Appendix E

Committee Meeting at Argonne National Laboratory-East October 26-27, 1998

OCTOBER 26, 1998 – OPEN SESSION
Agenda

8:30 a.m.

Demonstration Project Status - R. W. Benedict

9:00

Electrorefiner Development Testing - E. C. Gay

9:30

Plans for Waste Form Testing and Qualification - R. W. Benedict

10:15

Break

10:30

Metal Waste Form Testing and Plans - D. Abraham

11:00

Metal Waste Form Modeling - M. C. Petri

11:30

Ceramic Waste Form Development - J. P. Ackerman

12:00 p.m.

Lunch

1:00

Tour of Development Laboratories (Building 205) C. C. McPheeters

2:30

Ceramic Waste Form Testing - W. L. Ebert

3:00

Demonstration Scale Ceramic Waste Processes - K. M. Goff

3:40

Ceramic Waste Form Fission Product Release Modeling -T. Fanning

4:00

Electrometallurgical Program Status - B. Cook

4:15

Electrometallurgical Treatment for Other DOE Fuels -C. C. McPheeters

4:45

Open Discussion

5:00 p.m.

Adjournment

Location: Argonne National Laboratory (ANL)-East Facility



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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 Appendix E Committee Meeting at Argonne National Laboratory-East October 26-27, 1998 OCTOBER 26, 1998 – OPEN SESSION Agenda 8:30 a.m. Demonstration Project Status - R. W. Benedict 9:00 Electrorefiner Development Testing - E. C. Gay 9:30 Plans for Waste Form Testing and Qualification - R. W. Benedict 10:15 Break 10:30 Metal Waste Form Testing and Plans - D. Abraham 11:00 Metal Waste Form Modeling - M. C. Petri 11:30 Ceramic Waste Form Development - J. P. Ackerman 12:00 p.m. Lunch 1:00 Tour of Development Laboratories (Building 205) C. C. McPheeters 2:30 Ceramic Waste Form Testing - W. L. Ebert 3:00 Demonstration Scale Ceramic Waste Processes - K. M. Goff 3:40 Ceramic Waste Form Fission Product Release Modeling -T. Fanning 4:00 Electrometallurgical Program Status - B. Cook 4:15 Electrometallurgical Treatment for Other DOE Fuels -C. C. McPheeters 4:45 Open Discussion 5:00 p.m. Adjournment Location: Argonne National Laboratory (ANL)-East Facility

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 Attendance: Committee members G. Choppin (chair), M. Apted, P. Baisden, E. Flanigen, C. Hussey, B. Kear, F. Mansfeld, E. McNeese, R. Osteryoung, and R. White; staff members D. Raber, C. Murphy Unable to attend: J. Williams Participants: R. Benedict, E. Gay, D. Abraham, M. Petri, J. Ackerman, C. McPheeters, L. Simpson, K. Goff, T. Fanning, B. Cook, and D. Funk Summary of Presentations Gregory R. Choppin convened the meeting at 8:30 a.m. in Conference Room A138 Building 208, ANL-E. Robert W. Benedict, ANL-W, described the Spent Fuel Treatment Demonstration project status. On August 1 ANL-W began treating blanket assemblies. Now 72 of 100 driver assemblies have been treated. ANL-W has have cast two of three waste metal ingots and is on track for hot cell HIP. Two incidents have caused a delay in the schedule. The first was contamination that occurred during sealed tube maintenance. The second, a platen malfunction, halted equipment transfers. These incidents have caused a delay of about six weeks. ANL-W will try to make up this lost time before the end of the demonstration, but there may be a delay beyond the scheduled June 1999 completion. Plans are to have a final report on the demonstration project out at the end of September 1999. ANL-W is planning an environmental impact statement (EIS), and the general counsel is reviewing the Notice of Intent for an EIS at the present time. ANL-W is planning to begin fuel treatment operations after the EIS, and this will last for 12 years. Significant accomplishments for driver treatment include eight assemblies treated in 1 month, 693 kg of low enriched uranium cast, cathode processor batch size increased from 12 to 17 kilograms, casting furnace batch size increased from 36 to 54 kg, and 468 kg of low enriched uranium shipped. Blanket treatment started in August 1998. The Mark-V electrorefiner has run its first batch of irradiated baskets. Results from the Mark-IV anode-cathode module tests are used to design Mark-V improvements. Finally, the blanket element chopper is operational. Eddie C. Gay, ANL-E, talked about Electrorefiner Development Testing. In the past three months ANL-E has been conducting ACM tests in the CMT glovebox. They have 20 baskets, with a capacity of 150 kg of uranium. The cathode assembly for the 25-in.-diameter HTER can run over a variety of batch sizes. The support bars have beryllia insulation so as not to deposit uranium. They have transported 150 kg of uranium. In the original design the scrapers trailed the anode baskets, resulting in uranium hold-up between the faces of the anode baskets and the cathode tubes. This causes the anode driver to stop rotating. A dense deposit forms with increasing rpm. The initial resolution of this problem was to perform periodic stripping, which consisted of reversal of the polarity, and transport of the uranium to the anode baskets. The cathode strippers were modified so that the uranium is knocked off the cathode tubes in the space between the anode baskets. In the outer cathode channel, the number of scrapers per revolution was increased from one to three.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 Operating conditions for initial improved scraper tests in the 25-in.-diameter HTER included a 500 °C operating temperature, 40-rpm rotation speed for the anode drive, a cutoff voltage of 0.45 V, no stripping, cathode scrapers leading baskets, and 0.05 A/cm2 current density. Alternative methods to control dense uranium deposits on the cathode tubes include stripping and dissolution/deposition. Design improvements in the ACM include the following: beryllia covers were placed over concentric tube support bars, cathode scrapers that trailed anode baskets were replaced by leading scrapers, scrapers were provided to raise the fuel in the anode baskets above the taper in the outer cathode tube, and clearance between the anode baskets was increased. A procedure was developed for sustained operation of the ACM that used one anode basket in each cathode channel to enhance dissolution of the dense uranium deposit on cathode tubes. Operating steps consist of 200 A-h electrodeposition (uranium transport from the anode baskets to the cathode tubes), washing for 2 to 6 min with reverse rotation of the anode drive, 40-70 A-h stripping (uranium electrotransport from the concentric tubes to the fuel baskets), washing, and repeating the above steps. Operating conditions for sustained operation of the ACM consisted of an anode drive rotation speed of 40 rpm, voltage cutoff of 0.45 V, a 7 wt % concentration of uranium in the salt, a 100- to 200-A electrodeposition current, and a stripping current of 600 A. Planned near-term tests include the continued investigation of the uranium dissolution method to remove the dense uranium deposit from the cathode tubes, parametric tests with the ACM and 25-in.-diameter HTERs, and parametric tests that will include effects of current density, anode basket spacing, separation between the anode baskets and the cathode tubes, separation between the scraper and the cathode tubes, and rotation speed of the anode drive on the uranium throughput rate. Robert W. Benedict, ANL-W, spoke about plans for waste form testing and qualification. ANL has standardized on sodalite as the ceramic waste form. Sodalite can be used over a greater range of operating conditions than zeolite, and can be used at a higher temperature. An extensive characterization program is now under way. Cladding hulls are cast into a stainless steel zirconium waste form. Activities related to waste forms involve several organizations, with interactive feedback between the repository operator [DOE's Office of Radioactive Waste (RW)] and the waste form producer [DOE's Offices of Nuclear Energy and Environmental Management (NE & EM)]. High-level waste is being packaged for retrievable interim storage. A series of three HIP cans are placed into an ANL-W high-level can, which in turn is placed into a storage hole at the Radioactive Scrap and Waste Facility. Interim storage/retrieval at ANL-W can then be coordinated with the repository for high-level waste shipment. The ANL-W high-level waste can would be placed into a standardized canister, which in turn would be placed in a cask for transportation. This packing and shipping would occur at the Idaho Nuclear Technology and Engineering Center. The standardized waste canisters would then be transported to the repository, where they would become part of the waste package at the geologic repository surface packaging facility prior to final disposal. A time line was presented for waste production, waste qualification, waste document releases, and repository activities, beginning in 1997 and extending to a period past 2014 when the repository would be ready for DOE spent nuclear fuel. Qualification

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 activities focus on preproduction documents and repository data needs. The Waste Acceptance Product Specifications (WAPS) document describes the generic waste form and waste packages, including isotopic inventory, criteria for waste physical characteristics, phase stability, and package design. The Waste Compliance Plan (WCP) describes the experiments, tests, and measurements that will be performed to demonstrate that the waste processes will produce a product that complies with the requirements in the WAPS. It will include product consistency measures and important physical property measurements. The Waste Qualification Report presents the data, analyses, and results obtained by executing the WCP as well as data to support waste form product control. Waste forms testing and modeling provide data input to the repository performance assessment (modeling). It includes viability assessment and license inspection. Data has been supplied for the Yucca Mountain EIS. The data include the waste form description, including volume, mass, density, chemical composition, and metric tons of material. Packaging description includes the number of cans and the kilograms of material per canister. Additional data are included on the total number of radionuclides, the average nuclides per canister, and the highest concentration of radionuclides in a canister. Leach rate is assumed to be at least as resistant as for Environmental Assessment glass. Also included are the dose rate and a limited criticality assessment. These data were submitted through DOE-EM, with a first submittal in July 1997 and a final submittal in May 1998. Qualification activities to be completed by the end of the demonstration project include completion of the waste form testing matrices, except for radioactive samples, and tests that last longer than 1 year. Methods will be established for extrapolating both waste form models to the repository time frame (e.g., ~1 million years). In addition, sensitivity studies will be performed using bounding waste degradation models. These should help identify the major parameters that drive repository performance for the ceramic and metal wastes. For both waste forms, degradation modeling refinements based on experimental data will be incorporated into the simplified TSPA model. The Preliminary Waste Acceptance Product Specification and Waste Compliance Plan will be discussed with DOE-RW. Finally, future waste test matrices for licensed applications will be established. Daniel Abraham, ANL-E, spoke on metal waste form (MWF) testing and plans. The MWF test plan includes attribute tests that are designed to provide material property information, electron microscopy, X-ray analyses, neutron diffraction, and investigation of mechanical and physical properties. The total number of tests is 215, scheduled for completion in June 1999. All of the tests have been initiated, and 165 have been completed. Characterization tests are designed principally to generate information regarding mechanistic understanding of alteration. They include immersion and pulse flow corrosion tests. Of 194 tests, all have been initiated and 80 have been terminated (i.e., the testing has ended, but analysis is not complete). Accelerated tests are designed to accelerate the waste form alteration rate beyond rates measured under service conditions. They include high-temperature immersion, vapor hydration, and electrochemical tests. Of the 422 tests (360 of which are electrochemical tests), 130 have been completed. All tests are expected to be completed by June 1999. Service condition tests are designed to simulate actual service conditions. They include thermal aging, drip

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 tests, and materials interaction tests. Of 25 tests, 9 have been terminated, and all are expected to be initiated by June 1999. The total number of tests is 856, with 24 different test types, 39 alloy compositions, and 111 ingots. Characterization tests include immersion testing. These take place in sealed Teflon vessels at 90 °C, in simulated J-13 and deionized water. Two hundred and twenty of these tests have been initiated, and 80 have been terminated. Results so far show that the specimens are either unaffected or show minor tarnish. The test solutions have been submitted for elemental analysis. Alloy samples that have been tested for up to 3 years in J-13 show minor surface tarnish, considerable scatter in data, and that normalized losses are small, i.e., that there are low corrosion rates. Other characterization tests include pulse flow immersion tests. Immersion tests were interrupted to obtain leachate, which provides radionuclide release rates. All tests listed in the MWF Test Plan have been initiated, including 36 tests containing Tc and/or U. An SS-15Zr-2Tc sample was examined recently. It had been submerged in deionized water for 400 days at 90 °C. Mild surface tarnish was found, along with some rusting of pre-existing pits. The pulse flow immersion test data showed that Tc release was small (NL ~0.025 g/m2) and that U release was higher, but still small (NL ~0.15 g/m2). Accelerated tests include high-temperature immersion. These are designed to accelerate the corrosion rate. They take place at 200 °C, in deionized water, in a titanium vessel for 28 days. Tests were conducted on 6 alloy compositions, including 1 slag composition. There were a total of 14 specimens. The corrosion rates were found to be small. There was considerable scatter in the data. There was no correlation of elemental leaching with alloy composition. Normalized loss of the fission products was less than 0.1 g/m2. Elemental loss from the slag specimens was similar to those from the alloy. In vapor hydration tests, corrosion was accelerated by exposure to water vapor at 200 °C. Alteration of the layers was typically 1 micrometer or less in 56-day tests, even those with uranium or plutonium. Under similar conditions, borosilicate glasses alter much more rapidly —from 1 micrometer to complete alteration. Electrochemical corrosion testing measured corrosion rate by the linear polarization method (based on ASTM G 59). The corrosion rates of the MWF alloys are small and are similar to those for SS 316 and Alloy C 22. They are 2 to 3 orders of magnitude smaller than the corrosion rate for mild steel. Proposed tests beyond June 1999 include electrochemical tests at higher temperatures (40, 70 and 90 °C), electrochemical tests in chloride solutions, immersion tests in chloride solutions, studies of crevice corrosion, and study of corrosion mechanisms. Mark C. Petri, ANL-E, presented information on metal waste form release-rate modeling. Developing a radioisotope release rate model for the stainless steel metal waste form involves assessing experimental needs to support the modeling effort, and is a collaborative effort among RE, RA, CMT, and NT. Scheduled MWF modeling milestones include a status report on release mechanisms, which was released in September 1998, and determination of the methodology for accounting for radiation damage, which was completed in October 1998. In addition, identification of the methodology needed to generate input needed by the TSPA analysis was also completed in October 1998. Ongoing milestones include the establishment of and familiarity with

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 current Yucca Mountain MWF modeling facilities, and the development of any additional methodology needed to extrapolate metal test data to the repository time frame. By the end of the demonstration project, the results of the MWF modeling activities will be documented. Current uniform test results indicate that the MWF behaves much like stainless steel. Therefore, known stainless steel degradation mechanisms are used as a basis for MWF release modeling. Also, MWF corrosion data are used to empirically adjust models. Interrelated factors are important to radioisotope release from the MWF. In terms of MWF metallurgy, alloy composition, microstructure, and crystallography are all important. The degradation mechanism and environmental conditions, including temperature, humidity, water chemistry, and radiation effects also play a role. Potential degradation mechanisms include uniform aqueous and humid-air corrosion, crevice or pitting corrosion, microbially influenced corrosion, selective leaching, intergrannular attack, and galvanic corrosion. Study of repository water chemistry indicates that the water chemistry can vary from J-13 well water, and that some models predict salt concentrations greater than those in J-13 water. Waste canister modeling is concerned with breaching of the carbon steel and C-22 shell. Waste canister corrosion testing has used concentrated J-13 well water (up to 5000 x). Long-term canister corrosion models have used empirical multi-variate regressions. Key variables for steel corrosion include temperature, pH, and chloride content. Other variables include hydrogen peroxide content, and bicarbonate and nitrate contents. Current MWF corrosion testing includes immersion testing, electrochemical testing, and vapor hydration testing. Additional testing will include chloride contents up to 10,000 ppm, room temperature up to 90 °C, pH between 5 and 10, various MWF alloy compositions, and crevice-corrosion tests. John Ackerman, ANL-E, spoke about CWF development. Thermal stability in storage of the CWF was studied in a 3-month test, with the results forthcoming. One-year tests are under way, and exposure is expected to be complete by June 1999. “Cold” reference waste forms samples have been fabricated for waste form testing. In addition, uranium-bearing reference forms fabrication has taken place. Plutonium-bearing hot uniaxial pressed specimens have been provided. Of the parametric studies, the glass content studies indicate that a 25 wt % glass is recommended for production. A nominal zeolite dryness requirement of 0.3 wt % water has been set, and tests are to be completed in April 1999. The nominal temperature has been set at 775 K for blending, and these tests are also to be completed in April 1999. Particle-size studies have looked at several size combinations, and the samples were made and sectioned for analysis. Analysis of the data awaits leaching results, but there are no outstanding effects based on the data to date. In uranium studies, the concern was that UCl3 reacts to destroy the zeolite lattice. The concern was warranted, as reaction does occur, but UCl3 reacts with water, rather than zeolite, and after zeolite drying, adequate (by more than a factor of 2) water remains to consume UCl 3.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 For plutonium studies, there was concern that PuCl3might react like UCl3. This concern does not appear to be warranted. PuCl3, like the rare-earth chlorides, appears not to attack zeolite no matter how much excess PuCl3 is used. PuCl3 reacts with water to form PuO2. Waste forms and precursors show no degradation. Zeolite was exposed to varying amounts of PuCl3, and heated to 773 K as in fabrication. Examination using X-ray diffraction showed no changes in the zeolite pattern; no zeolite degradation products were seen, and variations in intensity and background effects all are attributable to the salt. Pu-bearing waste forms were fabricated (HUP). No evidence of degradation was seen. Waste forms are being investigated with the full range of Pu levels. Advance fabrication of the ceramic waste form includes high-temperature, pressureless consolidation. Pressureless consolidation serves as an attractive alternative to hot isostatic pressing (HIP). The powder preparation is identical to HIP. The balance of the process is greatly simplified and should result in a great increase in throughput. It requires a ~50 wt % glass versus 25 wt % for HIP. Initial scoping studies achieved attractive samples with >45% glass and 44- to 125-micrometer powders. Scale-up to 3-inch- and 6-inch diameter has been successful. The objective is to understand and demonstrate the scale-up. K. Michael Goff, ANL-E, spoke about demonstration-scale ceramic waste processes. The hot isostatic press (HIP) out-of-cell testing is essentially complete. Mock-ups for in-cell operations are complete. The hot fuel examination facility installation is scheduled for completion in December 1998. For the heated V-mixer, heater reliability issues have been resolved. V-mixer activities are focused just on the granulated zeolite. The best mixing results have been obtained with granulated material. Problems related to salt and zeolite particle size were encountered and included segregation of the salt, which in turns sticks to the walls of the V mixer. The problem is due to different salt and zeolite particle sizes. A design-of-experiments process has been implemented to develop a statistical process model. This process identifies important parameters and quantifies effects. The design-of-experiments process also develops an empirical model of response variables, and minimizes the number of tests required to define and optimize the production process. Development of witness tubes for characterization involves exploring the use of small sample tubes processed with HIP cans for characterization. The tubes would be subjected to characterization by a product consistency test (PCT) as an alternative to destructive analysis of the large-scale HIP can. PCT results from the witness tubes versus the HIP can have indicated that the witness tube has less porosity than the HIP can. Supporting testing for accelerated alpha decay studies includes work first performed with surrogate salt-loaded zeolite to assess the use of a hot uniaxial press (HUP) for sample production. Operations were then performed with depleted uranium for equipment check-out and with Pu-239 for equipment check-out and to assess product quality. The first batch of material containing Pu-238 was produced and X-ray diffraction results have been completed. Accelerated alpha decay studies on samples have produced the first of two batches of salt and salt-loaded zeolite. The salt contained 15 mole % plutonium and surrogates for fission products. The HUP sample contained more than 2 wt % plutonium.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 William L. Ebert, ANL-E, spoke about ceramic waste form testing. Laboratory tests were conducted to support performance calculations for the disposal system (TSPA-LA), to support development of models for long-term behavior (ANL model), and to monitor the consistency of the CWF product. Assessment of total system performance includes identification of key release modes under likely disposal conditions, determination of the mechanism for the corrosion of the waste form, determination of pathways for release of radionuclides and possible retention modes in near-field, and measurement or calculation of kinetics of waste form corrosion and radionuclide release. Support of model calculations includes identification of important corrosion/release processes to be modeled, providing parameter values for the rate expression to calculate long-term releases of nucleotides, and generation of a separate database to confirm model calculations. Monitoring product consistency includes identification of the properties to be monitored, development of a test method that is sensitive to variations in a monitored property, and generation of a database to confirm the adequacy of tests for product consistency over a range of processing conditions. The product composition is used to control the process, so consistency of the glass product composition is monitored. The product consistency test (PCT) is used to monitor the consistency of defense high-level waste (DHLW) glass. The partial dissolution test monitors only highly soluble components. The response of waste glass compared to the response of the environmental assessment benchmark is also monitored. A waste form that has acceptable PCT-A behavior is not viewed as ensuring that its performance will be acceptable for disposal. The DOE is at risk with regard to determining the acceptability of glass from Hanford and Idaho if only PCT-A data are available. Work is in progress to relate PCT-A to long-term performance for DHLW glass. Major components of the ceramic waste form are sodalite, glass 57, and a clay binder. Minor components consist of salt and other phases, including those from transformation of the zeolite during HIP, other reactions, and possibly small amounts of radionuclides. MCC-1 tests were performed with a reference waste form and with HIPed sodalite. Tests with glass 57 showed typical glass behavior and dissolution rate. Tests with sodalite showed rapid initial release of sodium, and this rate must be confirmed with dilute PCT. Tests with the reference waste form showed good agreement with sodalite. Long-term PCT was also run with 4-micrometer sodalite and with the reference waste form. The tests, at 2000 and 20,000 m-1, verify pH independence. Tom Fanning, ANL-E, presented information on ceramic waste form radionuclide release modeling. The purpose of the modeling effort is to develop a radionuclide release model for the ceramic waste form. This will allow prediction of long-term behavior based on short-term tests of important processes. It also provides conservative release estimates based on waste form behavior. Determination of the experimental data needed to support performance assessment activities is required. Incorporation of the radionuclide release model into the repository assessment model is planned. The initial approach to modeling is to identify important exposure mechanisms, to focus on the sodalite as the dominant radionuclide-bearing phase, to take advantage of similarities between sodalite and glass, and to treat corrosion of each phase independently. In the conceptual model, release to the environment depends on waste form exposure, radionuclide release, and groundwater transport. The release mechanisms in the model

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 hold that radionuclides in sodalite and oxide/oxychloride phases are released via ion exchange and/or dissolution. The assumption is that non-occluded ions (those in glass, clay, salt, secondary mineral phases) are released immediately. The model requirements include confirmation of the exposure mechanism, characterization of important radionuclide-bearing phases, determination of temperature and solution chemistry (including pH) effects on glass and sodalite corrosion, and investigation of ion exchange rates between solution and sodalite. Beverly Cook, DOE-NE, spoke about the electrometallurgical program status. DOE is in a state of reorganization, and it is now moving toward the Notice of Intent for an environmental impact study and that the committee 's assessment of the electrometallurgical process is important to this. Waste form issues are an important component of the committee 's work. A clear test plan for waste forms is needed with parameters. DOE now has large matrices and wants to see which ones drive the process. DOE's management plan is to stay in contact with the Department on Repository Waste as it develops criteria for waste forms. DOE would like to have the committee's assistance with this. She stated that DOE is committed to completing the demonstration project and treating all EBR-II spent fuel. Discussions between the committee and Beverly Cook took place during this presentation. Charles C. McPheeters, ANL-E, spoke on electrometallurgical treatment of other DOE fuels. An overview of the lithium reduction concept showed lithium reduction of oxide spent fuel in molten LiCl. The salt-soluble fission products and the particulate fission products are then separated and made into a ceramic waste form, while uranium, transuranic elements, and noble-metal fission products are separated and sent to the electrorefiner. The focus of the oxide work is twofold. Laboratory-scale work is done to investigate the limits for reduction of PuO2, the kinetics of UO2 reduction, electrowinning the cathode for Li handling, and electrowinning anode materials. Engineering-scale work is done to study the effect of scale on reduction kinetics, the fuel basket for reduction and electrorefining, and the lithium handling method development. A porous cathode for lithium electrowinning from LiO2 has been developed that has a steel rod attached to lithium metal, which is wrapped by a stainless steel screen. For the anode, candidate materials are being investigated. Platinum has the advantage of performing well, having a low overpotential, and a backlog of experience. Its weaknesses include expense and the fact that it reacts with lithium and chlorine gas. Fe3O2 has low resistance, is a corrosion product, and is cheap, but fabrication presents difficulties and there is the problem of thermal shock. For SnO2, advantages include its commercial availability, its low resistance, and its low cost. Its weakness is that it has a high overpotential for the generation of oxygen gas. On the engineering scale, a lithium reduction facility, reduction vessel, electrowinning (ES-8), and cathode have all been developed. Electrorefining objectives include characterization of the reduced product for slat content, residual LiO2, and residual lithium. Other objectives include investigations into the effect of fuel morphology and the effect of basket design. Treatment goals for aluminum fuel include the minimization of high-level waste. Options should include disposal in defense waste glass, and the use of electrometallurgical treatment to minimize costs. A fission

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY AS OF FALL 1998 product salt/metal distribution study is needed for understanding the initial fluxing step. It would provide data needed to determine activity coefficients (or ratios) of fission products in aluminum and in the salt. Thermodynamic calculations predict that the salt phase will contain active metal and rare-earth fission products, and the metal phase will contain uranium, plutonium, and noble-metal fission products. OCTOBER 27, 1998 — EXECUTIVE SESSION 8:00 a.m. Opening Remarks (G. Choppin) 8:15 Review of the Previous Day's Presentations at ANL-W 10:15 Meeting and Report Planning 10:45 Report Writing Assignments 11:00 Report Writing Session 12:00 p.m. Lunch 1:00 Report Writing Session 2:00 p.m. Adjourn The committee met from 8:00 a.m. to 2:00 p.m. at the conference facilities at the O'Hare Hilton in Chicago, Illinois. The chairman and committee followed the agenda listed above.