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Table of contents :
ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT
Copyright
Acknowledgments
Contents
Executive Summary
INTRODUCTION
BACKGROUND
WASTE FORM QUALIFICATION AND ACCEPTANCE
Metal Waste Forms
Ceramic Waste Form Testing and Plans for Qualification
ADDITIONAL CONSIDERATIONS RELATED TO THE DEMONSTRATION PROJECT
Disposition of the Uranium Product Stream Produced by the EMT Process
The Quantity of EMT Waste Forms Produced from EBR-II Spent Nuclear Fuel
1 Introduction
PRIOR COMMITTEE FINDINGS AND RECOMMENDATIONS ON WASTE FORMS
2 Background
3 Waste Form Qualification and Acceptance
METAL WASTE FORMS
Background
Metal Waste Form Testing and Plans for Qualification
Success Criteria Goals
CERAMIC WASTE FORM
Background
Ceramic Waste Form Testing and Plans for Qualification
Issues
Fabrication of Ceramic Waste Forms
Zeolite Column Operation
Success Criteria Goals
Development
4 Additional Considerations Related to the Demonstration Project
DISPOSITION OF THE URANIUM PRODUCT STREAM PRODUCED BY THE EMT PROCESS
QUANTITY OF EMT WASTE FORMS PRODUCED FROM EBR-II SPENT NUCLEAR FUEL
Appendixes
Appendix A Charge for the Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment (Phase 1 and 2)
Appendix B ANL's Electrometallurgical Demonstration Project Success Criteria
DEMONSTRATION PROJECT SUCCESS CRITERIA
Appendix C Meeting Summary, January 28-29, 1999
JANUARY 28, 1999
Open Session
JANUARY 29, 1999
Closed Session
Appendix D Meeting Summary, March 15, 1999
EXECUTIVE SESSION
OPEN SESSION
EXECUTIVE SESSION
Appendix E ANL-DOE Interactions Related to Waste Forms Generated by ANL's EMT Program
Attachment 1: Repository EIS Meetings
Attachment 2: Spent Nuclear Fuel Program Meetings
Attachment 3: DOE High-Level Waste Programs and DOE-RW Meetings
Appendix F Abbreviations and Acronyms Used in the Main Text of This Report

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT An Assessment of Waste Form Development and Characterization

Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment Board on Chemical Sciences and Technology Commission on Physical Sciences, Mathematics, and Applications National Research Council

National Academy Press Washington, D.C.

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NOTICE: The project that is the subject of this report was approved by the Governing Board of the National Research Council, whose members are drawn from the councils of the National Academy of Sciences, the National Academy of Engineering, and the Institute of Medicine. The members of the committee responsible for the report were chosen for their special competences and with regard for appropriate balance. The National Academy of Sciences is a private, nonprofit, self-perpetuating society of distinguished scholars engaged in scientific and engineering research, dedicated to the furtherance of science and technology and to their use for the general welfare. Upon the authority of the charter granted to it by the Congress in 1863, the Academy has a mandate that requires it to advise the federal government on scientific and technical matters. Dr. Bruce Alberts is president of the National Academy of Sciences. The National Academy of Engineering was established in 1964, under the charter of the National Academy of Sciences, as a parallel organization of outstanding engineers. It is autonomous in its administration and in the selection of its members, sharing with the National Academy of Sciences the responsibility for advising the federal government. The National Academy of Engineering also sponsors engineering programs aimed at meeting national needs, encourages education and research, and recognizes the superior achievements of engineers. Dr. William A. Wulf is president of the National Academy of Engineering. The Institute of Medicine was established in 1970 by the National Academy of Sciences to secure the services of eminent members of appropriate professions in the examination of policy matters pertaining to the health of the public. The Institute acts under the responsibility given to the National Academy of Sciences by its congressional charter to be an adviser to the federal government and, upon its own initiative, to identify issues of medical care, research, and education. Dr. Kenneth I. Shine is president of the Institute of Medicine. The National Research Council was organized by the National Academy of Sciences in 1916 to associate the broad community of science and technology with the Academy's purposes of furthering knowledge and advising the federal government. Functioning in accordance with general policies determined by the Academy, the Council has become the principal operating agency of both the National Academy of Sciences and the National Academy of Engineering in providing services to the government, the public, and the scientific and engineering communities. The Council is administered jointly by both Academies and the Institute of Medicine. Dr. Bruce Alberts and Dr. William A. Wulf are chairman and vice chairman, respectively, of the National Research Council. Support for this project was provided by the Department of Energy. Any opinions, findings, conclusions, or recommendations are those of the author(s) and do not necessarily reflect the views of the agency that provided support for this project. Copyright 1999 by the National Academy of Sciences . All rights reserved. Available from: Board on Chemical Sciences and Technology National Research Council 2101 Constitution Avenue, NW Washington, DC 20418 Printed in the United States of America

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COMMITTEE ON ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT GREGORY R. CHOPPIN, Florida State University, Chair MICHAEL J. APTED, QuantiSci, Inc. PATRICIA A. BAISDEN, Lawrence Livermore National Laboratory EDITH M. FLANIGEN, UOP (retired) CHARLES L. HUSSEY, University of Mississippi FLORIAN MANSFELD, University of Southern California L. EUGENE MCNEESE, Oak Ridge National Laboratory ROBERT A. OSTERYOUNG, North Carolina State University PAUL G. SHEWMON, Ohio State University RALPH E. WHITE, University of South Carolina Staff CHRISTOPHER K. MURPHY, Program Officer MARIA P. JONES, Senior Project Assistant DOUGLAS J. RABER, Director, Board on Chemical Sciences and Technology

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BOARD ON CHEMICAL SCIENCES AND TECHNOLOGY JOHN L. ANDERSON, Carnegie Mellon University, Co-Chair LARRY E. OVERMAN, University of California, Irvine, Co-Chair GREGORY R. CHOPPIN, Florida State University BARBARA J. GARRISON, Pennsylvania State University ALICE P. GAST, Pennsylvania State University LOUIS C. GLASGOW, E.I. du Pont de Nemours & Company JOSEPH G. GORDON II, IBM Almaden Research Center ROBERT H. GRUBBS, California Institute of Technology KEITH E. GUBBINS, North Carolina State University JIRI JONAS, University of Illinois at Urbana-Champaign GEORGE E. KELLER II, Union Carbide Company (retired) RICHARD A. LERNER, Scripps Research Institute GREGORY A. PETSKO, Brandeis University WAYNE H. PITCHER, JR. , Genencor Corporation KENNETH N. RAYMOND, University of California at Berkeley PAUL J. REIDER, Merck Research Laboratories MARTIN B. SHERWIN, ChemVen Group, Inc. CHRISTINE S. SLOANE, General Motors PETER J. STANG, University of Utah WILLIAM J. WARD III, General Electric Company JOHN T. YATES, JR., University of Pittsburgh Staff DOUGLAS J. RABER, Director RUTH MCDIARMID, Senior Program Officer CHRISTOPHER K. MURPHY, Program Officer SYBIL A. PAIGE, Administrative Associate MARIA P. JONES, Senior Project Assistant DAVID A. GRANNIS, Research Assistant (through July 20, 1999)

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COMMISSION ON PHYSICAL SCIENCES, MATHEMATICS, AND APPLICATIONS PETER M. BANKS, ERIM International, Inc., Co-Chair W. CARL LINEBERGER, University of Colorado, Co-Chair WILLIAM BROWDER, Princeton University LAWRENCE D. BROWN, University of Pennsylvania MARSHALL H. COHEN, California Institute of Technology RONALD G. DOUGLAS, Texas A&M University JOHN E. ESTES, University of California at Santa Barbara JERRY P. GOLLUB, Haverford College MARTHA P. HAYNES, Cornell University JOHN L. HENNESSY, Stanford University CAROL M. JANTZEN, Savannah River Westinghouse Company PAUL G. KAMINSKI, Technovation, Inc. KENNETH H. KELLER, University of Minnesota MARGARET G. KIVELSON, University of California at Los Angeles DANIEL KLEPPNER, Massachusetts Institute of Technology JOHN KREICK, Sanders, a Lockheed Martin Company MARSHA I. LESTER, University of Pennsylvania M. ELISABETH PATE-CORNELL, Stanford University NICHOLAS P. SAMIOS, Brookhaven National Laboratory CHANG-LIN TIEN, University of California at Berkeley NORMAN METZGER, Executive Director (through July 1999) MYRON F. UMAN, Acting Executive Director (as of August 1999)

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ACKNOWLEDGMENTS

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Acknowledgments

This report has been reviewed by individuals chosen for their diverse perspectives and technical expertise, in accordance with procedures approved by the National Research Council's (NRC's) Report Review Committee. The purpose of this independent review is to provide candid and critical comments that will assist the authors and the NRC in making the published report as sound as possible and to ensure that the report meets institutional standards for objectivity, evidence, and responsiveness to the study charge. The contents of the review comments and draft manuscript remain confidential to protect the integrity of the deliberative process. We wish to thank the following individuals for their participation in the review of this report: John Ahearne, Sigma Xi, The Scientific Research Center, Stephen Berry, University of Chicago, Robert Budnitz, Future Resources Associates, Inc., Kenneth Czerwinski, Massachusetts Institute of Technology, Lloyd Heldt, Michigan Technological University, Andrew Kadak, Massachusetts Institute of Technology, George Parshall, E.I. du Pont de Nemours & Company (retired), and Thomas Pigford, University of California at Berkeley. Although the reviewers listed above have provided many constructive comments and suggestions, they were not asked to endorse the conclusions or recommendations, nor did they see the final draft of the report before its release. Responsibility for the final content of this report rests solely with the authoring committee and the NRC.

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ACKNOWLEDGMENTS viii

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CONTENTS

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Contents

EXECUTIVE SUMMARY

1

1

INTRODUCTION

7

2

BACKGROUND

11

3

WASTE FORM QUALIFICATION AND ACCEPTANCE

13

4

ADDITIONAL CONSDERATIONS RELATED TO THE DEMONSTRATION PROJECT

27

A B C D E F

APPENDIXES Charge for the Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment (Phases 1 and 2) ANL's Electrometallurgical Demonstration Project Success Criteria Meeting Summary, January 28-29, 1999 Meeting Summary, March 15, 1999 ANL-DOE Interactions Related to Waste Forms Generated by ANL's EMT Program Abbreviations and Acronyms Used in the Main Text of This Report

31 33 37 49 51 59

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CONTENTS x

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EXECUTIVE SUMMARY

1

Executive Summary

INTRODUCTION The Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment has functioned since January 1995 to evaluate Argonne National Laboratory's (ANL's) spent fuel demonstration project for the Department of Energy.1 Over this period, the committee has operated in three phases and has released eight reports detailing various aspects of the demonstration project as it has proceeded. The present report is the ninth in this series and will be followed by a tenth, and final, report to be released after the completion of ANL's demonstration project. The Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment has three parts to its charge for phase 3: 1. Continue its ongoing evaluation of Argonne National Laboratory's (ANL's) demonstration project, and issue a final report at the end of the demonstration; 2. Review the viability of electrometallurgical technology in light of technical progress in other possible treatment technologies; and 3. Evaluate the criteria developed by ANL and the U.S. Department of Energy (DOE) to determine the success of the demonstration project. The committee in its seventh report2 addressed the second and third aspects of its charge. The committee's eighth report,3 the present (ninth) report, and the committee's final report all have addressed and will address the first aspect of the committee's charge for phase 3. The committee's previous report4 focused on the chemical and electrometallurgical process steps of electrometallurgical technology (EMT) and on ANL's progress in the Experimental Breeder Reactor-II (EBR-II) demonstration project. The committee stated in that report that it would defer the part of its charge involving assessment of the waste forms being developed by ANL for disposition of the products of the EMT treatment. The present report discusses these waste-form issues as they relate to the completion of the demonstration project, as well as postdemonstration testing that will be necessary for placement of these waste forms in a geologic repository.

1The

committee's charge for phases 1 and 2 is included in Appendix A.

2National

Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1998. 3National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity as of Fall 1998, National Academy Press, Washington, D.C., 1999. 4National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity as of Fall 1998, National Academy Press, Washington, D.C., 1999.

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EXECUTIVE SUMMARY

2

BACKGROUND Electrometallurgical processing technology produces two waste streams:5 • the metallic waste form (MWF) and • the ceramic waste form (CWF). The novel structure and composition of the MWF require a verification approach that can demonstrate its suitability as a final waste form for geologic disposal. R&D and evaluation led to the selection of glass-bonded sodalite as the reference CWF. The committee in this report assesses waste qualification activities related to the MWF and CWF produced by this process and provides an evaluation of the technical consistency and comprehensiveness of the EMT activities that support ultimate acceptance of EMT waste forms by DOE's Office of Civilian Radioactive Waste Management (RW). WASTE FORM QUALIFICATION AND ACCEPTANCE DOE, through its Office of Civilian Radioactive Waste Management (DOE-RW), is assessing the viability of permanent disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a deep geologic repository at Yucca Mountain, Nevada.6 The performance and compatibility of the ANL waste forms must be assessed within a system context of overall repository safety. DOE asked the committee to evaluate ANL's progress in taking appropriate steps for obtaining the necessary regulatory approvals in the future. The EMT Program has developed a waste qualification program that is patterned after the protocols used for the waste qualification of defense HLW (DHLW) borosilicate glass.7 To date, both commercial spent fuel and vitrified defense HLW have been subjected to detailed characterizations conducted with respect to their performance in a geologic repository.8 For the EMT waste forms, preliminary evaluation is being performed by the DOE-RW, and no final decision on the EMT waste form has been made by RW. Issues remain that affect the ability of the EMT Program to fully document its plans and schedule for achieving a future waste-acceptance decision. First, the EMT Program is concluding a directed demonstration phase that supports issuance of an Environmental Impact Statement (EIS) regarding continued application of the EMT to process the remaining inventory of EBR-II spent fuel. A final consideration is that the initial draft of RW's Acceptance Criteria document should be issued for review in 1999. This new document may modify the actual waste-acceptance strategies and waste-acceptance criteria that the EMT Program is currently following.

5This

process also produces a uranium stream. ANL does not consider uranium a waste product. Assessment of a Repository at Yucca Mountain, DOE/RW-0508, U.S. Department of Energy Office of Civilian Radioactive Waste Management, Washington, D.C., 1998.

6Viability

7DWPF Waste Acceptance Reference Manual (U), WSRC-IM-93-45, Westinghouse Savannah River Company, Savannah River Site, Aiken, SC, 1993. 8Mined Geologic Disposal System Waste Acceptance Criteria Document, B00000000-01717-4600-00095 REV 00, TRW Environmental Safety Systems, Inc., Las Vegas, NV, 1997, pp. 5-1 – 5-8.

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EXECUTIVE SUMMARY

3

Finding: From interactions with RW, ANL has developed a strategy appropriately based on RW's waste acceptance criteria for the characterization of its MWF and CWF for eventual acceptance by RW. This strategy encompasses its characterization protocols, including short-term test procedures, for its ceramic and metal waste forms. Conclusion: Continued interaction between ANL and RW will become even more important in the postdemonstration phase. Conclusion: There remains uncertainty regarding which DOE organization will be charged with the ultimate responsibility for performance-confirmation testing of waste forms suitable to support a repository licensing decision. As this uncertainty in responsibility could lead to costly duplication of effort and lack of consensus among DOE organizations regarding data supporting future decisions, DOE should take the lead in achieving a documented resolution to this issue. Metal Waste Forms The electrometallurgical treatment of spent EBR-II reactor fuel involves a set of operations designed to disassemble driver and blanket fuel pins, to refine and recover the uranium metal contained therein, and to segregate the radioactive waste components. For the entire EBR-II spent fuel inventory, the base metal waste stream composition is stainless steel containing approximately 15 weight % zirconium, labeled SS-15Zr by ANL personnel. 9 Thus, MWF testing at Argonne National Laboratory-West (ANL-W) has focused primarily on this and similar alloys. The corrosion resistance of SS-15Zr alloys has been determined using immersion tests, electrochemical tests, and accelerated corrosion tests (vapor hydration, high-temperature immersion, and product consistency tests). Good progress seems to have been achieved in the identification of the various microstructures of SS-15Zrtype materials. Noble metal-rich precipitates have not been observed. Most of the corrosion tests have shown either no corrosion or only slight tarnish. However, it was found that corrosion rates were greatly accelerated by exposure to steam. Surface corrosion was observed in the pulsed-flow immersion test of the SS-15Zr MWF. No data have been presented to the committee thus far for “standard” SS-15Zr samples. Corrosion appears to be retarded by the formation of a passivating oxide layer that may trap the fission products and actinides, limiting their release. Finding: Some of the corrosion products, which may sequester radionuclides, might remain on the sample surface and might not be detected by solution analysis. Conclusion: Corrosion behavior data for the SS-15Zr MWF standard need to be obtained.

9D.

P. Abraham, Metal Waste Form Handbook, NT Technical Memorandum No. 88, Argonne National Laboratory, Argonne, IL, 1998, p. 3.

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EXECUTIVE SUMMARY

4

Recommendation: Surface analysis by x-ray photoelectron spectroscopy (XPS) or Auger electron spectroscopy (AES) should be performed for selected samples to determine the chemical composition of passivation filings and/or corrosion products. Because a large number of samples to be tested differ only slightly in minor alloying elements, it is recommended that only a few of these samples be subjected to full characterization. These samples should be selected using a statistical analysis approach. The corrosion resistance of the MWF appears to be dominated by the passivation behavior of the alloy, and dissolution is not considered to be a dominant release mechanism of the radionuclides. Finding: Results from corrosion testing of the MWF in rather benign environments suggest that the corrosion behavior of the MWF is similar to that of stainless steel. Finding: At the present time, ANL has not indicated how it plans to conduct crevice corrosion studies. Finding: ANL has carried out a number of corrosion tests using mild solutions. Under these conditions, significant corrosion damage to the MWF is not expected. Recommendations: Instead of continuing to conduct a large number of corrosion tests using mild conditions, it would be better to subject a few carefully selected samples to additional evaluation by surface analysis to determine the chemical composition of the corrosion products. It may be better to concentrate on a few key samples, expose them at higher temperatures, and then obtain electrochemical and surface analysis data. Ceramic Waste Form Testing and Plans for Qualification ANL reported10 on the detailed testing to support CWF qualification using “scoping” tests.11 These tests are carried out to provide data for assessment of the CWF's dissolution and to provide data for the development of a model to predict waste form stability in the repository. Finding: ANL's tests of several months' duration indicate that the CWF dissolves at a rate equal to or less than that of the reference DHLW borosilicate glass. Conclusion: If dissolution remains the dominant release mechanism under actual

10Presentation

to the committee by William Ebert, National Academies Beckman Center, Irvine, CA, January 28, 1999. 11For background on the development of these testing methods as they apply to the ceramic waste form produced by ANL's demonstration project, see L. J. Simpson, D. J. Wronkiewicz, and J. A. Fortner, Development of Test Acceptance Standards for Qualification of the Glass-Bonded Zeolite Waste Form Interim Annual Report: October 1995 – September, 1996, ANL Technical Memorandum No. 51, Argonne National Laboratory, Argonne, IL, 1997.

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EXECUTIVE SUMMARY

5

repository conditions, then the release performance of the CWF will be at least comparable to that of DHLW borosilicate glass. ANL reported12 on modeling long-term behavior based on short-term data. The model is based on a transition state theory approach for the rate of silicate mineral dissolution.13,14 ANL15 discussed accelerated alpha damage testing on simulated CWF doped with 0.2 to 2.5 weight % 238Pu 239 or Pu. Finding: During the conduct of the alpha-decay tests, plutonium oxide was observed as nanocrystals in the grain boundaries. Conclusion: Plutonium may not be in the sodalite phase.16 Its presence in potentially colloid-sized products may have implications for the long-term release behavior of plutonium and any other radionuclides that also segregate into such colloid-sized phases. The EMT Program's waste-form qualification program is based on adaptation of models and test protocols developed for DHLW borosilicate glass. Recommendation: The EMT Program should continue to evaluate and demonstrate that test protocols and conceptual models developed for monolithic single-phase borosilicate glass can adequately represent the behavior of the nonhomogeneous multiphase EMT CWF. There is to date little or no direct evidence quantifying the distribution of radionuclides among the phases. Finding: The Material Characterization Center test (MMC-1) and product consistency test (PCT) designed to model the release behavior of inert, major components of the CWF may be irrelevant with respect to evaluating the release of plutonium and other actinides partitioned into separate oxide phases. Conclusion: The committee believes that ANL is taking appropriate steps to coordinate its wastequalification program with the DOE-RW repository program. It remains undemonstrated, however, that direct adaptation of test procedures and models developed for measuring the rate of general corrosion of the matrix of homogeneous, vitrified HLW forms are sufficient for evaluating the performance of the heterogeneous, crystalline CWF under long-term repository conditions.

12Presentation

to the committee by Thomas Fanning, National Academies Beckman Center, Irvine, CA, January 28, 1999. Glasstone, K. Laidler, and H. Eyring, The Theory of Rate Processes, McGraw-Hill, New York, 1935. 14A. C. Lasaga, and R. J. Kirkpatrick, eds., “Transition State Theory” in Kinetics of Geochemical Processes, Reviews in Mineralogy, Mineralogical Society of America, Washington, D.C., 1981. 15Presentation to the committee by Stephen Johnson, January 28, 1999. 16Sodalite is the name of a group of alumino-silicate framework materials formed by linkage of SiO and AlO tetrahedra 4 4 that form internal cavities that can be occupied by chloride or other ions. 13S.

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EXECUTIVE SUMMARY

6

Conclusion: These continuing concerns are not expected to jeopardize the timely completion of the EBR-II demonstration project in 1999, but attention should be devoted to their resolution prior to extending the EMT process past the demonstration. When waste form acceptance criteria for geologic repository placement are adopted by RW, test procedures for the waste forms produced by the electrometallurgical process may require modification. The committee believes, however, that the test procedures used for the MWF and CWF are appropriate for the completion of ANL's demonstration project. ADDITIONAL CONSIDERATIONS RELATED TO THE DEMONSTRATION PROJECT Disposition of the Uranium Product Stream Produced by the EMT Process The EMT Program staff has discussed applicable product purity levels with the staff of the Oak Ridge National Laboratory's Y-12 Plant for Y-12 acceptance of the uranium metal. If DOE decides that a commercial disposition option is desirable, some additional purification must be sought. To date, these options have not been studied. The Quantity of EMT Waste Forms Produced from EBR-II Spent Nuclear Fuel The committee notes that both the quantity and radionuclide inventory of EMT waste forms are extremely small relative to those of commercial SNF and DHLW. DOE should evaluate whether small quantities, both in terms of volume and radionuclide inventory, of novel waste forms should be characterized and qualified to the same level of detail as major waste forms. However, because final qualification criteria have not been set, it is difficult to assess whether the testing is excessive or not at this stage. Conclusion: Alternative, conservatively bounding strategies for assuring safe disposal of such relatively small quantities of novel HLW may result in significant cost avoidance while still protecting public safety. However, if EMT is to be used for other DOE SNF, a full qualification of the waste forms would be required.

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INTRODUCTION

7

1 Introduction

In the fall of 1998, the Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment entered its last year as an operating body. The committee, appointed by the National Research Council (NRC), has functioned since January 1995 to evaluate Argonne National Laboratory's (ANL's) spent fuel demonstration project for the Department of Energy. Over this period, the committee has released eight reports detailing various aspects of the demonstration project as it has proceeded. The present report is the ninth in this series and will be followed by a tenth, and final, report to be released following the completion of ANL's demonstration project. This ninth report, as with the preceding eight reports, is a result of a request to the NRC made by the U.S. Department of Energy's Office of Nuclear Energy, Science, and Technology (DOE-NE). As a result, the committee's evaluation of ANL's demonstration project, including its findings and recommendations, is a response to this request. This report is technical in nature and contains abbreviations and acronyms related to the demonstration project, although the report is available to the public. For clarity, a list of abbreviations and acronyms is included in Appendix F. References to previous committee reports are noted in the report. The committee has operated in three phases, each of which had a different charge. The charges to the committee for Phases 1 and 2 of its work are given in Appendix A. Phase 3, which began in 1998 and will continue through 1999, has three parts to the committee's charge: 1. Continue its ongoing evaluation of ANL's demonstration project, and issue a final report at the end of the demonstration; 2. Review the viability of electrometallurgical technology in light of technical progress in other possible treatment technologies; and 3. Evaluate the criteria developed by ANL and DOE to determine the success of the demonstration project. The committee in its seventh report1 addressed the second and third aspects of its charge. The committee's eighth report,2 the present report, and the committee's final report all have addressed and will address the first aspect of the committee's charge for phase 3. As the committee began its final year, representatives from DOE and the National Research Council discussed plans for reports for the committee's final year of operation. In addition to a status report on the demonstration project (which addressed the first part of the committee's charge for phase 3) and a final report covering all aspects of this project (scheduled for release in late 1999), the DOE requested that the committee produce a report addressing aspects of waste form testing and evaluation as they relate to

1National

Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1998. 2National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity as of Fall 1998, National Academy Press, Washington, D.C., 1999.

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INTRODUCTION

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ANL's demonstration. The committee's previous report3 was focused on the chemical and electrometallurgical process steps of electrometallurgical technology (EMT) and on ANL's progress in the EBR-II demonstration project. The committee stated in that report that it would defer the part of its charge involving assessment of the waste forms being developed by ANL for disposition of the products of the EMT treatment. This report addresses that aspect of the committee's overall charge. Over the course of the committee's existence, it has from time to time discussed waste form issues in its reports. The present report discusses these waste forms issues as they relate to the completion of the demonstration project, as well as postdemonstration testing that will be necessary for placement of these waste forms in a geologic repository. ANL has developed a waste qualification program for its EMT demonstration project that is patterned after the protocols used for the waste qualification of Defense Program high-level waste (DHLW) borosilicate glass. 4 The early phases of waste-form acceptance modeling and data collection activities by the EMT program are being conducted to provide quality assurance characterization data during a demonstration project for the EMT process. To support a final waste-acceptance decision, however, major qualification/characterization activities will be necessary beyond the end of the demonstration project. ANL's demonstration project for the electrometallurgical treatment of SNF produces two waste forms: a metallic waste form (MWF) and a ceramic waste form (CWF). In addition to these two waste forms, a third product stream, metallic uranium, is produced by EMT. At the present time, DOE has made no decisions about the disposition of this material, and it is the committee's understanding that there are no plans for disposing of it as a waste. Consequently, the committee has not attempted to evaluate the suitability of the uranium product as a possible waste form. PRIOR COMMITTEE FINDINGS AND RECOMMENDATIONS ON WASTE FORMS From its inception, the committee has produced recommendations regarding these waste forms. In a preliminary assessment report, 5 the committee noted that ANL was conducting research and development on both the CWF and the MWF. The qualification of the forms would be conducted in accordance with a waste form qualification plan to be issued in fiscal year 1995. In its first report,6 the committee noted:

3National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity as of Fall 1998, National Academy Press, Washington, D.C., 1999. 4DWPF Waste Acceptance Reference Manual (U), WSRC-IM-93-45, Westinghouse Savannah River Company, Savannah River Site, Aiken, SC, 1993.

5National

Research Council, A Preliminary Assessment of the Promise of Continued R&D into an Electrometallurgical Approach for Treating DOE Spent Fuel, National Academy Press, Washington, D.C., 1995. 6National Research Council, An Assessment of Continued R&D into an Electrometallurgical Approach for Treating DOE Spent Nuclear Fuel, National Academy Press, Washington, D.C., 1995, p. 39.

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INTRODUCTION

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The electrometallurgical technology would generate new waste forms. The fate of the cladding-metal waste form is a major open question, and qualifying the zeolite waste form for burial could present major challenges. In the July 1996 report,7 the committee concluded: Greater priority should be given to the development of a strategy and a relevant test protocol to demonstrate acceptability of waste forms. This activity is of the highest importance relative to all other aspects in the development of the electrometallurgical techniques for weapons plutonium disposition. As a result, the following recommendations were made: A schedule should be developed and implemented for demonstrating waste form performance over a time period commensurate with DOE's plans for treatment of spent nuclear fuel and conversion of weapons plutonium to a form suitable for ultimate disposal. Evaluation of waste form performance is of equal concern for application of the electrometallurgical techniques to treatment of DOE spent nuclear fuel, although the latter application is governed by a different schedule. Waste-form testing should be conducted on the “as-produced” zeolite host phase for radionuclides, as well as on the glass-bonded zeolite waste form.

Continued concern about DOE's acceptability of the waste forms as qualified by the ANL R&D program led to the following recommendations in the status report of the committee in 1996:8 The committee recommends that ANL's ongoing studies be extended to include efforts aimed at defining the phase changes in the salt-loaded zeolite during hot isostatic pressing and determining the fate of the salt, which would no longer be as well isolated from the environment. The committee recommends that attention be given to establishing the performance of both the zeolite and metal waste forms under conditions relevant to their disposal in a geological repository. The committee recommends that the several aspects of ANL's substantial effort in waste form development be integrated into a formal, comprehensive program plan.

7National Research Council, An Evaluation of the Electrometallurgical Approach for Treatment of Excess Weapons Plutonium, National Academy Press, Washington, D.C., 1996, pp. 28-31. 8National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: A Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1996, pp. 8-9.

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INTRODUCTION

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Overall concerns about the fate of the waste forms after the end of the demonstration project were described in 1997:9 The planned development schedule for waste forms pushes strategically important parts of the work beyond the end of the demonstration programs; the effect of such delays on the research schedules needs to be carefully evaluated by ANL and DOE. ANL should develop and implement immediately an overall strategic plan that defines the following:

• the planned state of waste form development at the end of the demonstration phase and the objectives that will remain to be addressed; and • the methods for ensuring optimal, synergistic use of all ANL resources for ceramic waste form development and evaluation. The recommendation for an overall plan for acceptance criteria for the waste forms, agreed upon by DOE and ANL, was repeated in report 6 (1997). 10 In report 7 (1998)11 the recommendation and finding were as follows: Confirmation that the waste forms produced by EMT are acceptable within the DOE's Office of Radioactive Waste (DOE-RW) Office of Civilian Radioactive Waste Management (OCRWM) program for final geological disposal must be a key component in a full qualification of the EMT process. The committee finds that the criteria established by DOE are reasonable for judging the success of the EBR-II spent fuel treatment demonstration.

The issue of final disposition of the waste forms produced by the electrometallurgical process is essential to determining the usability of this process for the treatment of SNF. The present report reviews the waste forms produced by this process and examines this subject in light of DOE requirements for acceptability of waste forms for placement in a geologic repository.

9National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Fall 1996 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1997, p. 11. 10National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity Through Spring 1997, National Academy Press, Washington, D.C., 1997, p. 10. 11 Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1998, pp. 24-25.

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BACKGROUND

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2 Background

Figure 1 shows a schematic of the electrometallurgical processing technology including the production of two waste streams:1 • the metallic waste form (MWF) and • the ceramic waste form (CWF).

FIGURE 1. EBR-II spent fuel treatment flow sheet. The MWF is composed of a mixture of an iron-chromium-nickel alloy (“stainless steel”), 5 to 20 weight % zirconium, and up to 11 weight % uranium.2 The MWF also contains noble metal fission products at a combined concentration of 0 to 4 weight % and lower concentrations of actinides including plutonium. The microstructure of the MWF consists of a stainless-steel phase containing ferrite and austinite, and a zirconium-rich intermetallic phase that contains most of the uranium and actinides present in the MWF. The novel structure and composition of the MWF require a verification approach that can demonstrate its suitability as a final waste form for geologic disposal. The second waste stream is a glass-bonded ceramic waste form (CWF) that will contain the bulk of remaining fission products and transuranic elements, including 1This 2S.

process also produces a uranium stream. ANL does not consider uranium a waste product.

M. McDeavitt, D. P. Abraham, J. Y. Park, and D. D. Keiser, Jr., Metal Waste Form Alloys from the Electrometallurgical Treatment Process, NT Technical Memorandum No. 24, Argonne National Laboratory, Argonne, IL, 1996, pp. 3-10.

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BACKGROUND

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plutonium. This waste form results from the use of zeolite 4A for extraction of fission products and actinides from the internally recycled EMT chloride processing salt.3 Early development of the CWF was based on hot isostatic pressing (HIP) of the zeolite 4A with a borosilicate glass binder to produce a densified waste form. Subsequent R&D showed that a significant fraction of the zeolite 4A was converted to sodalite during HIP. Further evaluation led to the selection of glass-bonded sodalite as the reference CWF. This report provides discussion of critical issues and outstanding questions related to the activities shown in Figure 2 , as well as committee comments and recommendations. This analysis is based on several recent presentations by Argonne National Laboratory (ANL) and documents covering the EMT Program, the committee's continued monitoring of ANL progress during the last 4 years, and published reports on the EMT process. This report assesses the technical consistency and comprehensiveness of the EMT activities that support ultimate acceptance of EMT waste forms by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management.

FIGURE 2. Waste qualification activities involve several organizations. DOE = U.S. Department of Energy; RW = DOE's Office of Civilian Radioactive Waste Management; NE = DOE's Office of Nuclear Energy; EM = DOE's Office of Environmental Energy, WASRD = Waste Acceptance Systems Requirements Document, MGDSWAC = Mined Geologic Disposal Waste Acceptance Criteria.

3Waste Form Acceptance Requirements for Molten Salt Electrorefining of Spent Nuclear Fuel, NT Technical Memorandum No. 8, Argonne National Laboratory, Argonne, IL, 1995, pp. 8-9.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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3 Waste Form Qualification and Acceptance

The U.S. Department of Energy (DOE), through its Office of Civilian Radioactive Waste Management (RW), is assessing the viability of permanent disposal of spent nuclear fuel (SNF) and high-level waste (HLW) in a deep geologic repository at Yucca Mountain, Nevada.1 Long-term isolation of these nuclear materials from the biosphere is planned through a “defense-in-depth” approach of multiple natural and engineered barriers. The performance and compatibility of the Argonne National Laboratory (ANL) waste forms must be assessed within a system context of overall repository safety. The committee recognizes that while general specifications have been identified for acceptance of DOE spent fuel and waste forms into the repository program,2 explicit criteria for evaluating such waste forms under relevant repository conditions are not yet available. Furthermore, the committee recognizes that it is not waste-form performance per se, but rather performance of the integrated system of engineered and natural barriers that must be demonstrated for safety. Nevertheless, DOE asked the committee to evaluate ANL's progress in taking appropriate steps that would be necessary for obtaining the necessary regulatory approvals in the future for the waste forms produced by the electrometallurgical process. DOE-RW is charged with responsibility for determining overall waste-acceptance product specifications.3 However, at this time RW has not yet promulgated waste-acceptance product specifications for any waste form destined for placement in a geologic repository. As a result, any evaluation of ANL's testing program for its ceramic waste form (CWF) and metal waste form (MWF) must be performed within the context of this lack of criteria. The committee does not intend to evaluate whether these waste forms will eventually be accepted in a geologic repository, but rather whether ANL's present testing protocol is adequate to characterize the CWF and MWF. The actual safety standards for a geologic repository have yet to be promulgated by the U.S. Environmental Protection Agency (EPA). There has been considerable debate concerning an appropriate safety standard, a topic on which the National Research Council has published a special study.4 The recent study on Yucca Mountain by RW, 5 for example, indicates that if an arbitrary 10,000-year cutoff is assumed for dose modeling, an isolation strategy based on extended containment (i.e., preventing groundwater from contacting waste forms) would greatly mitigate the importance of waste forms for repository safety. Until EPA finalizes a safety standard, it is difficult for

1Viability

Assessment of a Repository at Yucca Mountain, DOE/RW-0508, U.S. Department of Energy Office of Civilian Radioactive Waste Management, Washington, D.C., 1998. 2Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste between the Assistant Secretary for Environmental Management (EM) U.S. Department of Energy (DOE) and the Director Office of Civilian Radioactive Waste Management (RW) U.S. DOE, Washington, DC, Department of Energy, Washington, D.C., September 1998. 3Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste between the Assistant Secretary for Environmental Management (EM) U.S. Department of Energy (DOE) and the Director Office of Civilian Radioactive Waste Management (RW) U.S. DOE, Washington, DC, Department of Energy, Washington, D.C., September 1998.

4National

Research Council, Technical Bases for Yucca Mountain Standards, National Academy Press, Washington, D.C., 1995. 5Viability Assessment of a Repository at Yucca Mountain, DOE/RW-0508, U.S. Department of Energy Office of Civilian Radioactive Waste Management, Washington, D.C., 1998.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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the U.S. Nuclear Regulatory Commission and DOE to determine exactly what data will be required for safety assessment in support of repository licensing. The MWF and the CWF are HLW forms intended for final disposition in a geologic repository. The electrometallurgical technology (EMT) Program has developed a waste-qualification program that is patterned after the protocols used for the waste qualification of Defense HLW (DHLW) borosilicate glass. 6 The early phases of waste-form acceptance modeling and data collection activities by the EMT Program are being conducted during ANL's demonstration project for the EMT process. To support a final waste-acceptance decision, however, major qualification and characterization activities will continue beyond the end of the demonstration project. To date, both commercial spent fuel and vitrified defense HLW have been subjected to detailed characterizations conducted with respect to their performance in a geologic repository. 7 Such characterization data for borosilicate glass have been used to guide isolation strategies as well as initial design of engineered barrier systems to assess the viability of a geologic repository. 8 At present, only DOE SNF is being grouped by RW with respect to common characteristics. For the EMT waste forms, preliminary evaluation is being performed, and no final decision on the EMT waste form has been made by RW. This preliminary evaluation is using bounding data on these waste forms.9 To assure coordination between RW and DOE's Office of Environmental Management (EM), the “Memorandum of Agreement (MOA) for the Acceptance of DOE Spent Nuclear Fuel and High-Level Radioactive Waste” was issued.10 This MOA establishes the terms and conditions under which RW will make available disposal services to EM for DOE SNF and HLW. The responsibilities of RW and EM relative to data collection, transportation, storage (if needed), safeguards, characterization, and final acceptance for disposal of these materials are identified. The responsibility to treat the EBR fuel rests with DOE's Office of Nuclear Energy, Science, and Technology (NE), but the ultimate disposition of this treated EBR fuel and any HLW waste forms generated is the responsibility of EM. Hence the waste-acceptance activities of the EMT Program will be guided by the MOA. Figure 2 in Chapter 2 shows a flow diagram of the interrelated waste characterization and verification activities of both RW and the producers of DOE HLW (EM and NE).11 From these activities, documents are to be produced that will support the decision for acceptance of EMT HLW forms by RW. This flow diagram is broadly similar to flow diagrams presented in “Appendix C: Subagreement on the DOE SNF and HLW Technical Baseline” in the MOA and the Savannah River Laboratory Defense

6DWPF Waste Acceptance Reference Manual (U), WSRC-IM-93-45, Westinghouse Savannah River Company, Savannah River Site, Aiken, SC, 1993. 7Mined Geologic Disposal System Waste Acceptance Criteria Document, B00000000-01717-4600-00095 REV 00, TRW Environmental Safety Systems, Inc., Las Vegas, NV, 1997, pp. 5-1 – 5-8. 8Mined Geologic Disposal System Waste Acceptance Criteria Document, B00000000-01717-4600-00095 REV 00, TRW Environmental Safety Systems, Inc., Las Vegas, NV, 1997. 9Personal communication, Steven Gomberg, DOE-RW. 10Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste between the Assistant Secretary for Environmental Management (EM) U.S. Department of Energy (DOE) and the Director Office of Civilian Radioactive Waste Management (RW) U.S. DOE, Washington, DC, Department of Energy, Washington, D.C., September 1998.

11Presented

by Robert W. Benedict to the committee, National Academies Beckman Center, Irvine, CA, January 28, 1999.

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Glass Program.12 In particular, Appendix C of the MOA addresses issues related to the development, concurrence, distribution, compliance, and conformance verification of acceptance criteria for DOE SNF and HLW. The specific waste qualification requirements for waste-form producers (NE and EM) by the repository operator (RW) are shown in Figure 2 of Chapter 2 and are defined in Appendix C of the MOA 13 as follows: Acceptance: The transfer of responsibility, custody, and physical possession of DOE SNF or HLW from EM to RW at the EM site. Acceptance Criteria: All technical and programmatic requirements that must be satisfied by DOE SNF and HLW for the repository program to meet regulatory requirements. RW is currently preparing a “Civilian Radioactive Waste Management System (CRWMS) Acceptance Criteria” document. Waste Acceptance Product Specifications (WAPS): The documentation by a HLW producer that identifies the technical specifications for the HLW waste forms. Waste Form Compliance Plan (WCP): The documentation prepared by a HLW producer describing planned analyses, tests, and engineering development work to be undertaken, as well as information to be included in wasteform production records to demonstrate compliance of the proposed waste form with waste acceptance specifications. Waste Form Qualification Report (WQR): The documentation prepared by a HLW producer which describes results of analyses, tests, and engineering development work performed to demonstrate waste-form compliance with acceptance specifications.

Broadly, two types of data requirements can be envisioned within the waste-form acceptance activities shown in Figure 2 of Chapter 2. The first type of data is collected to verify that the as-produced HLW waste forms consistently conform to acceptance specifications for disposal, including topics such as particulates, pyrophoricity, dimensions, radionuclide inventories, and heat-generation rate. The second type of data are those more directly connected to the long-term (10,000 years or more) performance characteristics of such HLW forms under expected repository conditions. The “Waste Testing and Qualification” activity shown in Figure 2 of Chapter 2 is conducted by the waste-form producer to provide specific input to RW's repository performance assessment task. The MOA's section VII “Acceptance Criteria”

12DWPF Waste Acceptance Reference Manual (U), WSRC-IM-93-45, Westinghouse Savannah River Company, Savannah River Site, Aiken, SC, 1993. 13Mined Geologic Disposal System Waste Acceptance Criteria Document, B00000000-01717-4600-00095 REV 00, TRW Environmental Safety Systems, Inc., Las Vegas, NV, 1997, Appendix C.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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notes that RW shall be responsible for the long-term performance characterization of HLW starting in fiscal year 1999.14 Issues remain that impact the ability of the EMT Program to fully document its plans and schedule for achieving a future waste-acceptance decision. First, the EMT Program is concluding a directed demonstration phase that supports issuance of an Environmental Impact Statement (EIS) regarding continued application of the EMT process to the remaining inventory of EBR-II spent fuel. As previously noted, the EMT Program must orient its current activities to provide evidence of successful compliance with demonstration criteria. 15 A final consideration is that the initial draft of RW's Acceptance Criteria document should be issued for review in 1999. This new document may modify the actual waste-acceptance strategies and waste-acceptance criteria that the EMT Program is currently following. Finding: From interactions with RW, ANL has developed a strategy appropriately based on RW's waste acceptance criteria for the characterization of its MFW and CFW for eventual acceptance by RW. This strategy encompasses its characterization protocols, including short-term test procedures, for its ceramic and metal waste forms. Conclusion: Continued interaction between ANL and RW will become even more important in the postdemonstration phase. Conclusion: There remains uncertainty regarding which DOE organization will be charged with the ultimate responsibility for performance-confirmation testing of waste forms suitable to support a repository licensing decision. As this uncertainty in responsibility could lead to costly duplication of effort and lack of consensus among DOE organizations regarding data supporting future decisions, DOE should take the lead in achieving a documented resolution to this issue. METAL WASTE FORMS Background The electrometallurgical treatment of spent EBR-II reactor fuel involves a set of operations designed to disassemble driver and blanket fuel pins, to refine and recover the uranium metal contained therein, and to segregate the radioactive waste components. The radioactive waste components are consolidated into two forms, CWF and MWF. The CWF includes transuranic elements and fission products in a glass-ceramic matrix, whereas the MWF contains noble metal fission products in a fuel-cladding matrix. The

14Appendix C of the MOA contains a listing of interaction between ANL personnel with DOE programs associated with the geological repository and waste form qualification over the past two years.

15

National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Fall 1996 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1997.

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MWF may contain up to 4 weight % noble metal fission products and up to 11 wt % uranium. The long-term corrosion behavior of this type of alloy is not known and needs, therefore, to be determined in the corrosion tests carried out at Argonne National Laboratory-East (ANL-E). The EBR-II driver fuel is primarily a uranium-10 weight % zirconium alloy with type 316, D9, or HT9 stainless steel cladding. Because zirconium is a principal component of the driver fuel, zirconium will be a significant component of the metal waste stream. For the entire EBR-II spent fuel inventory, the base metal waste stream composition is stainless steel containing approximately 15 weight % zirconium, labeled SS-15Zr by ANL personnel.16 Thus, MWF testing at ANL-W has focused primarily on this and similar alloys. Metal Waste Form Testing and Plans for Qualification The MWF is obtained by melting at 1,600 °C in an inert atmosphere the cladding residue that remains from the electrorefiner step. The molten residue is adjusted to contain 15 weight % zirconium and then cast into ingots. Corrosion resistance and noble metal fission product retention are the principal requirements for the safe application of the MWF. Waste-form qualification involves experimental testing and model development. Research at ANL-E has evaluated alloy metallurgy and alloy properties, including mechanical properties, thermophysical properties, and corrosion behavior. 17 The corrosion resistance of SS-15Zr alloys has been determined using immersion tests, electrochemical tests, and accelerated corrosion tests (vapor hydration, high temperature immersion, and product consistency tests). Plans for qualification testing beyond June 1999 and testing highlights have been presented at the committee's recent meetings. 18, 19 The MWF test plan consists of attribute tests, characterization tests, accelerated tests, and service condition tests. The attribute tests, as defined by ANL, are designed to provide material property information using electron microscopy, x-ray analysis, and neutron diffraction. Good progress seems to have been achieved in the identification of the various phases of SS-15Zr-type materials. Noble metalrich precipitates have not been observed. The characterization tests consist of immersion testing in sealed Teflon™ vessels at 90 °C in J-13 (simulated Yucca Mountain well water) and deionized water. The tests that have been terminated showed either no corrosive attack or only minor tarnish. The test solutions have been submitted for elemental analysis. Justification for whether the planned total of 856 tests is necessary to achieve the goals of the project has not been provided to the committee. The accelerated tests are designed to shorten the test period and consist of immersion in deionized water in a titanium vessel at 200 °C for 28 days. Six alloy compositions were tested. Corrosion rates were very low and no correlation of elemental leaching with alloy composition was found.

16D.

P. Abraham, Metal Waste Form Handbook, NT Technical Memorandum No. 88, Argonne National Laboratory, Argonne, IL, 1998, p. 3. 17Presentation by Stephen G. Johnson and Daniel Abraham to the committee, ANL-W, June 25-26, 1998. 18Presentation 19Presentation

by Stephen G. Johnson and Daniel Abraham to the committee, ANL-W, June 25-26, 1998. by Daniel Abraham to the Committee, ANL-E, October 26-27, 1998.

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Electrochemical corrosion testing is based on the polarization resistance technique (ASTM G59),20 which is used to measure instantaneous corrosion rates. 21 Electrochemical corrosion testing is used to screen out alloy compositions that may not be suitable for repository disposal. An example was given for an SS-1Ag-2Nb-1Pd-1Ru alloy that did not contain zirconium and had high corrosion rates at pH 2. Corrosion rates for alloys that contained from 15 to 20 weight % zirconium were low. Corrosion rates of the MWF alloys in J-13 and in solutions of pH = 2, 4, and 10 were low and similar to those of SS316 and alloy C22. These results are not surprising considering that the solutions tested did not contain chloride ions that could have initiated localized corrosion. Corrosion rate data for MWF materials were also compared to those of copper and mild steel. The results from pulsed-flow immersion tests of SS-15Zr alloys containing Nb, Pd, Rh, Ru, and Tc in J-13 water at 90 °C for up to 275 days showed a sudden increase in Tc release rates after about 150 days; however, the overall release rate remained relatively small. The cause of this behavior is under investigation. The results of the immersion tests, which have shown that only small amounts of fission and actinides are dissolved in the test solution, suggest that corrosion of the SS-15Zr MWF is not a dominant release mechanism. Corrosion appears to be retarded by the formation of a passivating oxide layer that may trap the fission products and actinides, limiting their release. Finding: Some of the corrosion products, which may sequester radionuclides, might remain on the sample surface and might not be detected by solution analysis. Vapor hydration tests have been performed in sealed SS vessels for 56 and 182 days. It was found that the corrosion rates were greatly accelerated by exposure to steam. The corrosion product layer for SS-15Zr has been estimated to have a thickness of about 1 µm after 56 days based on visual observations. Samples containing less than 5 weight % zirconium (or no zirconium) were heavily rusted and contained numerous pits. The chemical nature of the corrosion products is under investigation. No data have been presented to the committee thus far for “standard” SS-15Zr samples. ANL personnel did discuss corrosion testing of SS-15Zr MWF samples at a 1998 meeting, concluding that “SS-Zr waste forms are very resistant to the normal corrosion conditions envisioned at the proposed Yucca Mountain geologic repository.” 22 Conclusion: Corrosion behavior data for the SS-15Zr MWF standard need to be obtained.

20ASTM

G 59-91, “Standard Method for Conducting Potentiodynamic Polarization Resistance Measurements.” Mansfeld, “The Polarization Resistance Method for Measuring Corrosion Currents,” in Advances in Corrosion Science and Technology, Vol. 6, p. 163 (1976), Plenum Press. 22D. P. Abraham, L. J. Simpson, M. J. Devries, and S. M. McDeavitt, “Corrosion Testing of Stainless Steel-Zirconium Metal Waste Forms,” Scientific Basis for Waste Management XXII, MRS, Boston, 1998. 21F.

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The effect of radiation on corrosion behavior has been discussed only briefly. Calculations carried out at ANL seem to suggest that radiation levels in the MWF are too low to affect the corrosion resistance. The toxicity characteristic leaching procedure (TCLP) test data suggest that the MWF passes the TCLP test. The results for the release of Ag, As, Ba, Cd, Cr, Hg, Pb, and Se were below the detection limits of acceptable methods. Recommendation: Surface analysis by x-ray photoelectron spectroscopy (XPS) or Auger electron spectroscopy (AES) should be performed for selected samples to determine the chemical composition of passivation filings and/or corrosion products. Because a large number of samples to be tested differ only slightly in minor alloying elements, it is recommended that only a few of these samples be subjected to full characterization. These samples should be selected using a statistical analysis approach. Based on the results from the various corrosion tests, ANL concluded that SS-15Zr shows corrosion behavior similar to that of stainless steels such as SS316. High corrosion rates were observed in electrochemical and vapor hydration tests for alloys with less than 5 weight % zirconium. In the immersion tests, high release of silver was observed for an alloy that did not contain zirconium. The significance of these results is not clear because the alloy composition was below the lower limit of the zirconium specification. The corrosion resistance of the MWF appears to be dominated by the passivation behavior of the alloy, and dissolution is not considered to be a dominant release mechanism of the radionuclides. Finding: Results from corrosion testing of the MWF in rather benign environments suggest that the corrosion behavior of the MWF is similar to that of stainless steel. Finding: At the present time, ANL has not indicated how it plans to conduct crevice corrosion studies. The tests to be performed after June 1999 have not been finalized. Electrochemical tests are to be performed at elevated temperatures in order to shorten the test period. It was suggested by the committee that ANL concentrate on a few key samples, expose them at higher temperatures, and obtain electrochemical and surface analysis data. Tests are also to be conducted in chloride solutions with concentrations of up to 10,000 ppm, which are credible conditions that might be encountered in the repository. Finding: ANL has carried out a number of corrosion tests using mild solutions. Under these conditions, significant corrosion damage to the MWF is not expected. Recommendations: Instead of continuing to conduct a large number of corrosion tests using mild conditions, it would be better to subject a few carefully selected samples to additional evaluation by surface analysis to determine the chemical composition of the corrosion products. It may be better to concentrate on a few key

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samples, expose them at higher temperatures, and then obtain electrochemical and surface analysis data. Guidance for carrying out the needed pitting scans can be obtained from ASTM G61. 23 Corrosion rate data could also be obtained from such measurements. The proposed study of crevice corrosion needs careful design of an artificial crevice with consideration of the proposed application of these alloys. ASTM G48, 24 which describes a multiple crevice assembly, could serve as guidance. Success Criteria Goals Two of the Success Criteria Goals for ANL's demonstration project relate to MWF specifically. These criteria and associated work to date are as follows: Criterion Goal 2-2. Develop metal waste specifications that are based on performance characterization results of small samples with variations in the principal constituents: zirconium, uranium, technetium, plutonium, neptunium, and noble metals. Determine performance characterization with electrochemical techniques, corrosion tests, vapor hydration tests, and attribute tests. Accomplishments as of June 15, 1999. The analysis of 110 samples, 80 of which were spiked with radioactive constituents, is 99% complete and will be summarized in the Metal Waste Form Handbook. Analyses were performed with a wide variety of techniques and procedures, including scanning electron microscopy (SEM) and transmission electron microscopy (TEM), immersion tests, vapor hydration tests, thermal aging tests, crevice corrosion tests, neutron diffraction analysis, bulk material tests, electrochemical corrosion tests, and TCLP. These tests provide the basis for performance of the MWF with reasonable variations in composition of the major components (zirconium and SS) and minor components (U, Tc, Pu, Np, and noble metals). Because Pu and Np are trace constituents in the MWF, the metallurgical behavior of these elements was investigated with small ingot samples that were spiked with up to 6 weight % Pu and up to 2 weight % Np. Criterion Goal 2-3. Develop metal waste process specifications for major process variables: operating temperatures, hold time, and cooling rate. Accomplishments as of June 15, 1999. Three ingots have been cast with irradiated cladding hulls from the electrorefiner and have shown acceptable casting parameters at 1600 °C and a 2-hour hold time. Characterization of these ingots has shown that they are similar to the metal waste test matrix samples. The operating parameters are summarized in the Metal Waste Form Qualification Plan, which is being compiled.

23ASTM

G61-86, “Standard Test Method for Conducting Cyclic Potentiodynamic Polarization Measurements for Localized Corrosion Susceptibility of Iron-, Nickel-, or Cobalt-Based Alloys.” 24ASTM G48-97, “Standard Test Method for Pitting and Crevice Corrosion Resistance of Stainless Steels and Related Alloys by Use of Ferric Chloride Solution.”

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In addition to the tests that have been completed as part of the demonstration project, there are a number of long-term tests on the metal waste form whose results will not be known until after then end of the project. As a result, full characterization of the metal waste form may not be complete by the end of the demonstration. CERAMIC WASTE FORM Background As described in Chapter 2, the reference CWF for the demonstration is glass-bonded sodalite formed from the hot isostatic pressing (HIP) process. 25 Sodalite is the thermolysis product formed during the HIP of the saltloaded zeolite 4A. As shown in Figure 1 in Chapter 2, the salt from the electorefiner containing transuranic elements and fission products is blended with dried zeolite and is heated to occlude the salt into the zeolite. The salt-loaded zeolite is then densified into the CWF by HIP. By appropriate choice of temperature and pressure, the zeolite is converted into sodalite during the HIP process. Scale-up problems previously encountered in the heating of the salt/zeolite blend, and particle size mismatch, were successfully resolved. 26 Ceramic Waste Form Testing and Plans for Qualification reported27

ANL on the detailed testing to support CWF qualification using “scoping” tests. 28 These scoping tests include studies of solution exchange with the CWF; product consistency tests in which the waste form, which is crushed and sieved to achieve suitable particle sizes and washed to remove fines, is leached; the Material Characterization Center Test-1 (MCC-1), a static leach test that uses a monolithic sample; pH stat tests; accessible free salt measurements; and vapor hydration.29 These tests are carried out over a range of environmental conditions but do not provide data on the long-term release-rate performance of these waste forms with respect to relatively insoluble radioelements. Limited tests indicate that dissolution of the sodalite matrix, rather than ion exchange, controls the release to solution of radioelements such as cesium and

25L.

R. Morss, M. B. Clark, W. L. Ebert, W. Hoashi, M. A. Lewis, L. J. Simpson, Donglin Sun, S.W. Tam, C. W. Vander Kooi, D. J. Wronkiewicz, and V. N. Zryyanov, Preliminary Report on the Properties and Behavior of Glass-Bonded Sodalite, Argonne National Laboratory, Argonne, IL, 1999. 26National Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory's R&D Activity, National Academy Press, Washington, D.C., 1998, pp. 7-11. 27Presentation to the committee by William Ebert, January 28, 1999. 28For background on the development of these testing methods as they apply to the ceramic waste form produced by ANL's demonstration project, see L. J. Simpson, D. J. Wronkiewicz, and J. A. Fortner, Development of Test Acceptance Standards for Qualification of the Glass-Bonded Zeolite Waste Form Interim Annual Report: October 1995 – September 1996, ANL Technical Memorandum No. 51, Argonne National Laboratory, Argonne, IL, 1997. 29For background on vapor phase hydration testing, see J. K. Bates, M. G. Seitz, and M. J. Steindler, “The Relevance of Vapor Phase Hydration Aging to Nuclear Waste Isolation,” Nuclear and Chemical Waste Management, Vol. 5, pp. 63-73, 1984.

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strontium.30 The MCC-1 test reflects early dissolution rates, and the PCT tests reflect somewhat longer time dissolution. Neither test, however, reflects the open-system, mass-transport conditions that govern the actual release rates of most radionuclides from the waste package system for disposal.31, 32 The release of highly soluble radioelements (e.g., iodine and cesium), however, probably will be limited by the dissolution rate of the encapsulating phase. Tests to date show that the CWF dissolves at a rate equal to or less than reference defense high-level waste (DHLW) borosilicate glass. This anticipates that the CWF repository performance will be comparable to that of the reference borosilicate glass. Finding: ANL's tests of several months' duration indicate that the CWF dissolves at a rate equal to or less than that of the reference DHLW borosilicate glass. Conclusion: If dissolution remains the dominant release mechanism under actual repository conditions, then the release performance of the CWF will be at least comparable to that of DHLW borosilicate glass. As stated previously, the performance and compatibility of the CWF must be assessed within a system context of overall repository safety. However, until that assessment has been completed, it is significant that the dissolution rate of the CWF is lower than that of reference DHLW borosilicate glass when the two waste forms are subjected to comparable test conditions. The committee is aware that long-term test results could alter this observation regarding the relative dissolution rates of the two waste forms. ANL reported33 on modeling for waste form lifetime in the repository. The model attempts to predict longterm behavior based on short-term data. The model is based on a transition state theory approach for the rate of silicate mineral dissolution. 34, 35 The model incorporates two effects. The first is the forward reaction rate in the absence of dissolved silicic acid. The forward rate is both temperature and pH dependent. The second effect takes into account the relative degree of saturation of silicic acid with respect to the solubility product of the dissolving solid. In previous applications to borosilicate glass, the solubility product of a proxy solid phase (for example, amorphous silica) is used. This approach has been used to model dissolution rates of high-level and low-level borosilicate waste glass for performance assessments. ANL's use of the model is aimed at predicting degradation and radionuclide release for the CWF. The model assumes congruent dissolution of the sodalite— i.e., aluminum and silicon (and presumably other

30L. R. Morss, M. B. Clark, W. L. Ebert, W. Hoashi, M. A. Lewis, L. J. Simpson, D. Sun, S.-W. Tam, C. W. Vander Kooi, D. J. Wronkiewicz, and V. N. Zyryanov, Preliminary Report on the Properties and Behavior of Glass-Bonded Sodalite, Argonne National Laboratory, Argonne, IL, 1999, pp. 15-16. 31 National Research Council, A Study of the Isolation System for Geologic Disposal of Radioactive Wastes, National Academy Press, Washington, D.C., 1983. 32Total System Performance Assessment for Viability Assessment, DOE/RW-0508/V3, U.S. Department of Energy Office of Civilian Radioactive Waste Management, Washington, D.C., 1998.

33Presentation

to the committee by Thomas Fanning January 28, 1999. Glasstone, K. Laidler, and H. Eyring, The Theory of Rate Processes, McGraw-Hill, New York, 1935. 35A. C. Lasaga and R. J. Kirkpatrick, eds., “Transition State Theory,” Kinetics of Geochemical Processes, Reviews in Mineralogy, Mineralogical Society of America, Washington, D.C., 1981. 34S.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

23

minor and trace components) are released to solution at rates whose ratio equals their stoichiometric proportions in the sodalite phase. 36,37 This assumption should be validated. Input data to the model is based on MCC-1 (forward rate) and PCT data (relative saturation) on both simulated and actual CWF. ANL has yet to demonstrate that its resulting model adequately represents measured data on the rate of general corrosion of the crystalline CWF. ANL38 discussed accelerated alpha damage testing on simulated CWF doped with 0.2 to 2.5 weight % 238Pu 239 or Pu. The 238Pu is shorter lived, and hence generates a larger alpha flux than would be found in the CWF. 39 Early results to date at low doses indicate limited cumulative damage. Results were from x-ray diffraction lattice parameters, PCTs, and some SEM and density examination. SEM indicated that plutonium oxide nanocrystals were formed and clustered along grain boundaries. There was some discussion by the committee that the grain size of the plutonium oxide, might lead to increased solubility and has the potential for colloidal particle formation. 40 Finding: During the conduct of the alpha-decay tests, plutonium oxide was observed as nanocrystals in the grain boundaries. Conclusion: Plutonium may not be in the sodalite phase. 41 Its presence in potentially colloid-sized products may have implications for the long-term release behavior of plutonium and any other radionuclides that also segregate into such colloid-sized phases. Issues I. The CWF is a multi-phase, nonhomogeneous material that introduces several issues. SEM images showed the presence, for example, of separate phases of cerium and neodymium oxide, as well as separate nonoccluded salt phases. Many actinides might preferentially partition into such rare-earth oxide phases, and fission products might partition into the free-salt phases. The EMT Program's waste-form qualification program is based on adaptation of models and test protocols developed for DHLW borosilicate glass.

36Sodalite is the name of a group of alumino-silicate framework materials formed by linkage of SiO and AlO tetrahedra 4 4 that form internal cavities that can be occupied by chloride or other anions. 37J. K. Bates, A. J. G. Ellison, J. W. Emery, and J. C. Hoh, “Glass as a Waste Form for the Immobilization of Plutonium,” Material Research Society Proceedings, Vol. 412, W. Murphy and D. Knecht, eds. Materials Research Society, Pittsburgh, PA, 1996, pp. 57-64. 38Presentation to the committee by Stephen Johnson, January 28, 1999. 39S. M. Frank, D. W. Esh, S. G. Johnson, M. Noy, and T. P. O'Holleran, “Effects of Alpha Decay Damage on the Structure and Leaching Rates of a Glass-Bonded Ceramic High Level Waste Form,” Conference Proceedings Material Research Society, Symposium: Scientific Basis for Waste Management XXII, Fall Meeting, Boston, Massachusetts, November 30 December 4, 1998. 40For a study on the potential impact of plutonium impact on repository performance, see: A. B. Kersting; D. W. Efurd, D. L. Finnegan, D. J. Rokop, D. K. Smith, J. L. Thompson, “Migration of Plutonium in Ground Water at the Nevada Test,” Nature, Vol. 397, 1999, pp. 56-59. 41Sodalite is the name of a group of alumino-silicate framework materials formed by linkage of SiO and AlO tetrahedra 4 4 that form internal cavities that can be occupied by chloride or other ions.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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Recommendation: The EMT Program should continue to evaluate and demonstrate that test protocols and conceptual models developed for monolithic single-phase borosilicate glass can adequately represent the behavior of the nonhomogeneous multiphase EMT CWF. II. There is to date little or no direct evidence quantifying the distribution of radionuclides among the phases. This is an area where sufficient information is needed to ensure adequate understanding of the material performance. SEM has shown that cerium, neodymium, and plutonium oxides exist as separate phases. Finding: The material characterization center test (MMC-1) and product consistency test (PCT) designed to model the release behavior of inert, major components of the CWF may be irrelevant with respect to evaluating the release of plutonium and other actinides partitioned into separate oxide phases. III. The committee has noted the importance of understanding the effects of alpha-recoil damage on the CWF as well as the effects of fission product decay on the stability of sodalite- and zeolite-based CWF. ANL has prepared CWF samples containing 238Pu for evaluating the effects of alpha-recoil damage to the CWF. During the operation of the electrometallurgical process, fission product elements that are thermodynamically less noble than uranium accumulate in the molten salt bath from the electrorefiner. These fission product elements then batch equilibrate dry Type A zeolite with the molten salt to produce a CWF that contains fission products. These samples will be evaluated subsequently to determine the effect of fission product loading on the CWF as well as effects of radiation damage to the CWF. The observations from SEM and TEM of the simulated CWF used for alpha-decay damage studies show that the plutonium is present as nanocrystals in a separate oxide phase, primarily at the sodalite grain boundaries. Because the plutonium exists as an oxide in the grain boundaries, it is problematical whether the effects of alpha recoil damage from plutonium decay would be observed by x-ray diffraction measurements of the unit cell dimensions (percent swelling) of the bulk sodalite phase. X-ray diffraction measurements may not be meaningful if the plutonium is not in the sodalite phase because the penetration distance of the alpha-recoil particles would occur predominantly at the grain boundary region. The current x-ray diffraction test would supply damage information only if the plutonium were distributed evenly in the sodalite lattice. Furthermore, the formation of nanocrystals of plutonium oxide after fabrication of the CWF raises questions as to whether plutonium might be released through a geologic repository system as plutonium oxide colloids rather than as dissolved plutonium. Fabrication of Ceramic Waste Forms HIP has been adopted as the reference technology for densification during the demonstration program. To date, tests have been successful using 4.5-inch cans (a limited number of tests with 8-inch cans have also been performed). A cooperative research and development agreement (CRADA) between ANL and the Australian Nuclear Science and Technology Organisation is developing an 8-inch HIP can

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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technology. Initial results appear promising. However, for production in the postdemonstration period, a scale-up to an 18-inch can is planned. ANL is seeking an outside contractor to develop the 18-inch can but, as of this writing, has not located a willing and able vendor. The planned scale-up in canister size, combined with remotehandled hot cell operation, may introduce an unresolved safety question.42 Zeolite Column Operation It has been ANL's intent to use a zeolite column rather than the current batch equilibration method for processing the remaining EBR-II fuel in the postdemonstration period. Column operation is under development but is not part of the demonstration. In the batch mode, traces of water in the zeolite prevent attack of the zeolite by chlorides of uranium and plutonium. However, with column operation, water will not be present and therefore this protection will not occur. For column operation to be viable, scaled-up column performance needs to be evaluated. Success Criteria Goals Two of the Success Criteria Goals for ANL's demonstration project relate to CWF specifically. These criteria and associated work to date are as follows: Criterion Goal 2-4. Develop ceramic waste specifications that are based on performance characterization results of samples with principal constituent variations: glass fission products, uranium and plutonium. Determine performance characteristics with attribute, characterization, accelerated, and service-condition tests. Accomplishments as of June 15, 1999. Approximately 900 laboratory-scale and 83 demonstration-scale samples were produced under different conditions. The results from the characterization of these samples were used to establish glass-bonded sodalite as the reference CWF composition. The present status of the work is documented in the report entitled “Preliminary Report on Glass-Bonded Sodalite Properties and Behavior.” Criterion Goal 2-5. Develop ceramic waste process specifications for major process variables: free chloride, zeolite moisture content, and chloride per unit cell. Accomplishments as of June 15, 1999. Tests of the CWF have examined the importance of free chloride, zeolite moisture content, and chloride per unit cell. The process specifications for these variables have been defined. 43, 44

42ANL's research staff have conducted a risk assessment and arrived at a protocol that ANL finds acceptable. (USQ Safety Assessment: FCF Electrorefiner Waste Salt Processing in HFEF, Argonne National Laboratory, Idaho, Falls, ID, 1998.)

43L.

J. Simpson, D. J. Wronkiewicz, and J. A. Fortner, Development of Test Acceptance Standards for Qualification of the Glass-Bonded Zeolite Waste Form: Interim Annual Report, October 1995 – September 1996, NT Technical Memorandum No. 51, Argonne National Laboratory, Argonne, IL, 1997. 44L. J. Simpson, D. J. Wronkiewicz, and J. A. Fortner, Development of Test Acceptance Standards for Qualification of the Glass-Bonded Zeolite Waste Form: Interim Annual Report, October 1996 – September 1997, NT Technical Memorandum No. 92, Argonne National Laboratory, Argonne, IL, 1998.

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WASTE FORM QUALIFICATION AND ACCEPTANCE

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Analysis of the CWF samples and evaluation of the resulting data may not be complete at the end of the demonstration program. Hence, meeting the above two success criteria goals may be delayed beyond the end of the demonstration period until analysis and data evaluation can be completed. Development Advanced Fabrication. Fabrication technology alternatives to HIP have been considered by ANL. These consist of hot uniaxial pressing (HUP), where the amount of glass is increased to about 25 mass %, and pressureless sintering (PLS), where the amount of glass is increased to about 50 mass %. ANL has stated that it is only using HUP to prepare simulated CWF and is not considering it further. However, pressureless sintering is under development as an alternative to HIP for the postdemonstration period and appears to be an attractive alternative to HIP. PLS would avoid concerns about scale-up in canister size and unresolved safety issues associated with remote HIP in a hot cell. Conclusion: The committee believes that ANL is taking appropriate steps to coordinate its waste qualification program with the DOE-RW repository program. It remains undemonstrated, however, that direct adaptation of test procedures and models developed for measuring the rate of general corrosion of the matrix of homogeneous, vitrified HLW forms are sufficient for evaluating the performance of the heterogeneous, crystalline CWF under long-term repository conditions. It is significant that the tests conducted so far by ANL indicate that the rate of corrosion of the CWF is comparable to that of the reference borosilicate glass. It remains to be seen whether this behavior will continue in long-term testing. Conclusion: These continuing concerns are not expected to jeopardize the timely completion of the EBR-II demonstration project in 1999, but attention should be devoted to their resolution prior to extending the EMT process past the demonstration. When criteria for acceptance of waste forms for geologic repository placement are adopted by RW, test procedures for the waste forms produced by the electrometallurgical process may require modification. The committee believes, however, that the test procedures used for the MWF and CWF are appropriate for the completion of ANL's demonstration project.

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ADDITIONAL CONSIDERATIONS RELATED TO THE DEMONSTRATION PROJECT

27

4 Additional Considerations Related to the Demonstration Project In addition to the testing and ultimate fate of the metal and ceramic waste forms, there are additional considerations related to ANL's demonstration project that are associated with the issue of waste forms. DISPOSITION OF THE URANIUM PRODUCT STREAM PRODUCED BY THE EMT PROCESS As stated previously, the uranium-metal stream is considered to be a product stream that would be incorporated into the nation's enriched-U inventory. Two uranium stream products will be produced—fuel highly enriched (60 to 65%) with U-235, and blanket elements made from depleted uranium but that have been in the reactor for close to two decades and thus have a buildup of plutonium and some fission products. It is not now clear where either stream will ultimately go. The fuel uranium will be diluted down to 19% 235U before it is cast into uranium ingots. The 1996 Environmental Impact Statement1 says that this highly enriched uranium (HEU) will be disposed of by one of two means: 1. Conversion to low-enriched uranium (LEU) light water reactor fuel, or 2. Disposal as low-level waste (LLW) after blending down to 0.9% 235U. Conversion to LEU fuel, by dilution to 4.8% 235U, would still leave a product that would have orders of magnitude more Pu-Np, and more fission products (FP) than the ASTM acceptance standard for LEU fuel. Thus this would not be acceptable without further processing to remove Pu and FP. Disposal as LLW after blending down with depleted uranium (perhaps about 0.25% 235U) to 0.9% would produce a product with about 40 nCi/ gram of transuranic (TRU) waste activity and thus under the limit of 100 nCi/gram required for LLW. It would also require that the volume of the uranium waste be increased by a factor of about 100. ANL is proposing that the blanket elements also be run through the electrometallurgical process to produce a metal stream that is mostly depleted uranium. It is not clear yet whether the process can produce depleted uranium of −0.58, the exposed area < 1 m2 (q = 0.25 m3/y), the flow exceeds 6.5 m3/y (s = 25 m2), and the maximum flow in VA is < 1 m3/y. Sodalite dissolves more slowly than the typical HLW glass, and sodalite dissolution (grams/day) is nearly independent of the exposed surface area. The drip flow model assumes flowing water accumulates silica until it reaches the bottom of the waste form. The drip rate is calculated as 2.5 × 10−3 m3/y, and the drip area is 0.25 to 25 m2. Low flow rates lead to silica saturation. The affinity term, 1 – Q/Ksp, dominates the calculation. The term klong prevails under low flow (high surface area) conditions. Sodalite dissolves slower than HLW glass under these conditions. A lower concentration of silica is needed for saturation, and sodalite dissolution is 3.5 times lower than typical HLW glass. The sodalite dissolution (in grams/day) is independent of surface area. The conclusions from these models are that sodalite dissolves slower than HLW under the considered flow conditions, that the comparison to HLW is conservative, and that future calculations will include temperature and pH effects. Bill Ebert addressed Current Work and Test Matrix. Ongoing work and future plans involve measuring parameter values needed to model the waste form under controlled temperature and pH conditions. Also, data must be provided to demonstrate the use of the mechanistic model. A database must also be generated to evaluate potential methods for monitoring product consistency. The applicability of the model for behavior of demonstration-scale materials under conditions relevant to the disposal system must be confirmed. Finally, the identity and

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APPENDIX C

45

distribution of salts, PuO2, etc., formed during processing for demonstration-scale samples must be determined. The test matrix for qualification of the CWF contains a number of tests to provide model parameter values. The MCC-1 with glass and sodalite measures parameters (k0,
, and Ea), and with the reference CWF for model confirmation. Another test is the pH dependence on acid and base sides at 90 °C before the demonstration, and other temperatures after the demonstration. Long-term PCTs with glass, sodalite, and reference CWF measure model parameters [H4SiO4]sat and klong. Other tests include SEM, TEM, x-ray diffraction (XRD), the density of starting materials and reacted solids for phase identification, radionuclide distribution, and physical characterization. Other tests in the matrix confirm the applicability of the model for long-term disposal. These include VHTs and long-term PCTs with glass, sodalite, and reference CWF for advanced corrosion (e.g., long-term stability, and ID alteration phases). Also required are thermal degradation and low-temperature VHTs. Drip tests are needed to confirm the applicability of the behavior model under repository-relevant conditions with the reference CWF, reference CWF and U, and demonstration-scale samples. Finally, SEM, TEM, and XRD are required for phase identification, radionuclide distribution, and physical characterization. Tests are also required to monitor product consistency. Evaluation of MCC-1, PCT, and soluble salt are needed to measure consistency of the waste forms during production. Daniel Abraham presented the metal waste form qualification update. The focus of the work is on qualification of MWF alloys, specifically those waste forms suitable for repository disposal, and to generate input needed for TSPA analysis. Waste form qualification involves experimental testing, with data feeds into the model development, and model development. The MWF model will be incorporated into the Repository Integration Program (RIP) performance assessment software. Experimental testing encompasses study of alloy microstructures, mechanical and thermophysical properties measurement, and corrosion testing. The modeling approach uses known stainless steel degradation mechanisms as a basis for MWF modeling. It also uses functional dependencies developed for stainless steels. The baseline MWF is a SS-15Zr alloy. The bounding composition is SS-15Zr-11U-0.6Ru-0.3Tc-0.1Pd. The experimental ranges have a Zr content of 0 to 20 weight %, noble metal content of 0 to 4 weight %, Tc content of 0 to 2 weight %, and U content of 0 to 11 weight %. The number of alloy compositions is 39 and there are 111 ingots. A review of the alloy microstrucutre reveals an intermetallic phase [Zr(Fe,Cr,Ni)2+x], and a stainless steel phase containing ferrite and austenite. Noble metal-rich precipitates are not observed and actinides are present only in the intermetallic. Type MCC-1 immersion testing of the MWF takes place in Teflon™ vessels at 90 °C in simulated J-13 and deionized water. Current tests examine specimen corrosion behavior over a

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APPENDIX C

46

range of Zr and noble metal contents. The test duration ranges from 90 days to 365 days to longer terms. New data from the 90-day tests suggest that specimens are either unaffected or show minor tarnish, the weight change of the specimens is small, fission product content in test solutions is small, and normalized corrosion losses are small. These results are consistent with previous tests on the SS-15Zr alloys. Immersion testing at 200 °C is designed to accelerate the corrosion rate. Test data presented previously show the weight change of the alloy specimens is small, normalized loss of fission products is < 0.1 g/m2, and silver loss from SS-1Ag-2Nb-1Pd-1Ru (no Zr) alloy is ~144 g/m2. Examination of specimen surfaces by optical microscopy has revealed that 15 and 20 weight % Zr specimens show uniform corrosion, while a low-Zr alloy (SS-5Zr-2Nb-1Pd-1Ru) showed pitting. Pulsed-flow immersion tests were performed at 90 °C in a simulated J-13 solution/deionized water, for up to 629 days. Tests were interrupted periodically to obtain leachate solutions for elemental analysis. Thirty-six samples contained Tc and/or U. Current results extend previously presented dated for test periods up to 275 days. SEM examination of an SS-15Zr-2Tc sample that was tested in deionized water for 449 days showed mild surface tarnish and some corrosion in the casting pores. The immersion tests indicate that dissolution is not a dominant release mechanism. Also, corrosion appears to be retarded by passivation. Under these conditions, fission products and actinides may be trapped in the passivated oxide layer, hence limiting release. In vapor hydration tests, the corrosion rate of the MWF is accelerated by exposure to steam. The tests are performed in stainless steel vessels at 200 °C for 56 and 182 days. Previous experiments have shown that corrosion layers are typically 1 µm thick (56 test SS-15Zr). Current experiments examine corrosion behavior for seven alloy compositions. Fifty-six day tests were terminated recently. Results show that weight gains were small, and localized attacks (pitting) were observed on samples containing ≤ 5 weight% Zr. The corrosion layer thickness and the nature of corrosion products on the specimen surfaces will be determined in February, and the 182-day tests will be terminated. The purpose of electrochemical corrosion testing is to obtain a relative measure of corrosion rates and to screen out alloy compositions that may not be suitable for repository disposal. The corrosion rate measurement by the linear polarization method is based on ASTM G59. Tests on “cold” specimens are complete and tests on “spiked” specimens (containing Tc and/or U) have begun. Test data show that corrosion rates for alloys that contain from 5 to 20 weight % Zr are similar. A relatively high corrosion rate was observed for SS-1Ag-2Nb-1Pd-1-Ru (no Zr) alloy in a pH 2 solution. Radiation can affect corrosion in three ways. Changes in local water chemistry can have an effect due to formation of radiolysis products (e.g., H2O2). Structural damage may occur to the protective oxide layer. Radiation may also have an effect due to changes in the electronic properties of the oxide. Calculations performed at ANL-E suggest that radiation levels in the MWF will be too low to affect corrosion.

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APPENDIX C

47

In summary, corrosion problems, if any, may exist at low-Zr contents for the samples exposed in the electrochemical, vapor hydration, and immersion tests. Also, MWF corrosion appears to be limited by passivation behavior of the SS base alloy. Steve Johnson spoke about A Long-Term Study of the Effects of Alpha Damage to the Ceramic Waste Form. The effect of radiation damage, specifically alpha decay damage, on an ordered (crystalline) waste form has been studied by several groups but it varied depending on the crystalline host. In an alpha decay event there are two particles that may cause damage: the alpha particle and the recoil nucleus. Ionization damage may result in the following effects: covalent and ionic bond rupture, valence changes or localized electronic excitations, enhanced diffusion processes, and decomposition. Ballistic damage may result in atom displacement. The present study will evaluate the ceramic waste form and the long-term effect of alpha radiation on that waste form. The anticipated plutonium (239Pu) loading in the actual waste form to be produced is 0.2 to 1.0 weight %. Samples made with a higher loading (up to 2.6 weight %) of 239Pu confirmed the similarity of the final product to that made with a lower loading such as 0.6 weight %. This observation allows an increase in the Pu loading without changing the fundamental properties of the sample. The samples produced as a part of this study contain 2.5 weight % 238Pu, and fission products present equivalent to ~100 drivers processed. The loading of Pu utilized is 3 to 12 times that anticipated in the actual waste form to facilitate the experimental methods employed and to accrue the most data in the shortest time frame. The purpose of the test matrix is to develop methods and techniques to evaluate the effects on the CWF of long-term exposure to alpha radiation. These techniques must be sensitive to macroscopic and microscopic changes during the study. The time frame to the test is 4 years. Contained in the test matrix are measurements of density to determine macroscopic swelling. Product consistency tests (PCT) determine the release rates of all elements, in particular Pu, Cs, and I. X-ray diffraction (XRD) analyzes phase-specific swelling or change. Scanning electron microscopy (SEM) checks for microstructural changes, as does transmission electron microscopy (TEM). Thermal properties tested include specific heat and expansion behavior at different temperatures. These tests or methods when taken together will yield a complete picture of the effect of alpha radiation to the CWF. Preliminary results show that 238Pu and 239Pu samples are equivalent and valid for this study. Analysis by XRD shows that in the 239Pu-sodalite CWF, a plutonium oxide and a (minor) halite phase are present. The same result was seen with the 238Pu-sodalite CWF. Likewise, for SEM/TEM analysis both 238Pu and 239Pu samples show agreement with XRD and phases present with small grain size (less than 30 µm in diameter). In the leach test, 238Pu and 239Pu release rate results for Pu and Cl are comparable. Density is comparable for both products. SEM reveals three primary phases present in the 238Pu-loaded CWF: sodalite, glass, and plutonium oxide. TEM bright field imaging of 238Pu-loaded CWF also shows three primary phases: sodalite, glass, and plutonium oxide.

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APPENDIX C

48

Product consistency test/density results indicate that the normalized release rate is ~1E−4 g/m2 day for Pu. A full elemental analysis is still pending. The release rates will be monitored with time using “aged” material to probe for enhanced release of material. Typical increases in the release rates vary from 10 to 100 times those for other CWFs. The density is 2.42 g/cm3. This will be monitored with time to probe for macroscopic swelling. An accelerated alpha damage study is under way with sample production initiated in October 1998. The majority of the test matrix has been accomplished for the early time period samples. JANUARY 29, 1999 Closed Session Attendance: G. Choppin (chair), M. Apted, P. Baisden, E. Flanigen, C. Hussey, F. Mansfeld, L. E. McNeese, R. Osteryoung, P. Shewmon, R. White, C. Murphy. The entire meeting was conducted in closed session. Following a preliminary discussion of committee balance and composition, a detailed review of the previous day's presentations took place. The committee then discussed writing assignments and generated findings and conclusions for this report. Reviewers' comments for the committee's report 8 were also discussed.1

1National

Research Council, Electrometallurgical Techniques for DOE Spent Fuel Treatnent: Status Report on Argonne National Laboratory's R&D Activity as of Fall 1998, National Academy Press, Washington, D.C., 1999.

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APPENDIX D

49

Appendix D Meeting Summary, March 15, 1999

Location: National Academies Building, Washington, D.C. Attendance: G. Choppin (chair), M. Apted, P. Baisden, E. Flanigen, C. Hussey, F. Mansfeld, E. McNeese, R. Osteryoung, P. Shewmon, D. Raber, C. Murphy. Unable to attend: R. White. Speaker: Robert Benedict, Argonne National Laboratory-West. EXECUTIVE SESSION 7:30 am

Committee Breakfast and Preliminary Committee Discussions

OPEN SESSION 8:00 am

ANL's Spent Fuel Treatment Demonstration Project Status—Robert Benedict, ANL-W

EXECUTIVE SESSION 9:30 am

Waste Forms Report (Report 9) Breakout Session

11:00 am

Review of Report 9 Sections: Introduction (Choppin, Murphy)

11:30 am

Background (Apted, McNeese)

12:00 pm

Lunch

1:00 pm

Review of Report 9 Sections: Metal Waste Form (Mansfeld, ussey, Baisden, and Shewmon)

1:30 pm

Ceramic Waste Form (Flanigen, Osteryoung, White)

2:00 pm

Other Considerations/Conclusions & Recommendations (Committee)

3:00 pm

Report 8 (Fall '98 Status Report) Reviewers' Comments

4:30 pm

Final Discussions/Wrap-up

5:00 pm

Adjourn

Robert W. Benedict gave a brief overview of plans for proposed final reports for the spent fuel treatment demonstration. The overall demonstration reports will include the spent fuel treatment demonstration final report, a report on waste form production from the electrometallurgical treatment of sodium-bonded spent nuclear fuel, and a report on the analysis of spent fuel treatment demonstration operations.

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APPENDIX D

50

Treatment-operation reports will cover three areas overall—treatment reports, driver treatment reports, and blanket treatment reports. The overall treatment report will discuss the development of the cathode processor and casting furnace operating conditions. Driver treatment reports will cover the process description for driver fuel treatment operations, and development of the electrorefining process for driver fuel. Blanket treatment reports will include a report on process description for blanket treatment operations, and a report discussing development of the electrorefining process for blanket fuel. Waste operations and qualifications reports will cover three areas—overall waste reports, ceramic waste reports, and metal waste reports. Included in the overall waste reports will be reports on the waste form qualification strategy, the waste acceptance product specifications, the waste compliance plan, and a report on waste form degradation and repository performance modeling. Ceramic waste reports will include the ceramic waste form process qualification plan, and the ceramic waste form handbook. Metal waste reports will include the metal waste form process qualification plan, and the metal waste form handbook. Following Robert Benedict's presentation and questions from the committee, the committee entered closed session. In the closed session, sections of the waste forms report prepared by the committee were discussed, and a new draft of the report was generated. These discussions were followed by analysis of the reviewers' comments for the committee's Report 8. Before adjourning, the committee discussed plans for its upcoming meeting in Idaho Falls, ID, in July 1999.

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APPENDIX E 51

Appendix E

ANL-DOE Interactions Related to Waste Forms Generated by ANL's EMT Program

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APPENDIX E

52

ARGONNE NATIONAL LABORATORY - WEST P.O. Box 2528, Idaho Falls, Idaho 83403-2528Telephone: (208) 533-7106 May 7, 1999 Dr. Chris Murphy National Research Council 2101 Constitution Avenue Washington, DC 20418 Dear Chris:

As we discussed at the January 1999 Electrometallurgical review meeting, over the last two years Argonne's interactions with the Department of Energy (DOE) programs associated with the geological repository and waste form qualification have increased in order to better ensure that the high-level waste (HLW) generated from the electrometallurgical treatment (EMT) of sodium-bonded fuels are acceptable for disposal. These interactions have generally fallen into three categories: Yucca Mountain Repository Environmental Impact Statement, National Spent Nuclear Fuel Program, and Electrometallurgical Specific Meetings. The first series of interactions (Attachment 1) concerns the preparation of the Yucca Mountain Repository Environmental Impact Statement (EIS). After discussions with the National Spent Nuclear Fuel Program (NSNFP), Idaho National Engineering and Environmental Laboratory (INEEL) and DOE Environmental Management personnel, the electrometallurgical wastes were included as high level waste streams. Argonne participated in the data call and review of this document. The formal data submittals for the EMT waste forms from treating 60 MTHM of sodium-bonded fuel were done through the INEEL HLW office and DOE-EM. The EMT waste streams are included in Appendix A of the Yucca Mountain Repository EIS. Argonne also participates in regular meetings with personnel of the National Spent Nuclear Fuel Program and the INEEL Spent Nuclear Fuel Program. The purpose of the NSNFP is to determine the activities necessary to best integrate DOE-owned spent nuclear fuel into the repository. Issues addressed include disposal requirements, data needs, interfaces for standardized canisters, material shipments, and quality assurance programs. The INEEL Spent Nuclear Fuel Program is responsible for the activities for the specific DOE fuels that are stored at the Idaho Nuclear Technology and Environmental Center (INTEC). INTEC currently has 2.0 MT of Experimental Breeder Reactor-II (EBR-II) fuel stored and is the proposed location of the final packaging for

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APPENDIX E

53

shipment of DOE stainless steel clad fuels and the EMT wastes to the repository. Argonne has regular interactions with both groups as detailed in Attachment 2. Many of these meetings also include DOE Office of Civilian Radioactive Waste Management (RW) and DOE-EM personnel. Through these earlier mentioned interactions, Argonne has now started more direct meetings with personnel working on the HLW programs within DOE-EM and with DOE-RW and contractors supporting the repository. These meetings (Attachment 3) are to help ensure that Argonne is generating the data and meeting the criteria required for disposal of the EMT HLW. Through these meetings and interactions, we feel that we are keeping up to date with the waste acceptance criteria and performance assessment issues for the proposed geologic repository. As stated by Yucca Mountain Project personnel at the recent NSNFP strategy meeting, the waste criteria are well established and the future changes should primarily be a refinement of the requirements. Their opinion is the waste acceptance criteria are well enough established that waste producers should be able to proceed without major technical risks. If you need any additional information or have further questions, please call me at 208-533-7166. Cordially, Robert Benedict Director, Nuclear Technology Division

RWB:ljc Attachments: As stated. cc: J. P. Ackerman Y. I. Chang CF RWB.49

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APPENDIX E

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Attachment 1: Repository EIS Meetings June 9, 1997 – DOE-EM (NSNFP) Met with D. Fillmore to determine method of data submittal for Yucca Mountain Environmental Impact Study (EIS).

June 25, 1997 – DOE-EM (INEEL HLW personnel) Met to coordinate HLW data submittal.

June 26, 1997 – DOE-EM (INEEL HLW personnel) Met to coordinate HLW data submittal.

June 27, 1997 – DOE-EM (INEEL HLW personnel) Met to coordinate HLW data submittal.

July 1, 1997 – DOE-EM Met with HLW producers and DOE-EM to discuss data submittal for No-Action Alternative for Yucca Mountain EIS.

July 2, 1997 – DOE-EM (INEEL HLW personnel) Met to coordinate HLW data submittal. Joint submittal letter signed and sent to T. Wichmann. Data eventually submitted to DOE-RW through DOE-EM (K. Picha).

August 19, 1997 – DOE-RW and DOE-EM Video conference call to discuss data call for the No-Action Alternative.

August 20, 1997 – DOE-RW and DOE-EM Conference call to discuss data call for the No-Action Alternative. Appointed to the Working Group to assess assumptions used in the EIS, especially in relationship to the No-Action Alternative.

After its appointment to the Working Group, ANL has become a participant in the bi-weekly teleconference for the Yucca Mountain EIS preparation. This meeting includes individuals from DOE-EM (HLW and NSNFP) and DOE-RW. We are still participating in these meetings. September 18, 1997 – DOE-RW and DOE-EM Conference call to discuss data call for the No-Action Alternative.

November 17, 1997 – DOE-EM (INEEL HLW personnel) Data submittal for No-Action Alternative provided to INEEL HLW personnel. Forwarded to DOE-RW through DOE-EM.

December 15, 1997 – DOE-EM (INEEL HLW personnel)

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APPENDIX E

Cost data submittal for No-Action Alternative provided to INEEL HLW personnel. Forwarded to DOE-RW through DOE-EM.

March 19, 1998 – DOE-RW and DOE-EM Conference call to discuss the review of Yucca Mountain EIS Appendix A (Inventory).

March 1998 Reviewed Appendix A (Inventory) of the Yucca Mountain EIS. Comments submitted to DOE-RW through DOEEM (HLW).

March 31, April 1-2, 1998 – DOE-RW and DOE-EM EIS review meeting in Las Vegas.

April 15, 1998 – DOE-EM (INEEL HLW personnel) Data revision submitted for Yucca Mountain EIS to INEEL HLW personnel. Forwarded to DOE-RW through DOEEM.

May 20, 1998 – DOE-EM (INEEL HLW personnel) Data revision submitted for Yucca Mountain EIS to INEEL HLW personnel. Forwarded to DOE-RW through DOEEM.

November 1998 Reviewed Draft Yucca Mountain EIS. Comments submitted through INEEL HLW to EM (HLW) and then to RW. INEEL HLW and DOE-EM personnel represented us at a comment review meeting in December 1998.

Attachment 2: Spent Nuclear Fuel Program Meetings May 9, 1997 – DOE-EM (NSNFP) Reviewed issues associated with disposal of DOE SNF.

June 24-25, 1997 – DOE-EM (NSNFP) Attended NSNFP strategy meeting.

January 8-9, 1998 – DOE-EM (NSNFP) Attended NSNFP meeting on DOE-RW waste package internal components and High Integrity Canister.

January 12, 1998 – DOE-EM (NSNFP) Attended NSNFP workshop on standardized canister.

January 13-14, 1998 – DOE-EM (NSNFP) Attended NSNFP Strategy Meeting.

March 1998 to Present – DOE-EM (NSNFP)

55

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APPENDIX E

We have regularly scheduled conference calls with individuals of the NSNFP to discuss issues associated with the electrometallurgical treatment of spent nuclear fuel.

June 2, 1998 – DOE-EM (NSNFP) Attended NSNFP meeting on design basis events.

June 3, 1998 – DOE-EM (NSNFP) Attended NSNFP meeting on repository criticality safety.

June 4, 1998 – DOE-EM (NSNFP) Attended NSNFP meeting on total system performance assessment.

October 27-28, 1998 – DOE-EM (NSNFP) Attended NSNFP strategy meeting.

November 17-18, 1998 – DOE-EM (NSNFP) Participated in a meeting to discuss grouping of DOE SNF for disposal in a repository.

March 18, 1999 – DOE-EM (NSNFP) Attended NSNFP planning meeting.

April 26-28, 1999 – DOE-EM (NSNFP) Attended NSNFP strategy meeting.

May 5-6, 1999 – DOE-EM (NSNFP) Attended NSNFP transportation workshop.

DOE-EM (NSNFP) Ongoing Review Work: Viability Assessment Draft B Standardized Canister Documentation

Waste Acceptance System Requirements Document Revision 2 and 3A Attachment 3: DOE High-Level Waste Programs and DOE-RW Meetings June 17, 1997 – Performance Assessment Modelers from Sandia Met with PA personnel working for the NSNFP.

September 8, 1997 – DOE-EM and DOE-RW Discussed Baseline Change Proposal with DOE-EM personnel assigned to Yucca Mountain (T. Gunter). Noted that change is probably not needed until after the demonstration.

January 12, 1998 – DOE-RW

56

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APPENDIX E

Discussed ANL performance assessment models with Yucca Mountain Project personnel.

April 20, 1998 – Performance Assessment Modelers from Sandia Met with PA personnel working for the NSNFP.

April 22, 1998 – DOE High-Level Waste Steering Committee Hosted committee at ANL-West.

May 29, 1998 – Savannah River Personnel Met with personnel of SRS to discuss waste form and process qualification issues.

June 10-11, 1998 – Advisory Committee on Nuclear Waste (ACNW) Attended ACNW meeting on waste form degradation modeling.

June 24, 1998 – Nuclear Waste Technical Review Board (NWTRB) Attended NWTRB Summer meeting.

June 26-27, 1998 – DOE-RW Discussed ANL performance assessment models with Yucca Mountain Project personnel.

July 14, 1998 – Nuclear Waste Technical Review Board Hosted Dan Bullen and Carl Di Bella from the Nuclear Waste Technical Review Board.

January 19, 1999 – DOE-EM and Sandia Personnel Met to discuss documentation for next repository license application.

January 25, 1999 – NWTRB Attended NWTRB Repository Panel Meeting.

January 26-27, 1999 – NWTRB Attended NWTRB Panel Meeting on the Viability Assessment.

February 2-4 1999 – DOE-EM and DOE-RW Attended Features, Events, and Processes (FEPs) workshop for PAs.

February 11, 1999 – DOE-RW and DOE-EM Video conference call to discuss waste forms from electrometallurgical treatment of spent nuclear fuel.

April 14, 1999 – DOE-RW and DOE-EM Conference call to discuss HLW quality assurance program, waste form qualification strategy including document preparation, and data submittal for repository license application.

57

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APPENDIX E 58

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APPENDIX E

59

Appendix F Abbreviations and Acronyms Used in the Main Text of This Report AES ANL ANL-E ANL-W ASTM CWF DHLW DOE DOE-EM DOE-NE DOE-RW EBR-II EIS EMT EPA HEU HIP HLW HUP LEU LLW MCC MGDWAC MIC MOA MWF NE PCT PLS R&D RW SEM SNF SS TCLP TEM TRU WIPP XPS

Auger electron spectroscopy Argonne National Laboratory Argonne National Laboratory-East (Argonne, Illinois) Argonne National Laboratory-West (Idaho Falls, Idaho) American Society for Testing and Materials ceramic waste form Defense high-level waste U.S. Department of Energy U.S. Department of Energy Office of Environmental Management U.S. Department of Energy Office of Nuclear Energy U.S. Department of Energy Office of Civilian Radioactive Waste Management Experimental Breeder Reactor-II Environmental Impact Statement electrometallurgical technology U.S. Environmental Protection Agency highly enriched uranium hot isostatic pressing high-level waste hot uniaxial pressing low enriched uranium low-level waste Material Characterization Center Mined Geologic Disposal Waste Acceptance Criteria microbially influenced corrosion Memorandum of Agreement metal waste form U.S. Department of Energy Office of Nuclear Energy product consistency test pressureless sintering research and development U.S. Department of Energy Office of Civilian Radioactive Waste Management scanning electron microscopy spent nuclear fuel stainless steel toxicity characteristic leaching procedure transmission electron microscopy transuranic waste isolation pilot plant x-ray photoelectron spectroscopy

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APPENDIX E

XRD

60

x-ray diffraction