Appendix E
Abstracts of Invited Workshop and Poster Presentations

This appendix contains a collection of abstracts of individually authored background papers and posters that were presented at the May 13-15, 1996, Glass as a Waste Form and Vitrification Technology: An International Workshop, sponsored by the Board on Radioactive Waste Management of the National Research Council. These abstracts, intended solely for discussion at the workshop, have not been reviewed or approved by the National Research Council.



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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Appendix E Abstracts of Invited Workshop and Poster Presentations This appendix contains a collection of abstracts of individually authored background papers and posters that were presented at the May 13-15, 1996, Glass as a Waste Form and Vitrification Technology: An International Workshop, sponsored by the Board on Radioactive Waste Management of the National Research Council. These abstracts, intended solely for discussion at the workshop, have not been reviewed or approved by the National Research Council.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Contents Invited Presentations     Vitrification of Radioactive Waste: Past Accomplishments and Future Challenges, Werner Lutze, University of New Mexico   E6     Overview of Current DOE Plans and Activities Involving Vitrification, Stephen P. Cowan, U.S. Department of Energy   E8     Identification and Summary Characterization of Materials Potentially Requiring Vitrification, Allen G. Croff, Oak Ridge National Laboratory   E10     Hanford Wastes and Glass Composition, Pavel Hrma, Pacific Northwest National Laboratory   E12     Acceptance of Waste for Disposal in the Potential United States Repository at Yucca Mountain, Nevada, David Stahl, Framatome Cogema Fuels, Civilian Radioactive Waste Management System, M&O Contractor   E14     A Review of Status of Science of Vitrified Waste Form Development, George G. Wicks, Westinghouse Savannah River Technology Center   E16     Leach Tests and Chemical Durability, Robert H. Doremus, Rensselaer Polytechnic Institute   E18     Status of Vitrification Technologies, M. J. Plodinec, Westinghouse Savannah River Company   E20     The Chemistry and Kinetics of Waste Glass Corrosion, John K. Bates, Argonne National Laboratory   E22     Waste Glass Leaching and Long-Term Modeling, W. L. Bourcier, Lawrence Livermore National Laboratory   E24     Waste Glass Leaching and Long-Term Durability, Etienne Y. Vernaz, Commissariat à l'Energie Atomique, France   E26     Corrosion Behavior of Glass: Remaining Scientific Issues, B. Grambow, Forschungszentrum Karlsruhe, Germany   E28     Natural Glasses and the Verification of the Long-Term Durability of Nuclear Waste Glasses, Rodney C. Ewing, University of New Mexico   E30     Actinide Vitrification: Status of Savannah River Site Activities, William G. Ramsey, Savannah River Technology Center   E32     Thermal Stability of Waste Form Glass, N. Jacquet-Francillon, Commissariat à l'Energie Atomique, France   E34

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop     Mechanical Properties of the Waste Form Glass, Hj. Matzke, European Commission, Institute for Transuranium Elements, Germany   E36     Radiation Effects in Glass Waste Forms, William J. Weber, Pacific Northwest National Laboratory   E38     Experience in Belgium, M. Demonie, Belgoprocess, Belgium   E40     Vitrification Experience in France—Development and Perspectives, Antoine Jouan, Commissariat à l'Energie Atomique, France; Thierry Flament, Société Générale pour les Technologies Nouvelles, France; Hugues Binninger, COGEMA, France   E42     Experience in the United States: West Valley, New York, Victor A. DesCamp, West Valley Nuclear Service Company, Inc.   E44     Vitrification Experience of TVF in Japan, M. Yoshioka and Hiroshi Igarashi, Power Reactor and Nuclear Fuel Development Corporation, Japan   E46     Vitrification Experience in the UK, Graham A. Fairhall and Charles R. Scales, British Nuclear Fuels plc   E48     Experiences with Vitrification HL W and Development of New Approaches in Russia, A. S. Aloy, RPA, V. G. Khlopin Radium Institute, Russia; V. A. Bel'tyukov, PA ''Mayak,'' Russia; A. V. Detain, VNIIPIET, Russia; and Yu. A. Revenko, Mining and Chemical Association, Russia   E50     Vitrification Experience in China, Wang Xian de, Beijing Institute of Nuclear Engineering, China   E52     Vitrification Experience at the Defense Waste Processing Facility (DWPF), David B. Amerine, Westinghouse Savannah River Company   E54 Poster Presentations (Alphabetical, First Author)     Natural Glasses as Analogs for Nuclear Waste Glasses, Abdesselam Abdelouas and Werner Lutze, University of New Mexico   E56     Alteration of Nuclear Waste Glasses Characterized by Radon Emanation Method, V. Balek and Málek, Nuclear Research Institute Rez plc, and A. Clearfield, Texas A&M University   E58     DOE Regulatory Initiative For Vitrified Mixed Waste, Sandra J. Carroll, Westinghouse Savannah River Company, and James E. Flaherty, Science Applications International Corporation   E60     Stabilization of Plutonium in Hybrid Glass Materials Using a Cyclone Melter, N. V. Coppa, J. C. Simpson, and J. G. Hnat, Vortec Corporation   E62     DOE Regulatory Initiative: Immobilized Mixed Debris Proposal, Barbara Dubiel, Lockheed Idaho Technologies Company, and Susan Carson, Sandia National Laboratories   E64

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop     Reexamination of Variables Affecting the Hydration of Obsidians Using Empirical Chronological Data and Laboratory Induced Hydration Experiments, Jonathon E. Ericson and Stephen R. Lyon, University of California, Irvine   E66     A Plasma Arc-Vitreous Ceramic Process for Hazardous and Radioactive Waste Stabilization, Xiangdong Feng and Jeffrey E. Surma, Pacific Northwest National Laboratory; Clarence G. Whitworth, MSE, Inc.; Richard C. Eschenbach, Retech of Lockheed Martin; and Gary L. Leatherman, Science Applications International Corporation   E68     Direct Conversion of Metals, Ceramics, Amorphous Solids, Halogens, and Organics to Borosilicate Glass Using the GMODS Process, Charles W. Forsberg and Edward C. Beahm, Oak Ridge National Laboratory   E70     Surface Layer Formation of the French SON68 Nuclear Waste Glass During Vapor Phase Alteration at 200° C, W. L. Gong, R. C. Ewing, and L. M. Wang, University of New Mexico; E. Vernaz, Commissariat à l'Energie Atomique, France; J. K. Bates and W. L. Ebert, Argonne National Laboratory   E72     Potential Application of Electron Spin Resonance to the Study of Radiation Effects in High-Level Nuclear Waste Disposal Glasses, David L. Griscom, Department of the Navy   E74     Leaching Behavior of Pu and Cm From Waste Glass Under Reducing Condition, Y. Inagaki, A. Sakai, H. Furuya, K. Idemitsu, and T. Arima, Kyushu University, Japan; T. Banba, T. Maeda, S. Matsumoto, and Y. Tamura, Japan Atomic Energy Research Institute   E76     Glass and Glass-Ceramic Waste Forms Developed at the Idaho Chemical Processing Plant for Immobilizing HL W and Actinides, Dieter A. Knecht, Tom P. O'Holleran, Krishna Vinjamuri, Swami V. Raman, and Bruce A. Staples, Lockheed Martin Idaho Technologies   E78     Reproduction of Natural Corrosion by Accelerated Laboratory Testing Methods, J. S. Luo, J. J. Mazer, D. J. Wronkiewicz, and J. K. Bates, Argonne National Laboratory   E80     Colloid Formation During Waste Glass Corrosion, C. J. Mertz, E. C. Buck, J. A. Fortner, and J. K. Bates, Argonne National Laboratory   E82     Use of DC Graphite Arc Melter Technology For Production of Stable Vitrified Waste Forms, T. J. Overcamp and D. L. Erich, Clemson University; J. K. Wittle and R. A. Hamilton, Electro-Pyrolysis, Inc.; and P. J. Wilver, Svedala Industries, Inc.   E84     Hanford Low Level Waste Melter Tests, Ian Pegg, Pedro Macedo, Keith Matlack, Hamid Hojaji, and Shi-Ben Xing, Catholic University of America; Jaqueline Ruller and William Greenman, GTS Duratek   E87     Experimental Determination of Uranium Oxide Solubility in Hydrous Silicate Melts of Granitic Composition, Chantal Peiffert and Michel Cuney, CREGU and GdR, France   E88     Mixture Models Versus Free Energy of Hydration Models for Waste Glass Durability, Greg Piepel and Trish Redgate, Pacific Northwest National Laboratory   E90

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop     Microstuctures and Leach Rates of Glass-Ceramic Waste Forms for Immobilizing Plutionium and Associated Components, Swami V. Raman, Lockheed Martin Idaho Technologies   E92     Hot Isostatic Press (HIP) Vitrification of Radwaste Concretes, Darryl Siemer; Barry Scheetz, Pennsylvania State University; and Cliff Orcurt, American Isostatic Press   E94     Waste Glass Leaching During Open Site Tests, Igor A. Sobolov, Alexander S. Barinov, Michael I. Ojovan, and Natalya V. Ojovan, Scientific and Industrial Association "Radon," Russia   E96     In Situ Vitrification of Plutonium and Uranium Contaminated Buried Wastes: Microcompositional Analyses of Vitreous and Crystalline Phases and Corresponding Leach Test Results of the Vitrified Products, Leo E. Thompson, Dale M. Timmons, and Jack L. McElroy, Geosafe Corporation   E98     The Interaction Between HL W Glass and Clay: Present Status and Future Programme, Pierre Van Iseghem, Karel Lemmens, Marc Aertsens, Philippe Lolivier, Wei Jiang, and Pierre De Cannière, SCK·CEN, Belgium   E100     Verification Studies on the Pamela High-Level Waste Glasses, P. Van Iseghem, E. Hoskens, G. Smeyers, D. Huys, L. Sannen, and L. Vandevelde, SCK·CEN, Belgium   E102     Synroc - An Alternative Waste Form, E. R. Vance, K. P. Hart, and A. Jostsons, Australian Nuclear Science and Technology Organization   E104     Chemistry of Irradiated Solids, D. W. Werst and A. D. Trifunac, Argonne National Laboratory   E107     CORAL US: In-Situ Corrosion Test on Active HLW Glass, P.Van Iseghem, SCK·CEN, Belgium; E. Vernaz, Commissariat à l'Energie Atomique, France; and N. Jockwer, GRS, Germany   E108

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Workshop on Glass as a Waste Form and Vitrification Technology Washington, May 13-15, 1996, Washington, D. C. Vitrification of Radioactive Waste: Past Accomplishments and Future Challenges Werner Lutze Center for Radioactive Waste Management (CeRaM) and Department of Chemical and Nuclear Engineering The University of New Mexico Albuquerque, NM 87131 Abstract The history of vitrification of HLW is reviewed. The development of various vitrification technologies, pretreatment strategies, and the focus on two types of melters, the metallic and the ceramic melter, are explained. Various vitrification plants are in operation worldwide with France being the leader where three plants are in operation, AVM at Marcoule, R7, and T7 at La Hague. R7 and T7 are commercial plants. Another commercial vitrification plant, based on the French process, is in operation in Sellafield, England. Very recently, the first vitrification plant for defense waste was put into operation in the United States at the Department of Energy facility at Savannah River, S. C. Another vitrification plant in the U.S. will go into operation at West Valley Nuclear Services, N. Y. later this year. Experience with both melter types in hot operations (France, England and Germany) and lessons learned are highlighted. Research and development toward a new melter type, the cold crucible, is addressed and the potential advantages over existing melters are outlined. The main drivers for this effort are 1) significantly decreased melter corrosion and thus increased life time and operation safety, 2) greater flexibility in the melting temperature, and 3) greatly reduced contamination of the melter and less melter waste. For an overview of waste form development and vitrification technologies see reference 1 and for glass corrosion reference 2. The development and selection of waste forms, glass vs. crystalline materials, is addressed and the selection criteria are discussed. Borosilicate glasses has become the most widely used waste form for high-level radioactive reprocessing waste for both defense and commercial waste. However, large quantities of high-level waste are immobilized in phosphate glass in Russia. New glass waste forms are under development in the United States to vitrify and to dispose of surplus weapons plutonium and other waste streams high in actinides. Alternatively, a variety of ceramics is under investigation in Russia and in the United States.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Vitrified waste must meet certain waste acceptance criteria for disposal. These criteria are derived from glass properties measured in the laboratory. Glass properties are reviewed and the state of the art is presented. Glass properties are ranked in terms of significance for glass performance in the repository. Areas for further research, e.g., long-term radiation effects and long-term chemical durability and the use of natural glasses as analogs are pointed out. The dependence of glass performance on the geochemical (repository) environment is addressed. Finally, areas of research on alternative glass processing are addressed, for example sub-liquidus processing. The wide variety of existing waste streams calls for various types of pretreatment prior to solidification. The choice of the waste form and the process to make the waste form are important considerations for the design of waste pretreatment procedures. Glass and the melting process are not always the best choice no matter how much experience has been acquired with vitrification. The limits for glass as a waste form are demonstrated using particular waste streams as examples. References 1. Radioactive Waste Forms for the Future, W. Lutze and R. C. Ewing, eds., North Holland, Amsterdam 1988, 778p. 2. High-Level Waste Borosilicate Glass, A Compendium of Corrosion Characteristics, 3 volumes, DOE-EM-0177, 1994

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop OVERVIEW OF CURRENT DOE PLANS AND ACTIVITIES INVOLVING VITRIFICATION Summary of Presentation to: National Academy of Sciences Vitrification Workshop by Stephen P Cowan, U. S. Department of Energy The Department of Energy must address the legacy of over fifty years of nuclear weapons research, development, and production. Wastes with varying levels of a broad range of radionuclides and hazardous chemical constituents must be managed to minimize: current risks to workers; future risks to the environment and potential surrounding populations; and the costs to the nation. Newly generated wastes, and wastes in storage, are the responsibility of the Office of Waste Management. The fundamental problem the Department must address is that the wastes to be managed are not currently in the correct form or location for permanent isolation. Therefore, the waste must be treated. The question is: "What would be the ideal process for this treatment?". The ideal process would result in a reduced volume of waste in a stable, durable solid. The process should be relatively simple - suitable for remote operations in a radioactive environment, adaptable to a wide variety of feeds, and capable of destroying or immobilizing the hazardous constituents of concern as well as the radioactivity over the time period during which an unacceptable risk is posed. The vitrification process, and the glass waste form, are the best candidates the Department, and indeed other nations, have found. A brief history: Beginning in the 1940's, the Manhattan Project scientists knew that tank storage of liquid wastes was a temporary measure. The Atomic Energy Commission began examining immobilization and disposal options in the 1960's, and by the 1970's specific waste forms were being evaluated. Options examined included stabilizing the wastes in the tanks or making grout for underground injection beneath the sites where the wastes were stored. But the option that seemed consistently most appealing was the removal of the wastes from the tanks, separation of the high activity fraction from the low activity fraction, and immobilization of the high activity fraction in a stable, leach resistant form. Waste forms evaluated included concrete, a variety of glass formulations, and ceramics. Two criteria were of premier importance - the durability of the waste form and the simplicity of the process to enable remote operations in a radioactive environment. Through a formal Departmental process, the Department selected borosilicate glass for high level waste immobilization. A number of vitrification processes for low level, mixed, and environmental restoration wastes are in various stages of evaluation, demonstration, or implementation. Implementation of the vitrification waste form initially focused on high level waste (HLW) because it contains the largest inventory of curies. Of the four sites with HLW, the Savannah River Site has the most curies, followed by the waste stored at Hanford. The waste at both of these sites, as well as most of the volume at the West Valley site, is alkaline liquid stored in carbon steel tanks. The HLW at the Idaho National Engineering Laboratory is stored both as

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop an acidic liquid in stainless steel tanks and as calcined solids in stainless steel bins, with more than 90% of the radioactivity in the calcine. The total inventory of DOE HLW is approximately 950 million curies. Vitrification has been recognized by the Environmental Protection Agency as the "best demonstrated available technology" for HLW. Glass is also the waste form for immobilizing HLW in France, the United Kingdom, Belgium, and Japan. The Defense Waste Processing Facility at the Savannah River Site, initiated radioactive activities on March 12, 1996; the first canister of radioactive glass is scheduled to be produced in the very near future. Initiation of radioactive activities at the West Valley Demonstration Project are imminent, with production of the first canister of radioactive glass scheduled for late June. For the waste at the INEL, DOE is proposing a "Full Treatment Alternative" for integrating all the waste streams managed by the Office of Environmental Management. This alternative includes operation of a vitrification/ separations facility after 2017 which will process remote-handled transuranic waste, as well as the liquid HLW and the high-activity fraction of the calcine. DOE is also planning to use vitrification technology for low level wastes at the Fernald site, Mound, and at Hanford. Wastewater treatment sludges containing both hazardous constituents and radioactivity (i.e., mixed wastes) will be vitrified for disposal at Oak Ridge, Los Alamos National Laboratory, the Rocky Flats Environmental Technology Site, and the Savannah River Site.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop IDENTIFICATION AND SUMMARY CHARACTERIZATION OF MATERIALS POTENTIALLY REQUIRING VITRIFICATION1 Allen G. Croff Oak Ridge National Laboratory2 The United States and many other nations have and continue to produce a wide variety of radioactive materials that are wastes or may otherwise be declared to be surplus and be managed as if they were wastes. In either case, it is commonplace to subject these materials to treatment processes to appropriately condition them for subsequent storage, transportation, and disposal. Less toxic wastes are generally accorded minimal treatment whereas highly toxic wastes are accorded extensive treatment. One of the more common treatment technologies applied to the more toxic end of the waste spectrum is vitrification. The purpose of this paper is to identify those materials that have some potential for vitrification and to summarize their characteristics as background for subsequent papers. Identification of materials having some potential for vitrification begins by considering the broad categories of existing nuclear wastes or materials that may be declared surplus, which are listed in the first column of Table 1. Uranium mill tailings and enrichment plant tails are excluded because their large volume and low toxicity result in essentially no potential for vitrification. The next three columns of Table 1 provide summary information concerning the inventory and/or production rate of the materials, the radioactivity density, and the power (radioactive heat) density. The densities can be viewed as crude measures of the toxicity of the material. The trend of these numbers reflects a longstanding paradigm for management of radioactive materials that results in the high radioactivity/power materials being concentrated in relatively small volumes and conversely for the low radioactivity/power materials. In general, as the volume of the materials increase, the practicality of applying advanced treatment technologies such as vitrification decreases. Further, as the radioactivity and power density of the material increases, the appropriateness of a process such as vitrification that yields a high-integrity product increases. The fifth column of Table 1 provides a very brief description of the subject material. In some cases, the material is relatively uniform (e.g., LWR spent fuel, Cs/Sr capsules, surplus Pu). In other cases, the nature of the material in a waste category is extremely diverse ranging from liquids to soil to contaminated rabble to trash (e.g., transuranic waste, low-level waste). The above considerations strongly affect the extent to which various material types or sub-types are considered to be candidates for vitrification. However, at least as important as the technical considerations given above are precedent, current Federal policies, and stakeholder views (e.g., agreements negotiated between a state and the U.S. Department of Energy). As a result, combining all of these to arrive at a conclusion of the vitrification potential of a particular material is subjective. The author's views concerning the vitrification potential of the material types and selected sub-types is given in the last column of Table 1. 1   The submitted manuscript has been authored by a contractor of the U.S. Government under contract No. DE-AC05-96OR22464. Accordingly, the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or to allow others to do so, for U.S. Government purposes. 2   Managed by Lockheed Martin Energy Research Corp. for the U.S. Department of Energy under contract No. DE-AC05-96OR22464.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop TABLE 1. SUMMARY OF U.S. MATERIALS HAVING SOME POTENTIAL TO BE VITRIFIED Material Type Volume, m3 or m3/yra Radioactivity Density, Ci/m3 Power Density, W/m3 Material Description Vitrification Possibilities 1. Spent civilian nuclear fuel 12000 10,000,000 50000 Light-water reactor spent fuel Unlikely unless required by repository 2. DOE spent fuel 1200 Not quantifiable; Moderate-to-high Not quantifiable; Moderate-to-high Variety of spent fuels Likely for Al-clad fuels, possible for others 3. DOE "tank" wastes 375000 1,000 - 10,000 5 - 50 Alkaline liquid, saltcake, sludge; calcine Highly likely for essentially all retrieved tank waste 4. Capsules: Cs Sr 3.5 1.1 23,000,000 21,000,000 115,000 140,000 Capsules of CsCl Capsules of SrF2 Likely if overpack is unacceptable 5. Transuranic wastes       Wide variety of materials with TRU >100 nCi/g Likely for only a small fraction unless WIPP-WAC change substantially Remotely handled 2,500 + 14/yr 1,000 1 - 2     Contact handled 70,000 + 1500/yr 25 - 50 0.5 - 1.5     6. Low-level radioactive waste       Extremely wide variety of materials with <<100 nCi/g Likely for LLW from tank waste processing. Unlikely for most other LLW. DOE 38,000/yr 9 - 27 0.01 - 0.05     Commercial: Class A   0.6 0.03 - 0.1     Commercial: Class B 24,000/yb 60 15     Commercial: Class C   0.1 - 7,000 0.003 - 115     Commercial: > Class C 63 + 20/yr >0.1 - high >0.003 - high     7. Low-level mixed waste       Extremely wide variety of materials with <100 nCi/g and hazardous chemicals Likely in selected applications, but extent is unpredictable Commercial 2,100 Not quantifiable; low Not quantifiable; low     DOE 138,000         8. Surplus plutonium 2 11,000,000 44,000 Plutonium metal shapes Either vitrification or irradiation will be used 9. Environmental restoration 78,000,000 Not quantifiable; low with small-volume exceptions Not quantifiable Low with small-volume exceptions Extremely wide variety of materials and contamination High-toxicity wastes and some in-situ are likely. Unlikely for the bulk of the waste. a Fixed values are existing volumes which are given where production has essentially ceased or where disposal rates are approximately equal to production rates. Rates are given where volumes continue to increase significantly. b Sum of annual production rates for Classes A, B, and C.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop THE INTERACTION BETWEEN HLW GLASS AND CLAY: PRESENT STATUS AND FUTURE PROGRAMME. Pierre Van Iseghem, Karel Lemmens, Marc Aertsens, Philippe Lolivier, Wei Jiang, Pierre De Cannière Waste and Disposal, SCK·CEN, Boeretang 200, 2400 Mol, Belgium We have completed a large parametric study to identify the role of different engineered (container corrosion products, backfill materials) and natural (Boom clay) barrier materials on the dissolution behaviour of high-level waste glass. We considered the Cogéma R7T7 and the DWK/ Pamela glasses. This way we obtained specific information on the influence of the Al2O3 concentration in the glass. Other parameters were the temperature and the presence of a gamma radiation field. Different SA/V (surface area to solution volume) were applied as a means to accelerate the glass dissolution. By combining high SA/V conditions and interacting durations up to 2 years we reached high reaction progress values (up to 2200 y/m). The leaching behaviour of the long-living radionuclides was studied, through glasses doped with the radionuclides of interest (Pu-239, Am-241, Np-237, Tc-99), or fully active glass samples. In-situ tests were carried out to verify the results obtained in the laboratory. As a basic conclusion, we identified some main dissolution mechanisms: congruent glass dissolution in (mainly) clay media; we anticipate that the congruent dissolution is not permanent. ion exchange controlled dissolution at high reaction progress; the ion exchange between H3O+ / H+ and Na+/Li+ was proposed. Besides, discontinuous excursions in the glass dissolution due to secondary phase formation were observed in a number of situations, after the saturation concentration of certain elements was reached. We also obtained qualitative information on the influence of the presence of iron corrosion products and of bentonite or cement backfill. As a consequence of our new insight on the basic dissolution mechanisms we decided to develop a mathematical model for the glass dissolution, accounting for the ion exchange controlled dissolution. The model also considers the transport properties of SiO2 through the pore water in clay. Special migration tests were designed using Si-32 tracer. The in-situ tests were performed in the underground laboratory in clay beneath the SCK·CEN laboratory. We obtained data at temperatures of 16°C (rock temperature), 90°C and 170°C, for maximum durations of 7.5 years. At higher temperature the corrosion data were similar to the data from the laboratory tests. We evidence the yet important role of the glass composition: the current borosilicate glasses such as the R7T7 one corrode really congruently in Boom clay, whereas the high Al2O3 Pamela glass corrodes by selective dissolution. We have confirmed the fundamental difference in glass dissolution by profile analysis using SIMS (secondary ion mass spectroscopy). On the other hand, the glass dissolution is extremely small at ambient rock temperature of 16°C, and remains below 0.1 µm/y. This result together with the results from the laboratory tests provide large confidence in the quality of the HLW glass as an engineered barrier.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop We obtained a large data base on the leaching behaviour of Tc, Pu, Am and Np in claywater or clay slurries, over extended reaction progress. We measured both the radionuclide concentration in the solution (which is able to migrate through the clay rock) and the radionuclide inventory sorbed on the clay. The average mobile radionuclide inventories released are, in M, 10-8 (Pu), 10-6 (Np), 10-11 (Am), 10-6 (Tc). These data for Np, Tc and Am are lower than the resp. solubilities assumed in the Belgian performance assessment studies, and therefore add to the safety of the disposal concept. When bentonite substitutes part of the Boom clay, these concentrations decrease by as much as 100 times. The upcoming programme 1996- 1999 is focussed on three main areas: The dissolution behaviour of the glasses will be further modelled, by using mathematical and geochemical codes. Specific laboratory tests will be performed to determine specific parameters required by the model, such as the diffusion coefficient of Si through the glass surface layer and through the clay, the porosity of the surface layer. The identification of the complexes including Tc and Np in the interacting media. This study will consider as important parameter the presence of humic acids in the solution. The complexes will be characterized by various techniques, such as laser spectroscopy. The experimental data will be correlated with theoretical calculations of solubility. The demonstration by in-situ testing of the performance of an alpha active glass in disposal environments. This test is called "Coralus", and will consider the presence of a gamma irradiation field, and different interacting materials (Boom clay, bentonite backfill). This project is discussed in a separate poster. The presentation will summarize the actual state of knowledge, and present the main actions in the future programme. This programme is partly sponsored by NIRAS/ONDRAF and the European Commission.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop VERIFICATION STUDIES ON THE PAMELA HIGH-LEVEL WASTE GLASSES P. Van Iseghem, E. Hoskens, G. Smeyers, D. Huys, L. Sannen, L. Vandevelde SCK·CEN, 2400 Mol, Belgium As part of the Quality Assurance/Quality Control programme implemented by the national nuclear waste management authority (NIRAS/ONDRAF), various material properties are being verified. This is done through measurements on three types of samples :(1) laboratory made samples (MT1), (2) demonstration 1/1 scale drums (MT2) and (3) fully active samples (MT3). This paper deals with the Verification programme for the high-level waste glasses produced in the PAMEAL vitrification furnace, operated by DWK/Belgoprocess. In this plant, about 800 m3 high-level waste were vitrified so far. In 1988, a programme was launched by NIRAS/ONDRAF to measure a number of properties on the various material types, including MT1 samples, 4 MT2 drums, and 30 active MT3 samples. Three glass compositions were produced: SM513 and SM527/SM539, resp. for the low and high enriched waste concentrates. The investigations were carried out by the SCK·CEN. The following properties were checked: chemical and radiochemical analysis, to compare with the nominal composition (MT1, MT2, MT3 samples); homogeneity, on a microscale, or between various positions inside a container, or between containers. This is done through microscopical techniques (elemental X-ray mapping, SEM analysis, radiography), chemical or radiochemical analysis (MT1, MT2, MT3 samples); chemical stability, based on standard corrosion tests (the Soxhlet MCC5 flow test, and the static MCC1 test), on MT1, MT2 and MT3 samples; thermal stability, in terms of devitrification and phase separation behaviour (MT1 samples). Some of the main observations are: the chemical stability of the MT1 samples is comparable with other, actual or procursor HLW glasses; transmission electron microscopy analysis on the HEWC glass SM527 reveals some glass-in-glass phase separation. The size of the droplets is between 300 and 500 nm; some secondary phases, consisting of noble metals (Ru, Rh) or transition metals (Fe, Ni, Cr) are present in the active samples. The α and βγ active elements are homogeneously incorporated; for both MT2 and MT3 samples the glass matrix (based on Si analysis) is quite homogeneous; whereas minor components, e.g. rare earths are less homogeneous. Differences may be as large as 10%.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop the chemical as well as radiochemical (atot,241Am, ßtot,90Sr,137 Cs) composition of the active samples seems to quite reproducible: standard deviations are fairly small. The paper will review all data generated on two glass compositions (SM513 and SM527), with attention to the experimental procedures and the reliability of the data. Interpretation will be done with respect to the materials properties selected, and the various types of samples. This work was carried out under contract 758048 and 058068A2 with NIRAS/ONDRAF.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Synroc - an Alternative Waste Form E. R. Vance, K. P. Hart and A. Jostsons, Australian Nuclear Science and Technology Organization kph@ansto.gov.au It has been noted that given the compositional diversity of nuclear waste and the variations in potential repository geologies in different countries, it is important to have a ''menu'' of waste forms so that one is able to choose the waste form that best suits the type of waste and the geology of the repository (Lutze and Ewing, 1988). This paper summarises the work that has been carried out on Synroc to establish its viability as an alternative waste form for the safe disposal of nuclear waste. The feasibility of production of good quality Synroc on a commercial scale has been demonstrated in the Synroc demonstration plant at ANSTO on a non-radioactive basis. This plant has already produced more than 6 tones of Synroc and is used to optimise process steps and to provide information for the conceptual design of a radioactive plant. The plant has a capacity of 10 kg/hr and produces Synroc of a similar quality to that produced in the laboratory. Chemical durability studies of Synroc have been carried out on non-radioactive samples and Synroc doped with actinides and fission products. Leach tests initially concentrated on predicting the durability of Synroc under standard conditions but lately have been extended to include studies based on different repository geology and ground water compositions. Data will be presented for the release of actinides and fission products measured under standard and simulated repository conditions and also from in-situ studies carried out in the Mol underground facility. In general, the data show that long term release rates are low, << 10-5/g/m-2/d, and continue to decrease, albeit slowly, with time. Studies of radiation damage have been carried out by doping Synroc with 238Pu or 244Cm to simulate damage corresponding to about 105 years of storage in about 2 years. These studies have shown that the volume expansion of Synroc saturates at about 4 to 7% at doses of around 5 to 8 × 1018 a-decays/g, under ambient conditions. Storage of the specimens at 200°C reduces the rate of density change by about 30%. Samples that have accumulated damage equivalent to 13,000 years of storage have leach rates for Mo, Sr, Ca, and Ba which are less than a factor of 10 higher than those of undamaged Synroc. Cs leach rates increase by about a factor of 50 over this storage period. The design of Synroc was based on natural minerals which were known to be stable in the earth's crust for 106 to greater than 109 years. Samples of naturally-occurring zirconolite containing U and Th in trace to major proportions, resulting in a-decay doses from less than 1017 to greater than 1020 a/g allow the study of the effect of radiation damage incurred slowly under geological conditions to be studied. Studies of these samples have shown that

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop the crystal chemistry of the natural samples is close to that of the zirconolite phase in Synroc containing 10 and 20 wt% of simulated PW-4b waste and that the structure of these phases is rendered aperiodic at doses similar to those for the actinide-doped samples. In addition, the natural samples have remained as closed systems for actinide elements regardless of the amorphisation of their structure or contact with ground waters. The work carried out thus far has established that Synroc is a viable alternative for disposal of nuclear waste. Additional work is on-going, however, to further develop the Synroc concept and applicability of this waste form to some special waste streams including excess military Pu disposition. This work will also be detailed within the paper.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop CHEMISTRY OF IRRADIATED SOLIDS D. W. Werst and A. D. Trifunac Argonne National Laboratory, 9700 Cass Av, Argonne, IL 60439 werst@anlchm.chm.anl.gov and trifunac@anlchm.chm.anl.gov Studies of chemical mechanisms of radiation effects in solids address the need to provide a scientific basis for predicting the performance of solid radioactive waste forms. In the Radiation and Photochemistry Group at Argonne National Laboratory many state-of-the-art capabilities have been developed to identify reactive intermediates and elucidate radiolysis mechanisms in solids. Radiolytic H atoms in silica and ice have been studied by time-resolved electron paramagnetic resonance (EPR), with information gained about the activation energy for diffusion, diffusion coefficient and likely reaction partners and their properties. Radiolytic yields and annealing behavior of trapped H atoms have been measured by low-temperature EPR in microporous solids (sol-gel silica, porous vycor glass) and crystalline materials (zeolites). Charge transfer has been studied in sol-gel and zeolite systems by using variable-temperature EPR. Charge recombination reactions in frozen hydrocarbons have been probed by time-resolved fluorescence-detected magnetic resonance (FDMR). Measurements of hydrogen gas evolution from radiolyzed cement grout as a function of water content revealed higher than statistical energy absorption by water, leading to O-H bond rupture. A presentation of these experimental results will accompany discussion of proposed applications of modem radiation chemical techniques to understand fundamental chemistry of irradiated solids by using the unique facilities for pulse radiolysis and detection of reactive intermediates at Argonne.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop CORALUS: IN - SITU CORROSION TEST ON ACTIVE HLW GLASS P. Van Iseghem*, E. Vernaz**, N. Jockwer*** * SCK·CEN (B), ** CEA Valrhô (F), *** GRS Braunschweig (G) 1. Objectives The performance of waste forms in repository conditions must be known, to validate models and conclusions from laboratory tests, and to confirm data on the source term. The various interactions between the engineered and natural barriers must be evaluated, to determine and validate the parameters used in the safety assessment. In this proposal a new kind of in-situ test is proposed, which contributes to answer the problems mentioned above, and which is based on the experience gained in different in-situ tests developed in the past ten years: the Belgian in-situ corrosion tests in clay, the Belgian "CERBERUS" in-situ test, using an active 60Co source, the Swedish tests in the STRIPA mine in granite, the USA "MIIT" tests in salt. So far, in-situ tests on waste forms were restricted to small, inactive or doped samples. This proposal includes glass samples doped with large amounts of alpha emitters. This is consistent with the recommendations of an International Workshop on in-situ testing of nuclear waste glasses [1], stating "Future Tests With Waste Glasses Containing Radioactive Tracers are Recommended " The objectives of the proposed in-situ test are: to determine the dissolution of the glass in simulated disposal conditions. Both the global dissolution and the specific release of radionuclides will be measured. Because we will use coupon glass specimens, we will be able to compare these results with previous lab and in-situ tests, and interpret with the dissolution models based on laboratory tests. Surface and bulk studies of the reacted glasses will help elucidate corrosion mechanisms. to evaluate the migration of the radionuclides through the interacting media in a radiation field. We will investigate two reference situations, the first including both a gamma radiation field and the alpha activity in the glass. In the second situation no gamma radiation sources will be used. in both former objectives, also inactive simulants of other long-living nuclides of interest, such as 99Tc, 135Cs, 79Se, 93Zr, 107Pd, can be investigated in terms of release and migration; to measure various parameters in the contacting media ( dose rate, pH, Eh, gas generation, gas release, the petrophysical properties of the backfill) and their effect on waste form behaviour; to gain experience with techniques for in-situ testing and monitoring in more realistic repository-relevant conditions; to encourage further international cooperation in the Waste Management field. We will consider three interacting media: Boom clay (the Belgian candidate host rock), and two bentonite based mixtures, studied as backfill material within the EU.

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop 2. Work Content High-level waste glass doped with the real actinide concentration, and with the insertion of a gamma radiation source will be exposed to different surroundings in the underground laboratory in clay under the SCK·CEN site. As surroundings we select Boom clay, the Belgian candidate host rock, and two bentonite based mixtures, which are being proposed by the partners as backfill material in the repository concept. Modules corresponding to one kind of interacting environment (Boom clay, bentonite backfills) such as shown in Figure 1 will be assembled into tubes, and three of them will be introduced into openings in the underground laboratory in clay. Each module contains eight glass coupons, consisting of samples doped with one alpha emitter (3 samples total), an inactive sample, and duplicate samples. They will be exposed to media consisting of either the Boom clay or candidate backfill materials. The chemistry of the interaction environment will be carefully monitored during the test, together with the dose rate. The generation and release of gases formed (e.g., H2, CH4, H2S, HCl) will be monitored, and coupled with laboratory investigations of the gas release and of the petrophysical properties of the backfill materials. At well specified times the tubes will be retrieved by overcoring, and the interacting materials and the reacted glass will be analyzed by various radiochemical, chemical and surface analytical techniques. A view of a corrosion tube is shown in Figure 1. Glass samples of the Cogéma R7T7 composition, doped with about 0.85 wt% of NpO2, PuO2 or Am2O3 (the industrial R7T7 glass has a total actinide content of 0.85 wt%), as well as the inactive composition will be made available by CEA Valrhô (F). Because of the different specific activities of these actinides, we will be able to obtain different levels of alpha radiolysis in the interacting material. To demonstrate the feasibility of the assembly, operation, retrieval, dismantling and clay sampling of the active tubes, we will first assemble an inactive tube, loaded with inactive glass samples, without gamma sources. We foresee an operation period of about one year, after which the active tubes will be introduced in the underground laboratory. This test will identify and potentially remediate problems with the active set-ups. The test will also enable GRS (Germany) to demonstrate the gas sampling and analysis system. The work content in the present proposal covers a three year period. This will include the design, preparation and installation of the different tubes. Following our present planning the active tubes will be installed in the underground laboratory in the second half of 1999. Their scheduled operation time will be one and five years. Therefore the retrieval and analyses of the active tubes will must be carried out within a following contract. The paper will present the major conclusions form previous in-situ tests in the Mol underground laboratory, and the outline of the new in-situ test. Reference [1] T.McMenamin (editor), Conclusions of the Int. Workshop "In-situ testing of radioactive waste forms and engineered barriers", EUR 15629 (pre-print), 1994

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Glass as a Waste Form and Vitrification Technology: Summary of an International Workshop Figure 1: CORALUS Total view of a corrosion tube.