturation obtained fiom thermohydrologic modeling.

Diffusion Coefficient (m2Jyr)

9.87E-5

7.17E-5

6.94E-4

Porosity

0.429*

0.42*

0.56

Material

Basalt Gravel

Coarse Sand

Crushed Tuff

Saturation

0.0549*

0.0474*

-0.1 1 **

6. FINDINGS AND RECOMMENDATfONS

The base case for these analyses is a repository that contains 70,000 MTHM of waste, composed of 63,000 MTHM of commercial spent fuel and 7,000 MTHM of HLW with a thermal loading of 83 WM/acre . For this TSPA of DOE SNF rhe parameters for the site, the waste package, and the repository were updated to the currmt understanding of a potential repository at Yucca Mountain. Each category of DOE SNF was assumed to be disposed of in this repository environment. For ease in analysis, the amount of waste in each category (MTHM) was assumed to be added to the . inventory of the base c'ase repository. Each of the 13 categories of DOE SNF was analyzed with and without the co-disposed HLW, and the results of the latter case were compared to the results fiom an equivalent amount of commercial spent fuel (on an MTHM basis).

A composite of all DOE SNF (2,436 MTHM) and the co-disposed HLW (8,745 MTHM) was analyzed with and without the influence of the HLW, and these results were compared to an . equivalent amount of co~nmercial spent fuel. For the analyses of the individual categories of DOE SM: and for the composite of the 13 categories, the waste packages were assumed to be in waste Sckage failure Group 1 (i.e., the most conservative waste package failure group). This was done to avoid distributing small quantities of DOE SNF evenly among the eight waste package failure groups. Category 1 DOE SNF (2,136.8 MTHM, with N Reactor spent fuel being the dominant fuel type) was analyzed under the assumption of being distributed evenly amongathe eight package groups for comparison to the composite of DOE SNF. Category 1 is the largest contributor to peak dose at the accessible environment because it contains the majority of the spent he1 in the composite (88% on an MTHM basis).

The entire repository containing the DOE SNF md co-disposed HLW (74,181 MTHM) was analyzed and the results were compared to those of the base case. The dose attributed to the DOE SNF and co-disposed HLW was also scaled to examine the effects of a combined 7,000 MTHM (2,333 MTHM of DOE SNF and 4,667 MTHM of HLW) and added to the dose attributed to 63,000 MTHM of commercial spent fuel to examine the effects of the larger repository.

For the analyses in this performance assessment, 30% of the packages were assumed to be located where drips occur and the remaining 70% of the waste packages were at iocations where no drips occur. Where drips occur the conceptual engineered barrier release model "drips on the waste package" was used, and where no drips occur the conceptual release model for a capillary barrier was used. For both cases, release fiom the failed waste package was assumed to be by dimion, and for the capillary barrier there is an additional diffusion pathway through the crushed tuff invert.

The percolation flux through the repository horizon was abstracted from the calibrated site-scale mode1 for the unsaturated zone. For the six vertical columns used in the abstraction, the flux ranges. from 4 to 10 d y r with an average flux of 6.2 mmlyr. The repository drifts were assumed not to be backt?lIed, and no effects of climate were assumed. 7'he reason for not including climate cycles, which could have a significant effect on dose at the accessible environment, is that the site-scale model has not yet been calibrated to examine the transient effects of a climate cycle.

The DOE SNF was assumed to be packaged for disposal in HLW and co-disposal packages. High and medium enriched SNF were codisposed in HLW and co-disposal packages based on a maximum fissile content of 14.4 kg per package. Low enriched DOE SNF was assumed to be disposed in HLW packages that contain four canisters of spent fuel. The waste packages were assumed to have 75% galvanic protection for 90% of the packages when they are distributed evenly among the eight waste package failure groups. Waste package failure was assumed to occur when the outer barrier of the package developed the first through-going pit, and the release fiom a failed package over time was assumed to depend on the number of pits through the packgge. When all eight waste package failure groups are assumed, the first package failures begin to occlir after about 600 years and about 85% of the packages have failed in the first 1,000,000 years.

Transport parameters were updated to the current understanding of the repository environment. In particular the solubility of Neptunium was reduced by two orders of magnitude based on a reanalysis of existing data (Sassani and Siegmann, 1997).

Analyses of the dose history of an individual drinking two liters of ground water per day fiom a well located at the accessible environment (5,000 meters down gradient from the edge of the repository) were conducted for a time period of 100,000 years. The 100,000 year time period was found to be adequate to capture the peak dose h m the repository because of the lower solubility of Neptunium (i.e., dose peaks are fiom Technetium and Iodine which occur before 20,000 years). For some categories of DOE SNF the peak dose occurs beyond 100,000 years because of the r'etardation of Thorium (i.e., Category 8, Uranium-Thorium carbide spent fuel). However the magnitude of the . Thorium peak is small as compared to the peak dose fiom the entire repository (which is fiom Technetium and Iodine).

FINDINGS

• The dose at the accessible environment &om the composite of 13 categories of DOE SNF (2,436.lbfIlM) is somewhat less than that fiom an equivalent mount of commercial spent fuel. Compare Figures 4.2-4 and 4.2-5.

• The largest contributor to peak dose at the accessible environment from the composite of DOE SNF (2,436 MTHM) is Category 1 DOE SNF (2,136.8 MTHM) which is largely composed of N Reactor spent &I. Compare the peak dose on the upper curve of Figure 5.3- 1 to the peak dose on Figure 4.2-4, they are qpproximately the same.

* Distribution of the composite (2,436 MTHM) of DOE SNF evenly among eight waste package failure groups lowers the peak dose by about an order of magnitude. See Figure 5.3-1. The peak dose fiom the composite of DOE SNF, under this assumption, is about the same as that from the HLW used for co-disposal(8,745 MTHM). See Figure ES 2-5. The change fiom 30% of the waste packages being located where drips occur to 10% drips reduces the dose by about a factor of five Figure ES 2-2.

a Six categories of DOE SNF contribute significantly to the dose history at the accessible

environment fiom the composite of 13 categories of DOE SNF, Category 1,4,5,6,8, and 1 1. Of these six categories, Category 1 and Category 6 are the largest contributors to total dose at the accessible environment.

Dissolution rates of the waste based on assumed dissolution models range from approximately 1 d to 1 O3 dm$. The slowest rates occur after the repository temperature returns to neat ambient (i.e., after about 10,000 years). The dissolution rates in descending order are: metallic fuel under wet oxic conditions, metallic fuel under humid oxic conditions, oxide spent fuel, HLW glass, ceramic spent fuel, and carbide spent fuel. However, this five order of magnitude difference is muted by the diffusion rate out of the waste package. For example, the slow d i i i o n rate causes the release rate fiom packages containing different fuel types to be identical for different dissolution rates as long as there are more radionuclides dissolved than can be transported from the package by diffusion.

?he assumption of different engineered barrier release models has a significant effect on the dose at the accesslile environment. The peak dose assuming that 30% of the waste packages are located where drips occur (70% have no drips) and assuming that all of the waste package have no drips lowers the peak dose by more than an order of magnitude. When there are no drips the only mechanism for release from the engineered barrier system is by diffusion out of the waste package and through the invert. The assumption that the drips occur on the waste form rather than the package causes the dose peak from the alteration controlled radionuclides (Iodine and Technetium) to OCCLK earlier and to be a smaller value. T& peak occurs earlier because the advection is more rapid than diffusion, and it is lower because of depletion of alteration controlled radionuclides fiom the first waste packages that fail. For this case the Neptunium peak occurs at about 60,000 years and is somewhat higher than that fiom the alteration controlled donuclides.

For the assumption of a capillary M e r (no drips) the dose at the accessible environment was f w d to change sipiiicantly based on the type of material used to constr~ct the invert. Dose is lowered by nearly two orders of magnitude by using coarse sand rather than crushed tuff because the lower saturation of the sand produces a smaller coefficient of diffusion. When a capiIlary banier effect is assumed to be produced using a drip shield, Be lifetime of the drip shield becomes very important. If all of the drip shields are assumed to fail ~ ~ e o u s l y at the end of their life, the peak dose is moved out in time an mount equal to the assumed lifetime of the drip shield (i.e., there is very little decay because of the relatively long half-life of the radionuclides contributing to the peak). Thus, either the drip shields must last for a long period of time to reduce the peak, or they would have to fiil over a long period of time in order to allow for a significant decay of the Technetium. Technetium has a half-life of more than 200,000 years. If the drip shields are assumed to last forever, then the effects can be simulated by asmming no drips (the capillary barrier).

The assumed &centage cladding failure has a significant effect on peak dose at the accessible environment. However, the assumption of 100% cladding failure for DOE SNF is appropriate because of the characteristics of the two categories of spent fuel that contribute

much of total dose, Category 1 and Category 6. Category 1 is largely composed of N Reactor spent fue) which has a significant amount of cladding failure fiom pool storage at Hanford, and Category 6 DOE spent fuel has Aluminum cladding that would be expected to degrade rapidly after waste package failure (Appendix A). Category 5 spent fuel also contributes significantly to the total dose and it has disrupted cladding. Also, for this study no credit was taken for the canister that contains the waste which is a simplifying and conservative assumption.

a The effects of assumed dissolution rate and sudke area are small for Category 1 DOE SNF. Increasing the assumed dissolution rate of Category 1 SNF does not change the peak dose because release fbm the failed waste package is controlled by diffusion when the conceptual model "drips on the waste package" is assumed. Increasing the surf" area causes lower broader dose peaks because of dilution in the increased volume of water (volume = surface area x film thickness) and depletion of the source tenn fiom waste packages that fail early: When the conceptual waste package release model "drips on the waste fonn" is assumed the dose at the accessible environment does not change significantly for either a change of dissolution rate or of surfsace area This is because the alteration controlled radionuclides (Iodine and Technetium) are released into the flu through the waste package at the alteration rate and the solubiity limited radionuclides (Neptunium) is released into the flux according the their solubility.

RECOMMENDATIONS

The dose at the accessible environment increases significantly (about an order of magnitude for the 13 categories of DOE SNF) when the conceptual waste package release model "drips on the waste form" is assumed in place of the conceptual model "drips on the waste package". The assumption of the model "drips on the waste package" does not give conservative results because at some stage of waste package failure flux will be through the package Tather than around.it. This eiirninates the diffusion pathway. There may be a need to develop a switch fiom the conceptual release model of "drips on the waste package" to "drips on the waste fonn" at some point h time during the analyses.

Should it become necessary, the composite of the I3 Categories of DOE SNF could be approximately represented by a single spent he1 having the combined radionuclide inventories of Category 1 and Category 6 DOE SNF. This would reduce the number of spent fuel categories that have to be to be tracked in the RIP (Repository Integration Program) for the TSPA for the Viability Assessment (TSPA-VA). A more accurate composite is obtained using Categories 1,4,5,6,8, and 1 1.

From a radionuclide inventory point of view, the composite of 13 categories of DOE SNF yield a dose history at the accessible environment similar to that fiom an equivalent amount of commercial spent h l . The chemistry of the composite of DOE SNF and the codisposed HLW was assumed not to alter any of the near-field parameters. Geochemical analyses should be conducted to demonstrate that the DOE SNF and HLW do not produce a chemical

?; + .'.;. '1 ,' . . ,; t d ,

environment that could significantly alter near-field parameters.

Sodium bonded spent fuel and Navy spent fuel, Category 14 and 15, should bC included in future performance assessments (TSPA-VA).

work should continue to update DOE SNF data for futurr performance assessments (TSPA- VA and the TSPAfor the License Application).

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Appendix A

Allocation of repository space to DOE spent nuclev fuel (ShW and high-level wvre (HLW) glys has been identified as 10% of the 70.000 MTHlM total allocated to high-level nuclear waste disposal in the repository under the Nucleu W w Policy Act (198- '1 and its Amendment (1984). Within the 7.000 h m dlwation. 113 of that inventory (or 2333 MIHM) was to be dedicated to DOE-owned spent nuclear h L The balance of the inventory (4667 MElM equivalent) =mains for defense HLW placement within the repository.

The existing DOE SNF inventories include approximately 2700 MTHM of fuels considered suitable for repository disposal. A small quantity of other DOE S M wiU be excluded from consideration either because of their identification as fucls unsuitable for the repository. they were identified for pmccssing due to immediate vulnerabilities. or too few details were known to allow for assignment to a characteristic group.

Packaging fuels for aiticality safety conmls remains a significant isme not only due to the range of enrichments encountered with the heous fuel typesT but also because of the Wile material longevity beyond the design life of the repository. For purposes of packaging conceprc evaluated in tbis performance assessment, W e load b i t s were applied to package c o n f i ~ ~ t i o n s that included a co-diposal concept by mixing HLW and SNF canhers. To allow utilization of repository s t a d d was= package design and not generate excessive void volumes whilc complying with Bsik load limitsT a variety of SNF and HLW canister packaging con&prations was used. Thcse confgudons necessarily resulted in the inc1usion of much more MW than is currently allowed under the IWPA criteria- Waste packaging for this pcdormanoc ausmcn t wil l examine t&e combination of -2700 MIHM SNF (367 MIHM in excess of rhu identified for repositoty disposal) and 7977 MIHM [equivalent] M W (33 10 MlEM in excess) to promote efficient use of waste package volumes.

Defense High-Level Waste (DHLW)

DHLW wiIl consist of boro-silicate @ses produced at tht various facilities (West Valley. Savannah River, Hanford, and the INEEL). These facilities in the past had fuel nprocesSing capabilities which provided separation of fiiion products from the actinides. Generally. these wastes will have much lower aansuranic loadings than that associated with a spent fuel package. The glass ha propaties prescribed by a plars standard relative to panmcters deemed acceptable for disposal in a repository. W HLW glass generated across the DOE-complex will have similar compositions. 1

There are four identified sources of HLW glasses. those being West Valley (XY), Himford (WA). Savannah River (GA), and the Idaho National Engineering and Environmental Labontory (ID).

West Valley wvte glass originates from the only commercial fuel reprocessing facility operated in the US. The waste glass canisters from W D P are 61 cm in diameter and 199 cm long. Tney ye constxucted out of 304L SS and are approximately 0.95 cm thick. Tne projected -

numbers p h for 300 cilllisters3. The wvte in these packages represent fission product inventories from spent fucls irradiated in the 1950s.

Savvlnah River HLW glass represent the accumulated reprocessing wastes from the opention of the various production reactors at that site. These waste canisters also employ a standiud 61 crn diameter and a 300 cm length. Projections for canister count uses 5700 canisers3 These canisters also use 304L SS consauction materials with 0.95 ch wall thickness. Pu loading of up to 50 kg Pu/HLW 'log' has been proposed as a disposal scheme for Pu weapons-grade material, but m-ed glus inventory will not be incorporated as part of this PA

Existing tank farm wastes at Hanford has yet to be solidified into a HLW glass fonn: plans are to use existing boro-silicate glass formulations with similar properties to those used by West Valley and Savannah River. W o r d has requested a variance that wouId allow the use of a 61 cm diameter canister but with an overall length of - 450cm. Materials of construction and canister thicbmesses art expected to parallel those used in the Savannah River glass canisters.

A '

The total number of canisters ir expected m be 11,900 canirrcn5.

6EEL HLW annw have over the years been solidified as a non-conosive, dry p u l a r solid However, such a form would not meet existing ansporration and repository packaging requirements, so further processing of the dry solids will be required m convert tbis rnuerial into an acceptable HLW glass form. To maintain consistency with other sites, the HLW canister will hold to the 61. cm diameter, 0.95 cm wall thickness, and use a 450 cm length. There will be an

For purposes of estimating the MTHM equivalent contribution provided by each glass canister. this study twd a value of 0.5 MlHM for each 300 crn canister and 0.75 MTIIM for each 450 an canister. A single set of radionuclide invkntory dam will be used (on a per package basis) to represent the total HLW inventory even though the actual wastes originate from difFerent producers. The baseline inventory will represent a canister 61 cm in diameter and 300 cm long; a 50% increase in the radionuclide contedt will be used for the longer canisters.

In the inttrvenin_n yean since the cessation of fuel reprocessing, DOE SNF has been p u p e d in various ways dependiq on what analysis or disposd configuration was being studicd. Predominant among those grouping were fuels with common cladding characteristics (zirconium. duminum. stainless steel), similar fuel mimix or 'meat' composition (oxide, carbide. metal), and mpes of enrichments.

Furl grouping by cladding makes sense if claddins credits prove beneficid (or deuimend) to ovenll fuel performulct as measured by ndionuclide retud~uon. Groupins by fuel matrix is suggested when common chemical characteristics may impact leachability/solubility issues when no credit is available for cladding retardation. Where packaging concerns must deal with the issue of criticality safety, both in operational facilities and in the repository post-closurc environment. parameters such as fuel configuntion. fissiie loadings, and/or presence of neutron absorbers must be considered.

For purposes of this performance assessment up to 15 'categories' of fuels were to be evaluated. Evaiuation of those fuels was to be completed in conjunction with. high-level waste ohses incorporated in the repository waste packages using a co-disposal concept. The &egories of fuels selected for the PA modelins were selected based on the chemical composition of tht fuel matrix. This basis of selection evolved from the fact that cladding is not currently 'credited' for fuels identified for repository disposal. Therefore, the chemical constituency of tht fuel ma& will deternine the rate of relese of the various radionuclides based on leachability and soiubiLity.

The nature of the individual fuel packa_@ng approach resulted in a combination of SNF canisters with approximate diameters of 25.4.43.1, and 61 cm in both 300 and 450 cm Ien-&. This variety of fuel canister sizes, when placed with the HLW canisters resulted in a variety of repository waste package combinations within each fuel category. Generally, fuel types (as determined by the originating reactor) within a fuel carcgory were not mixed in common SNF canisters. This approach may create a sb@t increase in the SNF canister count and hence a corresponding increase in the HLW canisters needed to meet co-disposal requirements. However, such an approach does not affect the total MlHM fuel mass.

Determination of any fissile load limits on a per SNF canister basis have yet to be approved for the waste repository. For purposes of this study, W e load limits wen imposed on SM: canister loadings to determine how the package count might be afftmd These load limits were adopted from a study of alumin= fuel packaging and degradation scenarios '. These limits are not intended to be limiting values for any type or category of fuels proposed for rtpository disposaL The limits evaluated in this study are as follows:

HEU (>20%) should not exceed 14.4 kg U-235 equivalent MEU (>5MO%) should not exceed 40 kg U-235 equivalent LEU ( 4 % ) should not exceed 200 kg U-235 equivdent

The categories or fuel groups for this PA are comprised of one or more fuel types. These types may vary in term of physical geometries, total mass, enrichment, or burnup. While other groupings may have se-=gated the fuels by cladding, the categorization of fuels for this PA resulted in analysis of fuels types by fuel maaix composition. No emphasis was placed on any further se-gegation by fuel cladding or enrichments within a given category. However: generally fuel types, (e-g. from two different reactors) within a $vtn catesory wen not 'mixed' m the same SNF canister unless physical geometries, cladding, and BOL enrichments wen the same. There wen no attempts to load a variety of fuels in a canister to maximize fissile Ioding up to a prescribed limit or to minimize void volume.

I Diameter differences in the SNF canisters are not dictated by anything other

I .

I

f ' . :.: '. c,c than the cross-section dimensions of the fuel to be loaded. md only secondvily by the fssile loads. Canister lengths will be determined by fuel lengths, with the majority of fuels destined for loading within 300 cm long canisters. Fuel canisters 450 cm long will be reserved for those fuels requiring the length to avoid disassembly. Selectively, the longer SINF canisters could also be used to sack shortcr fuels. Co-disposal options for 450 cm SNF canisters should prove substvltial if OCRml approves the use of longer canisters in the HLW production facility intended for Hanford's liquid was tt treatment facility.

Canister desip will need to accommodate containment of the fuel load with a h u m pressure of 22 psia.5

Fisile load limits on any canister are not completely inviolate. In cases w h m the ksile quantity in a fuel or group of fuels within a single d t e r approaches the limit for that canister based on beginninpof-life (BOL) exuicbmenf the allowable fssile load may be exceeded by 20 8 of the established fissile limit This variance allowance will have to be pro.ved in a criticality safety evaluation for that specific fuel f i i l e Ioading.

Large disposal package(s) (LDPs) used for DOE EM SNFJHLW disposal will co'nsist of one of five types of available outer containers. In a l l cases, the LDP will consist of an outer carbon me1 layer that is IO-cm thick, and an inner Inconel 625 layer that is 2-cm thick LDP diameters will vary depending on the configuration selected for the individual fucl canisters based on the xespedve fatl lengths, cross-stctional areas, and bile contept

1)tsign parameters may vary somewhat as repository LDP design evolves. However t$e following design information was nsed for the packages assembled for this PA

l@ft standard '

empty package weight = 16730 kg extcmdltngth = 3871 mrn intend length = 3048 mm internal diameter = 1626 mm external diameter = 1725 mm

15-ft ~randard empty package weight = 23450 kg external length = 5247 mm internal len@ = 3810 mm internal diametcr = 1626 mm ex& diameter = 1725 mrn

10-ft 'super' empty package weight =22515kg . external length , = 3871 mm internal Ien-gh = 3048 mm

internal diameter external diameter

15-ft 'super' empty package wcizht external ien-gh . internal length internal diameter e x d diameter

Commercial . inmior len,d = 31413 kg extnnal Ien-gh = 5247 mm internal length = 3810 mm internal diameter = 1790 mm external diameter = 2030 mm

Figures A-1 and A-2 depict at least two of the 4x1 and 5x1 arrays proposed for study in this P A

1625 mm = 63.98 ' ID 266 mrn = 1 0 . 4 7 ' m

WOO rnmCS=4' a 2 0 mm Alloy 625 = .75'

Figure A-1: Proposed 4 W W ) x l(SW Co-disposal Package

Figure A-2: Proposed 50rr.W) x l(SNF) Co-disposal Packase

ShiF canisters .

The variety of individual fuels dictates both length and diameter considerations of the individual fuel canisters. The intent of canning the individual fuels is to faciliate handling the great vviety of fuels as well as the small p m and pieces.

This study will refer to 25.4 cm, 43.2 cm, and 70 cm diameter canisters; standard pipe information is used to determine weights and thicknesses. although actual design may use modified dimensions at some later date. All canisters will be constructed of 304L SS. Typical thicknesses should use 6.35 mm.

25.4 cm x 300 cm pipe - - 770 kg 25.4 cm x 450 cm pipe - - 1144 kg

43.2 crn x 300 cm pipe - - 1300 kg '

43.2 cm x 450 cm pipe - - 1822 kg

61 cm x 300 cm pipe = 61 cm x 450 cm pipe = 1

[specific to Hanfoa multi- = overpack (MCO)

5 HLW x 1 43.2 cm 0 SNF (requirts 'superpack') 4 HLW x 1 25.4 cm 0 SNF?

- 4 MCOs (Hadord N-reactor fuels) - 4 HLW only (no SNF package)

3 HLW x 1 25.4 cm, 432 crn(HEU or MEU), or 61 cm (LEU) 0 HLW x 4 43.2 or 61 cm SNF (LEU)

Note: all the aforementioned package cornbWons may be tncounttred in tither 300 or 4 5 k m length combinations (set tab& data associated with each fuel category)

HLW canister radionuclide inventories are predicated on a generic package loading base on a 300 cm long, 61-cm diameter standard. For those SNF/HLW packaze combinations using 450-cm HLW canisters, the inventory for ttte standard pacwe may be multiplied by 1.50.

Iidividud SNF packages are assigned a representative ndionuclide inventory by fuel category, based on a micat 43.2 sm dialpttcr and a 300 cm long SNF package.

Example calcdations:

0 x 4 (MCOs) = 3 * Category 1 radionuclide inventory

3 x 1 (10-ft) = 3 * HLW canister + 1 Category x ndionucIide inventory

5 x 1 (15-ft) = 5 * (1.5 * HLW canister) + 1 Catezory x radionuclide inventory

Category I fuels [U-met;ll/zirc]

?he X-reactor fuel elements consist of two concentic tubes made of uranium metal co-extnrded into Zircdoy-2 cladding. There art two basic types of fuel elements differentiated by their uranium enrichment Mark N fuels elements a pre-irradiation enrichment of 0.947 8

U-235 in both tubes (see Fi-pre) and an average uranium weight of 22.7 kilo-ms. The Mark IV fuels have an outside diameter of 6.1 cm and a length of 459.62 or 66 centimeten. &lark IA fuel elements have a pre-irradiation enrichment of 1.25 U-235 in the outer tube and 0.947% U-235 in the inner tube. They have an average uranium weight of 16.3 kilograms. Mark IA fuels have an outside diameter of 6.1 centimeters and a lenpth of 38.50. or 53 centimeters*

The de-ded condition of the N-reactor fuels has mated a vulnkbility issue rehtive to their continued storqe in a water environment Breach of the fuel element cladding and longtern water stowe has created an apparent uranium hydride formation Plamed remediation of these fuels currently ihcludcs drylug and controkd oxidation of the hydridc to an oxide for interim storage in a package labeled as a Mdti-Canincr Ovcrpack (Mco).~ The MCO has experienced evolutionary design changes; tfie basic unit will contain a close packed arrangement of cithcr Mark IV or Mark IA fuels. While the original concept of the MCO is not intended as a repository approved disposal package, no alternative or proposed packxge exists at this time. The physical size of the MCO is akin to the standard HLW @ass package, and will therefore be modeled as a 4-pack within the repository overpack.

Each MCO consists of a 61 centimeter outer diameter shell that is 416.6 cm long. - The package has a 0.95 cm wall thickness, and uses 304L W e s s steel

construction. The approximate mass of the tmpq MCO is 16735 kgs. InventoriesTiformation

fission products (41) distribution (refer to TSPA group listing data - Table 1) composition

breached fuel cladding -unnium metal with possible oxide surface coating

wet dissolution rate metal model

slrrtace 1 ~ ~ 3 (m2ig) 1.32E-06

clad failure friction 100%

free ndionuclide inventory fraction 0.01 8

Category 1 U metal, failed clad - ave package mass 34.9 (MT) - repository pkg count 2 - HLW can count. 6 - HLW mass (MT) 19.2 - SNF pkg count 2 - ave U mass (kg) 951.5

. - ave. fissile mass 1 1.25 (kg)

Tables A-1 md A-2 provide a summary listing of the various chemical components associated with the typical N-reactor fuels.

Table A- 1 : Pi-Reactor Fuel Element Description

I I i' . n a t t :Y rat2 I; I

L * . -Y - Z-5-i ryai i ;-:* I i G . : L ~ : E:~i:nac : .?:-G.:-~,. :AY;:ZZC ; . b "C ! - 1 - rsL' . . n z r i cknez; ci 4 - Z ! : I n i I I) - a 7":e-Ltnczh csce' i

I - i 22.5 L3.5 Ic.5 8 ! 2 i . j 22.2 :I*- o i c f i n ) I !

, E i emen: d i m e i t ? ( in ) 1 I i

1 I I ! Outer a f outet 2 . e 2.40 I

I ! - ! 2 . Inner o f outer 1 . R ! . ! I i I ! -2: ! 3 . 9u;cr a': innor ! 4

1.23 I t

1 0.48 ! 0 .&A ! A. lnner o f inner I I .

1 I t ' C!adcinc weight (19) I 1. ~ u t t t element I I 2.61 2.2s z.le 1.74 I 1 . 9 !.a 1.4: I :

_ - - I - - - - - - - I 0 . s I : p v W C f ~ ~ a* : 3.2+ Weight of uraniucr i n

2. Inner element 1 1.21 1.15 1.10 0 . w I 1 - 1 ~

- a u t e t ll b l I I

1. 1 0 . 4 4 X I 3 . 33.1 3.2 23.1 I I :

I . - 2. *!.25:: =Su] I Z L . ~ 22.9 r i . 3 ! .

W e f ~ b t o f urtniun i n 1 16.5 15.5 1 4 . i 10.9 I 12.1 11.3 E.6 1 . inner ( l b ) 0.947% I

I i i

Weighted average o f 50 .O iS .9 . uraniurrr in t l e ~ e n t ( i b l i Ri;!o of z i r = a l ~ y - t t o 140 141.: 143.2 134.1 171.0 1 i t . i 160-7 I I i uraniux I lbl ian) I UeichteC t v e . (Ib/tonl 1 140.6 I 171 .A 1 ': ci r o t a 1 c l~ments I E?. 1 3 7 i

.. % :f Ienpih type cf 1 78 10 I i I &i ! 0 3 1

TabIe A-1: Chemical Composition of i';-Reactor Fuel Elements .

Watniun I - I ZOO I 230 I m a a m I 2.0 1 u l SO

t rm I ?m-- I 0.m-0.m ~ t t I 0. 06-0.21 UC:

h a d I - I tco 1 ' 130

~ s u m a i u n I u 1 20 1 6n

)tanpsnae I ts 1 SO 1 60

rcslVtx!erra 1 - I SO 50

Nickel I 100 0.03-0.Oa w e 1 0.03-0.U utZ

a ~ :WOH 1 n l do 200

hwm . I I - I 2300

S i t i c n - 1 7% I too I s o I S d f ~ - I a to

Tin 1 - I 1 -20- r .a ~c: 1 t . i l - r .m ut:

I . - titrnita - I SO 1 30

tmats - I so 1 100

Uraniun 1 *atme= I 3-s I 4

V d i l r m I - I so I I0

tir:tnira I b~ I aatlrss I aatance

1. Concantrations givzn in parts per mil 1 ion {ppm) maximum or ppm range, unl ess indicatrd otherr~ise.

I ' *

Figure A-3 depicts a typic4 &-reactor fuel element: Figures A-4. and A-5 depict proposed layout of S-reactor fuel packaging within an MCO as it was evaluated in the PA.

105-N REACTOR MARK IV FUEL ELEMENT Asw&Y (f B1000 1 A)

Fi-gut A-3: N-Reactor Mark IV Fuel Element Assembly

- Hgun A-4a: Loading Arrangement for Mark IV Futl in MCO Container

Leadins A r r q u m f b r M L A F d in MCO Camzinc.

Fi=urc A-4b: Loading Amn~ernent for Mark LA Fuel in MCO Container

A A h.1lUA MCO Noma! Case 1 MKNM'CO ~ c m a l ~ o s e I

I

'CASE 1 CASE 2

Figure A-5: MCO Axial- Geometry Layout

Category 3 U-Zr

Typical: CP-5. HWCTR

[dnwing/sketc h?] Inventoriedinfomation

fssion products (31) dismbution (refer to TSPA ,goup listins data - Table 1)

composition wet dissolution rate

m t d model surface area (m21p)

5.00E-05 clad failure fraction

10% fret radionucIide inventory fraction

0.001%

Category 2 U-Zr, Zr clad - ave package mass 25.5 47.6 (MT) - repository pkg count 3 6 - HLW can count 12 30

- HLW mass (MT) 38.4 96 - SNF pkg count 12 6 - ave U mass (kg) 1.6 6.44 - ave. fissile mass 0.44 5.47

(kg)

Category 3 U-hIo/Zirconium

Typicd: ~ermi' Invemtories/ifomntion

fission products (4 1) distribution (refer to TSPA group listing data - Table 1) composition wet dissolution nrc

meal model x 10

surface area (m21g) - 6.80E-05

clad failure fraction 1090

free radionuclide inventory fraction 0.001%

# 5x1 # 5x1 # 0x4 - # 4x1 15ft lo f t 15ft 1 5 f t ,

# 4x1 fi 3x1

Category 3 U-Mo, Zr clad - ave package mass 33.7 (MT) - repository pkg count 55 - HLW can count 220 - HLW mass (MT) 473 - SNF pkg count 55 - ave U mass (kg) 71.45

- ave. fissile mass 18.2 (kg)

lo f t # 3x1

lof t 15ft

Category 1 U oxidel zirconium & stainless steel

Typical: Shippingpon (HEU). commercid (MEU), FFTF-TFA (LEU

(LEU), ML- 1 PBF

The Fuels in this category nenedy have the chmctcristics found in most of the commercial fuels (PWR Gd BWR). For one reason or another. these fuels have ended up in the DOE SNF inventory. For purposes of this PA. they will be analyzed as separate pachges of SNF. but thir analysis will be dom in conjunction with the PWR (Category 14) and B W (Category 15) representing the 63.000 MTHM portion of the repository design.

InventoriesTurfoma~~n fssion products (41) dismbuuon

(refer to TSPA group Listing data - Table 1) cooaposiuon wet dissoiution rate

commercial model surface area (m2/g)

5.179E-05 clad fail- fraction

commercial fice radionuclide inventory hct ion

. commercial

Category 4 U oxide, Zr/SST clad . - ave package mass 33.5

' (MT) - - repository pkg count 3 - HLW can count ' 9

-HLWmass(MT) 19.35 - SNF pkg count 3 - ave U mass (kg) 9.64 - ave. fissile mass 8.88

Figure A-8: 'Shippingport Core II.Blaaket Fuel Assembly

Category 5 , U oxide I failed clad & XI

Typical: SM-1A ORNL SST & Zr (MEU). ThU-2 (LEU). HFIR FRR. hlTR

The fuels in this category'represents those materials that arc tither already . damaged, disrupted, or considered the least robust in tenns of immediate fissile and fission product movement upon package breach. The fuels in.this category have been disrupted fiom heir original codgundon for a numbers of reasons. ?he bullc of .this category consists of the packaged TMI-2 debris. Primary issues related to packaging this material for disposal related to: (1) pach=in,o for criticality control and, (2) drying material toprevent gas g e n d o n .

InventoricsTiormation fission products (41) distribution

(refer to TSPA group listing data - Table 1) composition wet dissolution rate

commercial model surface area (m*lg)

5.10E-03 clad failure fraction

100% free radionuctidc inventory W o n

0.01%

Category 5 U oxide, ma& clad - - ave package mass 23.96 35.7 26.3 35.4 35.6 47.4

(MT) - repository pkg count 382 19 11 . 27 26 . 44 86 - HLW can count 1 146 57 44 1 08 130 220 - HLW mass (MT) 2463.9 182.4 94.6 345.6 279.5 704 - SNF pkg count 382 . 19 11 27 26 4 4 . - ave U mass (kg) 6.49 48.79 22.75 3.15 13.78 2.86 -ave.fissilemass 5.47 9.71 0.65 2.77 7.87 1.22

(kg)

Figure A-9: Buffaio Pulsrar Fuel Rod

Category 6 U-Al, / Al

This category includes fuels composed of a uranium-aluminum alloy. The cladding is assumed to be intact at this time, but is not considered to be a very durable material in lon,o-term storase conditions in wet environments. The nature of the cladding suggests application of a low- allowable centerline tempentun (-200 C) within a waste packa,oe.

lnventoridrnformation fission products (41) distribution (refer to TSPA group listing data - Table 1)

composition wet dissolution rate

metal model d c e area (m*ig)

130E-03 clad failure fraction

10040 frtt tadionuc1.de inventory fraction

0.01 %

Category 6 U alloy, aluminum clad - ave package mass

. - (hAT) - repository pkg count - HLW can count - HLW mass (MT) - SNF pkg COUM - ave U mass (kg) - ave. fissile mass

Figure A- 10: ATR Fuel Element

Category 7 U-Si

Typical: MTR. FRR (HEU, MEU)

Inventories/iionnation fission products (41) distribution

(refer to TSPA group Listing data - Table 1) cornposition wet dissoluaon rau

metal model x 0.1 surface area (rn2ig) -

13E-03 clad fadue fraction

100% fret radionuclide inventory fraction

0.01%

r

Category 7 U silicide, aluminum clad - ave package mass 29 35.9 25.53 36.62 33.56 47.67

(MT) - repository pkg count 2 21 1 1 1 1 99 - HLW can count 6 63 4 4 5 995

- HLW mass (MT) 12.9 ' 201.6 8.6 12.8 10.75 21 39.25 . - SNF pkg count 2 2 1 1 1 1 199

- ave U mass (kg) 29 65.2 25 28.4 20.216 48.75 - ave. fissile mass . 1206 9.83 4.75 5.6 3.996 . 4.39 (kg)

# 4x1 15ft

# 4x1 loft

# 3x1 loft

# 3x1 15ft

# 5x1 loft

# 5x1 15ft

ii 0x4 15ft

Category 8 U/Th carbide (hi-integrity)/ graphite

Ft St Vrain (HEU) consists of a mixture of ZiC-, - and ThC7 particles coated with - different pyrolytic carbon layers and a S i c protective outer coating, which is itself considered to be cladding. These panicles embedded in small diameter compacts that are in turn inserted in channels within large hexa,oonai blocks of graphite. These blocks are 14.172" across the flats x 8.102" on each side 3122" long.

Inventoridiormation fssion producrs (41) distribution

(refer to TSPA iioup Listing data - Table 1) composition wet dissolution rate

carbide model surface area (m*lg)

3.40E-03 clad failme -on

1% free radionuclide inventory fraction

0.001%

- ave package mass (MI) - repository pkg count - HLW can count - HLW mass (MT)

.- SNF pkg count - ave U mass (kg) - ave. fissile mass (kg)

#3xl 10ftd15ft loft l5ft loft 15ft

Category 8 Th/U carbide, graphite (FSV)

# 4x1 # 4x1 # 3x1 # 5x1 # 5x1

Rgvn A- 1 1 : Fr St. Vrah Gapbite Fuel . '

Category 9 U/Th carbide Oow-integrity)/ graphite

The ~eaihbottorn (HEU) fuels were the precursor fuels to the FL St Vnin fuel type. The fdure nte of the particles is eshatcd to be considerable higher than the FSV fuel panicles. Ail the fue t contain HEU, with the addition of Th-232 to U-233 conversion.

I

Inventories/information fusion products (41) disaibution

(refer to TSPA group listing data - Table 1) composition wet dissolution rate

carbide model surface uta (m21~)

5.179E-05 ciad failure fraction

35% free radionuclide inventory faction

0.0001%

- (MT) - repository pkg count - HLW can count - HLW mass (MT) - SNF pkg count - ave U mass (kg) - ave. fissile mass (kg) :

Category 9 ThN carbide, graphite (PB) - ave package mass 48.1 4

# 5x1 lof t

# 4x1 15ft

# 4x1 l o f t

# 3x1 lof t

# 5x1 15.e

# 3x1 15ft

# 0x4 I 15ft

figure A-12: Peach Bottom Unit 1, C o n 1 Fuel Element -wing not to scale; dimensions.in centimeters)

Fi-cuff A-13: Peach Bottom Unit 1. Core 1 Fuel Compacts (Drawing not to scale)

Category 10 U & UIPu carbide / non graphite

Typical: SRE (MEU FGE), FFE Carbide (MEU FGE)

Inventories/iomation fusion products (41) distribution

(refer to TSPA group listing data - Table 1) composition wet dissolution nu

carbide model surface area (rnSg)

5.179E-05 clad failure W o n

10% frtt radionuclide inventory fraction

0.001%

Category Pu carbide, SST clad 10 - - ave package mass 25.53 36.65

(MT) j - repository pkg count . 2 3 1

: - HLW can count 8 12 - - HLW mass {MT) 17.2 25.8 - SNF gkg count 2 3 - ave U mass (kg) 25.14 31.889 - ave. fissife mass 2.76 5.641 .

# 5x1 .15ft

# 3x1 loft

# 0x4 15ft

t 4x1 loft

# 3x1 15ft

# 4x1 15ft

# 5x1 lo f t

Category 11 MO' / (Zr) (SST) (other)

Typical: GE Test ((HEU FGE), FFE-DFA (HEU FGE, Fm-TFA-ACO (LEU & LMEU FGE)

MOX fuels ye composed of a mixture of unnium and plutonium oxides within various claddings. The uranium enrichment qualifies as 'low' but the plutonium content increases the effective enrichment above 158 .

Inventorieslinformarion frsion products (41) dismbunon

( d e r to TSPA group listing data - Table I) composition wet dissolution rate

commercial model surface m a (m2/g)

5.179E-05 clad failure faction

10% free radionuclide inventoty fraction

0.001%

f 0x4 1 5 f t I

Category 11 PuRl oxide, Zr/SST clad . .

- ave package mass 25.6 36.64 33.8 48.3 I (MT)

- - repository pkg count 16 327 7 2 - HLW count 64 1308 35 10 - - HLW mass (MT) 137.6 28122 75.25 32 - SNF pkg count 16 327 7 2 - ave U mass (kg) 9.41 24.15 48.75 458.86 - ave. fissile mass 5.04 7.18 3.67 8.01 . .

(kg)

# 3x1 lof t

# 4x1 15ft

# 3x1 # 4x1 15ft loft

# 5x1 10R

# 5x1 15ft

~ i g u ~ A- 14: KTF Standard Driver Fuel Assembly

FAST R U X TEST FACTLm FUEL PIN BUNDLE CROSS SECTION

(mocatr) .

Figure A-15: FFlF Pin Bundle Cross Section

catq~q 12 C/Th oxide I (Zr) (SST)

lypicd: LWBR (HEU FGE), Dresden mu FGE)

[bwinglsketf h?]

Inventorics/infonnation fission products (41) distribution

(refer to TSPA group listing data - TabT 1) composition . wet dissoiurion rate

c&c model s~rlact aru (mzig)

1 .lOE-04 clad failure fraction

30% free radionuclide hvenroy fraction

Category 12 TlrN oxide. Zirconium - ave package mass

- (MT)

# 0x4

- repositoly pkg count - HLW can count - HLW mass (MT) - SNF pkg count - ave U mass (kg) - ave. fissile mass

(kg)

# 5x1 3 5x1 10R

# 4x1 # 4x1 # 3x1 1Sft

# 3x1 lof t 15ft isn r o r rsa

< . Category 13 U-Zr hydride / (SST) (Incaloy) (other)

Typid: TRIGA Flip (HEU), TRIGA Std. (MEU). TR.IGA Alum (MEU). SNAP (HELr)

Category

Inventories/infoma~on fission products (41) distribution

(refer to TSPA group listins dam - Table 1) composition wet dissolution rate

commercial model x 0.1 sdace arc3 (rnZjg)

1.90E-3 clad failure fnction

0% free radionuclide inventory fraction

0.001%

clad - ave package mass (MT) - repository pkg

count - HLW can count - HLW mass (MT) - SNF pkg count - ave U mass (kg) - ave. fissile mass (kg)

#3x1 lo f t 15ft lo f t 15ft lo f t 15ft 1 5 f t ,

13 U-lr hydride, mixed

.#3x1 #4x1 #4x1 #5x l #5x1 #Ox4

Figm A-16: Smdd TRIGA Fuel Element

Category 14 Conunercid Fuels (PWR) ' (63.000 bITHb4 rep. balance)

Invcntories/infomation fission products (41) distribution

(refer to TSPA group listing data - Table 2) composition wet dissolution nte

NA

rutface aria (m2ig) NA

clad failure fraction NA

fret ndionuclidt inventory fraction ' NA

Figure A-17: Standard PWR Fuel Auernbly

S r c w A-A

Fi-mm A- 18: Proposed PWR Loading Inside a Large Disposal Package

Category 15 Coinmerad Fuels (BWR) (63,000 M'I3fiM rep. balance)

. Inventories/infonnation fssion products (41) distribution

(refer to TSPA group listing ~ak - Table 1) composition wet dissolutibn rate

NA surface are3 (rn2ig)

NA clad failure fraction

NA free radionuclide inventory fraction

NA

Fie- 'A-19: SmM BWR Fuel Assembly

Figure A-20: Roposed BWR Loading Inside a Large Disposal Package

Category 16. eeneric HLW SRS HLW product &u West Vdle y HLW product data INEEL HLW product dau Hylford HLW product dam

Inventorics/iiomation fusion products (4 1) distribution

(refer to TSPA group listing data - Table 3) . composition wet dissolution rate

EA glass standard

Figure A-2 1: Standard HLW Canister

Individual canisters 1 0 0 (for4-pack HLW + 1 S h i

Dimensions . - 26.6 cm OD, Sch 20 (0.635 cm wall thickness); both 300 cm (770 kg) and 450 cm (1 144 kg)

Materials - can: 304L - basket: bonted 300 series SS I

18" E) (for 5-pack HLW + 1 SNF) I - Dimensions - 45.7 cm OD, Such 10 (0.635 cm wall thickness); both 300 cm (1300 kg) and 450 cm (2103 kg)

Materials - can: 304L

i - bdcct: borated 300 series SS

24" 0 (for 3-pack HLW + 1 SNF) Dimensions - 61 cm OD, Sch 10 (0.635 crn wall thichess);

both 300 cm (1745 kg) and 450 cm (2442 kg) Materials - can: 3WL - basker boratcd 300 series SS

MCOs (specifk to N-reactor fifuels) (4-pack) Dimensions - 6 1 cm OD - length: 416.6 cm - weight 3685 kg Materials-U)4L

- HLwstandardcanister Dimensions - 61 an OD; both 300 an (1745 kg) and 450 cm (2442 kg) Matcriats - can: 304L

Criticality Issues

Us of ht& loadings on a per SNF canister b&is offer one approach to meeting of the two contingencies needed to assure criticality safety during the operational phases of fucl handling. Criticality safety evaluations (CSEs) will be needed for all SNF canisters prior to

. p.ansponation, ,md these CSEs 4.95 per regulations) will need to show safety for all operational phases (packaging, traqorwion, and storage) prior to repository ciosure. Adoption of double contingency controls implies both monitoring and remediation capabilities in operational facilities. For _geologic disposal facilities, neither monitoring nor remediation offer viable options. Furthermore, kcff = 1.0 define a criticality rather than the adminisaativt limit of S.95 associated with'operational facilities.

Use of other approaches to provide contingencies to assure criticality safety may allow credits for fur4 neutron absorbers andor controlled spacing of fuel assemblies if long-tenn perfonnanct can be demonsnttd. None of the fissile limits refertnccd in the PA study an

intended to preclude use of one or more of these options in lieu of the fissiie m a s Iirniu md water exclusion bases used for this PA

Then will be a set of dau needs to meet licensing requirements for waste packages intended.for the repository. Some of this chmcr iu t ion data is expect& to be fuel specific due to the variety of the fuel matrices and the resultant chemisuies for a hrlly-flooded waste package. Panmeten such as leachability, solubility. maui.~ dissolution rates. and sm-ace arcs arc among those which can affect mobilization and an rpon of chemical species in a breached waste package. Very lilrle of this rype of L u specific to the repository environment is a c W y available. Use of coase~aave estimates predominate the numbers used to calculate release and aansporr of the various chemical species

ORIGEEr' code data for the DOE S 8 F relied on ORIGEN 11 mu for one or more fuel rypes in each category. These runs provided a generic set of source term inventories (by category) needed to calculate both waste package thermal loads and radiation shielding. They idso provide the bask for calculating rhc oPie quanudes of radionmiides avaiMle for release

. and transport at any point in the ftm.

Cladding composition and mass an available for this mrdy on to the e x t a t they may affect the rare of gas generation. At t&c present h e . no cladding sredia arc being allowed for any of the fuels identified for repository disposal. This approach represents a bounding limif since any allowmce for retardation of radionuclide transport by the prumce of iaracr cladding can only improve rhe results calculated for the waste packages.

. . Packaoinsc corn binations

N-reactor fuels, which constitute -80% of the DOE ShT inventory, are in the process of being removed from their wet stomse environment As pm of this process, they are ~ceiving some minimal ueaunent to address breached andfor damaged fuels prior to dry packaging in multi-canistcr overpacks (MCOr). While these s taide~s steel packages arc intended to provide only interim storage for these hrels. there are no identified packaging alternatives for geologic disposal. Therefore this PA wil l evdutc the N-reactor fuel performance based on a default packaging conccpk evaluating the MCOs inside a repository overpack.

An MCO consists of either 48 Mxk 1A fuelsflayer, with 6 laye&CO, or 54 Mark IV futWlayer with 5 layerdMC0) 7 N-reactor clemenul canister]. '

7 elements / Mark I1 caaisur 28 elements / layer 140 elements / MCO 560 elements / repository overpack

FL St Vrain fuels - stack 5-high in a 44.4 cm (OD) canism iasm

All other remaining (DOE-EM) fucls wiU be contained in: 26.6 cm OD canister, or 45.7 cm OD canister, or 61cmODcanisn

in their qpropriatr Icngth.

intend basket designs may be incorporated into the SNF canincr to faditate packaging andlor a method of neutron absorber fixation as pan of a criticality safev assmancc. Such an intemal basket wodd be s h i b to that being proposed for commercial fuels in the LDP.

Repository overpack 10-cm thick carbon stet1 outer layer 2-cm thick Inconel 625 in& basket design

matuial composition .- thickness

inumd dimensions diamew

(1426 cm) [current standard design - PWR] (160.9 cm) [current standard d e d p - DHLW (proposed 'super pack' to deal with Navy DPCs and 5 x 1 packages]?

, length c450 cm

Combinauons 1 25.4 cm + 4 HLW (standard HLW overpack dwip) 1 43.2 cm + 5 HLW (proposed co-disposd overpPdc size?) 1 432 cm + 3 HLW (1 supplanted HLW) 1 61.0 cm + 3 HLW (1 supplanted HLW)

Terminology

leu ( 4% ) meu( >?% ~ 2 0 % ) heu( >30% )

m e s e enrichment vdues were used as the division criteria for the Sav~mah River Aluminum fuels study.]

mix & matching enrichmenu for similar fucb in a singe SXF canister may be considered if proven beneficial to improved loading efficiencies

fissile load limits based on leu, meu, heu values?

MCO (multicanistcr overpack) this is a Hylford &ism that while intended only for interim dry stomge of Ei-

- reactor fuek, may suf'fice for disposal when coupled with the repository overpack in a multi-pack array

MPUDPC this ori@ concept (54" OD; 15 ft long) has encountered design holdups, but may still considered a viable paclcase suitable for repository disposal.

repository overpack dual Iayer approach - carbon sted(lO.Ocm)/ Inconel 625(2.0cm) 'sunrivability' of this package will nly on the assumptions by RW as tocrtdit

for its longevity and puformancc with DOE SNF contained within the RW design overpack 'oversizec!' to accommodate a 'five +1 concept'?

ORIGEN codes - cdnrlare radionuclide inventory (& bumup) needed for a variety of reasons -two predominant codes - ORIGEN-I1 and ORIGEN-S

FGE (fissiIe gram equivalent) uses the following quation = U-235 conc. + 2*(U-233 conc.) + 2*(Pu-239 conc.)

I ~ P Bisposal Eackage) replaces the terminology that previously addressed what was h o w n as a 'repository overpack' (that would accept eithtr an MPC/DPC or 4 MLW canisters); ?he dcsip of this canister is still in a state of design flux to accornrnodate the proposed variant Ioadings to accommodate the HLW/SM: co- disposal concept d the proposed Navy fuel package sizing rtquirtrnents.

)vet dissolution rate (gmlm2lyr)

surface area (m2igm) d u d e s the surfac:: of lvel ma& marial subject to leaching after some d e g e of irradiation. A values of 3.96 x10-3 m2/g (Gray and Wilson, 1995)

, hap been recommended as a reaso.pable minimum surface area for typic@ spent fuel A

rnz(rimum value oi0.1 mllg (Gray md Waon. 1995) could be used for Kverely disrupted fuels such ;rs the ThfI debris.

(assi-ment of) clad fdure fraction allows some credit to be &en for the contribution any intact cladding may provide upon initial breach of the EBS

6

free radionuclide inventory friction describes the fiaction of radionuclides within a package available for immediate n l e s e upon initial package breach

. . gap radionuclide inventory &tion is that portion of the radionuclide inventory between the fuel cM and the he1 matrix itself.

References:

1. CSER 96-019. Rev.1: Criticality Safety Evduation Report for Spent Nuclev Fue! Processing and Stofage Facilities. WHC-SD-ShT-CSER-005. Octoba 1996.

2. Hanford Irradiated Fuel ~ n v e n t o j Bwline. WHC-SD-CP-TI-175. Febrwry. 1993. .

a 3. Integrated Data Base Repon - 1994. DoE/Rw-Mx)~. Rev 1 I. Seprcmbc: 1995.

4. ~urnrniuy'~~aer ~ e ~ o r t . Exisdng Radionuclide Inventory Data X e d . GW-03-96, .

Pjovember 1996. \

5. Waste Acceptance System Requiremeno Document (WASRD). Rev. 0% September 1995

6. Technical Strategy for the Treatment, Packas@ng. and Disposal of Muminurn-Based Spent Nuclear Fuel. VoL 1, DOE ORce of Spent Fuel Management June 1996.

Table A-3 -c. - , -

! - CCF41FlX .-.------ Cdb1 TSPA -.- Caiqcq,~-=r- c a t e ~ o ? T S ? ~ Catescr, TSPA Cattsc.?, T S ' i Catesoy, TSPA C a ; ~ ~ ~ p A - -- SfiIF, not HLIN 1 1 2 2 3 3 1 _-- _ - ------- -- - - - - ---

kTiiiM packages -- IMTihf sackages klTH~l Packaaes I

INEZ 3a. 6 13 0.W C 3.93 5 5 ; I

S?S 1 Hanfora 2102.2' . *= 103 J 1 Total 2136.8 . 1J8 0. Oc S 3.93 3 - 31 -i

:otal curies Cifpkg total curies Cilpka

Isotopes C 14 6.72Et02 5.70€+00 6.54E-05 7.27E-06 . 1.58E+01 2.87E-01

1 CL 36 0.00€+00 0.00E+00 0.00€+00 0.00€+00 4.01 E-02 7,30E-04 Nl 59 3.798.1.01 3.21 E-01 0.00E100 0.00E+00 4.94Ec01 8.97E-01 NI 63 3.57E+03 3.03E101 .0.00E+00 0.00E+00 i . 0 3 E ~ 0 3 ~ 1 .$8E+01 SE 79 8.33E+01 7.06E-01 2.35E-01 2.61 5-02 1.20E+OI 2.18E-01 S R90 4.67E106 3.96€+04 5.63E+04 6.26E+03 1.07€+04 1.95E+02 ZR93 3.91€+02 3.31E+00 1.21EcOO. 1.34E-01. 4.78€+01 8.69E-01

I

totat curies C i l p ) total cunes C i l ~ k g total curies C i i 0 ~ 2 total curies ,

l r j ? A Catego- TSPA Catason TS?A Catecor, TSPA Carecar, TSPA Catecor, iSPA Cateccr, -- --._.--- -.---_ _ - - - - - -. 1 1 - 1 1 12 12 13 13 -.

hrriM packages hrTihrl ~ e ~ e s lGiihr1 oackaces _--- 1.55 25 4S.66 64 1.96 49

I

I

total curies Ci/pkg total curies ,Ci/pko total curies Ci/pk$

sum pks INEEL 1517.9 sum pks SRS 1424.03.

complex ;total curies

fadcagt Design 31 PWR

- . . laole A 4 Sper.: Fuel Was:: Invec;o;?/ (CocfzccC)

-

Multi-Barrier Waste Package Design - 31 P);IX -

. . (CUpkg)

'Assumes 40.785 PWR with a burnup of 59.651 M W m , and 22311 MN BWR with a burnup of 31,186 MWd(MTU 9-74 ~ c o u t a i n e : , 21 PWR case 'Carbon. marine. and Iodine inventory assumed m be gwout reiease

iabie A-5 DELFv' Was;: Inv tnroq~ (Conccce~)

'Assumed 4 cznistZn per conkner. ' Source: DOE..(2987). Same inventory as DHLW inventory in TSPA-1993.

Iso tope DHLW Inventory (Cirpkg) '

-st I 9.1 SE-2 I IJ1sm I 0

I I% n o - I v c I 330EO

=% 1 1.51E-5

% . 1 124E-5

=Th 1 1.0524 .

d 5.84E4

Documents

OFFICE OF CIVILIAN RADIOACTIVE WASTE MANAGEMENT QA: N .../67531/metadc... · ACKNOWLEDGMENTS . . This document was authored by James 0.Duguid, Jerry McNeish, Vinod Vallikat, and Nelson