Appendix D

Committee Meeting at Argonne National Laboratory-West June 25-26, 1998

JUNE 25, 1998 – OPEN SESSION
Agenda

8:15 a.m.

Fuel Treatment Economic Evaluation

Y. I. Chang

8:45

Project Status and Success Criteria Accomplishments

R. W. Benedict

9:45

Break

10:00

Waste Qualification Strategy

T. P. O'Holleran

10:20

Repository Modeling

L. L. Briggs

10:40

Ceramic Waste Demonstration Scale Tests

K. M. Goff

11:00

Waste Form Development & Qualification Testing

J. P. Ackerman

12:15 p.m.

Lunch

Engineering Conference Room

1:00

Tours

--- Fuel Conditionig Facility

--- Ceramic Waste Process

R. A. Evans

K. M. Golf

2:00

Irradiated Metal Waste

D. D. Keiser

2:30

Metal Waste Qualification Testing

S. G. Johnson

3:00

Electrorefiner Development Testing

E. C. Gay

3:30

Irradiated Fuel Electrorefining Experience

R. D. Mariani

4:00

Electrorefining Model Development

R. K. Ahluwalia

4:30 p.m.

Open Discussions

Location: Argonne National Laboratory (ANL)-West Facility, Idaho National Engineering and Environmental Laboratory

Attendance: Committee members G. Choppin (chair), M. Apted, P. Baisden, E. Flanigen, C. Hussey, B. Kear, F. Mansfeld, L. E. McNeese, R. Osteryoung, R. White, and J. Williams; guests Y. Chang (ANL), R. Benedict (ANL), T. O'Holleran (ANL), L. Briggs (ANL), K. Goff (ANL), J.



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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 D Committee Meeting at Argonne National Laboratory-West June 25-26, 1998 JUNE 25, 1998 – OPEN SESSION Agenda 8:15 a.m. Fuel Treatment Economic Evaluation Y. I. Chang 8:45 Project Status and Success Criteria Accomplishments R. W. Benedict 9:45 Break 10:00 Waste Qualification Strategy T. P. O'Holleran 10:20 Repository Modeling L. L. Briggs 10:40 Ceramic Waste Demonstration Scale Tests K. M. Goff 11:00 Waste Form Development & Qualification Testing J. P. Ackerman 12:15 p.m. Lunch Engineering Conference Room 1:00 Tours --- Fuel Conditionig Facility --- Ceramic Waste Process R. A. Evans K. M. Golf 2:00 Irradiated Metal Waste D. D. Keiser 2:30 Metal Waste Qualification Testing S. G. Johnson 3:00 Electrorefiner Development Testing E. C. Gay 3:30 Irradiated Fuel Electrorefining Experience R. D. Mariani 4:00 Electrorefining Model Development R. K. Ahluwalia 4:30 p.m. Open Discussions Location: Argonne National Laboratory (ANL)-West Facility, Idaho National Engineering and Environmental Laboratory Attendance: Committee members G. Choppin (chair), M. Apted, P. Baisden, E. Flanigen, C. Hussey, B. Kear, F. Mansfeld, L. E. McNeese, R. Osteryoung, R. White, and J. Williams; guests Y. Chang (ANL), R. Benedict (ANL), T. O'Holleran (ANL), L. Briggs (ANL), K. Goff (ANL), J.

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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 Ackerman (ANL), R. Evans (ANL), D. Keiser (ANL), S. Johnson (ANL), E. Gay (ANL), R. Mariani (ANL), and R. Ahluwalia (ANL) Summary of Presentations Gregory Choppin, committee chair, opened the open session with an introduction of the committee members. Yoon I. Chang, ANL-E, presented a talk evaluating the economics of DOE spent fuel treatment. The justification for electrometallurgical treatment was discussed. The current DOE/EM approach is to stabilize vulnerable spent nuclear fuel (SNF) and store it in a safe condition for an interim period (assumed to be 40 years). Direct disposability of RCRA-characteristic, pyrophoric, or highly enriched DOE spent fuel types remains in question. Electrometallurgical treatment resolves these issues and has the potential for reducing the DOE spent fuel disposal life-cycle costs significantly. Electrometallurgical treatment results in additional benefits. It eliminates technical uncertainties associated with the problematic spent fuel types concerning whether they require treatment or can be qualified for direct disposal. It produces robust and common waste forms for interim storage as well as longer-term storage in a repository. The uranium by-products are separated, eliminating criticality concerns in the repository. DOE SNF disposal costs without treatment ranged from a low, using existing facilities, of $16 billion, to an upper range, with new facilities, of about $25 billion. This was compared with the range of costs saved through the use of treatment. The lower avoided costs were $10.7 billion, and the upper range was $14.2 billion. The electrometallurgical treatment strategy is to utilize existing hot cell facilities at ANL-W, Idaho National Engineering and Environmental Laboratory, and Hanford. The decision for implementation will be made in 2001, and the treatment will be completed during the period 2006 to 2015. Electrometallurgical treatment costs were estimated for waste storage at $200 million for capital costs plus $400 million for operating costs. Waste disposal costs were based on reduction of canistered waste volume by a factor of five, leading to an estimated cost of $740 million. The total treatment and disposal cost were estimated to be $2.9 billion. Cost savings with electrometallurgical treatment were set between a lower and upper range of $7.8 billion to $9.8 billion. Robert W. Benedict, ANL-W, spoke on the spent fuel demonstration status and success criteria accomplishments. The EBR-II spent fuel treatment flow sheet was reviewed, demonstrating the separation of uranium, and the ceramic and metal waste forms individually. Following a review of upcoming demonstration project milestones, a response to the committee's spring 1997 report1 was given. The committee had stated in that report that it looked forward to receiving the demonstration project implementation plan after DOE had approved it. The revised Work Breakdown Structure describes implementation in section 11. DOE has verbally approved. The committee had stated that before completion of the demonstration, DOE should establish criteria for success in the demonstration phase to allow evaluation of the electrometallurgical technology for future use in treating DOE spent fuel. DOE has now issued the criteria and specific goals. A status report with accomplishments was issued by DOE in June 1998. In addition, an 1   Electrometallurgical Techniques for DOE Spent Fuel Treatment: Status Report on Argonne National Laboratory's R&D Activity Through Spring 1997, National Research Council, National Academy Press, Washington, D.C., 1997.

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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 environmental impact statement is scheduled to begin in July 1998. The committee also stated that DOE's offices of Nuclear Energy (NE) and Environmental Management (EM) should maintain close contact to ensure proper coordination of their activities. EIS activities are being coordinated with DOE EM, which monitors the demonstration project. In response to the committee's recommendation that DOE should establish acceptance criteria for waste forms scheduled for storage in a geologic repository, it was stated that data has been submitted for a Yucca Mountain environmental impact statement. In addition, waste form qualification documents are being drafted. They will be reviewed internally and then sent to Yucca Mountain to make sure that these documents are compatible with Yucca Mountain's requirements. The committee also recommended that ANL utilize external technical experts in specific areas of the program. It was stated that ANL is utilizing University of Chicago peer review groups, which issue reports to DOE, and experts in specific topics, such as nuclear material control and accountancy, to review special areas. A review of the current status of the demonstration project, in light of the success criteria, followed. A demonstration summary noted that 64 driver assemblies have been treated and process specifications have been drafted. Unirradiated blanket fuel is being processed. Metal waste laboratory-scale samples and one demonstration-scale sample have been cast, and qualification testing has started. The majority of laboratory-scale and some demonstration-scale ceramic waste samples are prepared. Qualification test methods have been modified for ceramic waste samples. Finally, interactions have been initiated with DOE Environmental Management and Civilian Radioactive Waste Management, and the Yucca Mountain project. Tom O'Holleran, ANL-W, presented information on the waste qualification strategy. The objective is to ensure that the full-scale waste “package” will qualify for repository disposal. The criteria for waste acceptance are defined in the Waste Acceptance System Requirements document, and in the Mined Geologic Disposal System Waste Acceptance Criteria. Borosilicate glass is the standard waste form for high-level waste. Some waste forms may not meet all current criteria, but may be acceptable with respect to operational safety and repository performance. The need to modify criteria to accept other waste forms is recognized, but the process (the Waste Acceptance Criteria, or WAC, process) is not yet defined. The waste qualification strategy is summarized in a draft Waste Form Qualification Plan. The existing document hierarchy will be used to qualify the metal and ceramic waste forms as additional standard high-level waste forms. The applicability of most existing requirements was emphasized. A “standard” canister design is to be used so that no changes will be required to repository operations. One set of documents will be prepared to qualify both waste forms since co-packaging will result in a single canistered waste form. Also, the “WAC Petition Process,” when it is defined, will be exercised to qualify the waste forms for disposal. This approach maximizes the likelihood of waste acceptance by the Civilian Radioactive Waste Management System. Waste form qualification takes place within a complex regulatory framework, involving not only Radioactive Waste, Nuclear Energy, and Environmental Management within DOE, but also the Nuclear Regulatory Commission, the Environmental Protection Agency, and the state of Nevada. ANL 's qualification strategy focuses on preproduction documents and repository data needs. The Waste Acceptance Product Specifications (WAPS) describes the generic waste form. The Waste Compliance Plan (WCP) describes the experiments, tests, and measurements that will be performed to demonstrate the ability to comply with the requirements in the WAPS. The

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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 Waste Qualification Report (WQR) presents the data, analyses, and results obtained by executing the WCP. Additional waste form data on long-term radionuclide retention provides input to repository performance assessment (modeling). Few changes in glass specifications are needed for qualifying electrometallurgical waste forms. Proposed changes reflect differences between glass and the electrometallurgical waste forms. Chemistry must allow for crystalline waste forms. Phase stability must identify the lowest temperature that causes phase transitions or other changes. Dimensions must be specified to ensure that solid waste forms can be inserted into containers. The storage can takes advantage of existing facilities, waste will not be re-packaged prior to shipment, and it must not be specified as part of the canistered waste contents. Metal and ceramic waste forms will be co-packaged, which may be necessary to help meet weight limitations. This will give one set of requirements for the canistered waste form, which provides the primary interface with repository operations. One set of documents will therefore cover both waste forms. Implementation of the waste form qualification plan was started by preparing preliminary documents. Revision 1 of the preliminary Waste Acceptance Product Specification document is in draft form. Revision 0 of the preliminary Waste Compliance Plan has been prepared, and revision 1 is awaiting revision 1 of the preliminary WAPS. These documents will be revised as new information becomes available and will evolve into the final waste form qualification documentation. Laurel L. Briggs, ANL-W, spoke on repository modeling. Waste-form-specific factors in repository performance include the radionuclide inventory, waste form heat generation, waste form degradation, and the waste form radionuclide release mechanism. For EIS data call input ORIGEN-II analysis was used to generate a conservative radionuclide inventory for the 60 metric tons of DOE sodium-bonded spent nuclear fuel. Pyroprocess flow sheets were then used to partition the inventory between the ceramic and waste metal waste forms. These provide source team information for thermal analyses and repository performance assessment. In the packaging design study, the first approach is to produce an optimal electrometallurgical technology (EMT) waste packaging approach, taking into account thermal, operational, and transportation constraints while maintaining consistency with DOE packaging strategies. Surveyed DOE documents on standardized containers are developed in order to establish an envelope within which EMT packaging must operate. Packaging must be suitable for interim storage in the Radioactive Scrap and Waste Facility (RSWF) at ANL-W and for loading into DOE standardized canisters at the planned surface facility at Idaho National Engineering and Environmental Laboratory. Study results indicate that ceramic and metal wastes can both be packaged in storage containers currently in use at RSWF (HFEF-14 cans). These will fit into 24-inch-diameter DOE standardized canisters. The greatest packaging efficiency is achieved with 20-inch-diameter ceramic waste HIP cans (these appear feasible to produce). Metal waste ingots are currently planned to be 9 to 10 inches in diameter and will be placed three to a layer. Materials selections, such as the stainless steel shield plugs to replace the lead ones currently used by the HFEF-14 cans, are driven by geologic repository considerations. Thermal analysis included a conservative evaluation of both ceramic and metal wastes under interim storage and permanent disposal conditions. The interim storage analysis results indicated that peak temperatures for both wastes stay below 200 °C, which is well below limiting temperatures for either waste form. The geologic repository analysis results indicated that peak temperatures within the waste forms do not exceed 300 °C, well below limiting temperatures for either waste form. Metal waste form

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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 degradation modeling will provide input needed by the repository performance assessment. The major degradation mechanisms are similar to those which have been identified for the waste package corrosion allowance material. Radioactive release can be effected by many categories of factors: metallurgical, physical, chemical, radiation-induced changes, and mechanical. Evaluation and model development will rely on results from the metal waste experiment matrix. Ceramic waste form degradation modeling will provide the input needed by the repository performance assessment. Initial ceramic waste degradation modeling is based on established degradation modeling approaches for various types of spent nuclear fuel and defense high-level waste glass. The major waste form phases and their radionuclide release mechanisms have been identified. Sodalite experiment results will provide corrosion rate data needed for model refinement. Performance assessment sensitivity studies will identify the dominant model requirements where refinement is needed. For the repository performance assessment, ANL plans to do scoping studies and parameter sensitivity studies using performance assessment. The tool for these studies is the Repository Integration Program (RIP) code from Golder Associates used by the Yucca Mountain Project (YMP) for the 1998 viability assessment (TSPA-VA), as well as earlier performance assessment studies. In April 1998, the YMP provided ANL with the TSPA-VA base case model. ANL and the National Spent Nuclear Fuel Program have been working with Golder Associates to implement a simplified version of the TSPA-VA base case to use for scoping and sensitivity studies—the YMP is to review this simplified model. The radionuclide inventory evaluations and waste form degradation and release modeling specific to the EMT waste forms will serve as input to ANL's sensitivity studies and eventually to the YMP assessments for licensing. Future work will include the continued development of degradation modeling and radionuclide release modeling for both waste forms. Also, ANL will develop experience with the TSPA-VA base case and work with the YMP and Golder to perform sensitivity studies focused on EMT waste performance. ANL will use performance assessment results as input to the experimental matrix to aid in decisions about where best to focus experimental efforts. As needed, ANL will modify the design envelope for EMT waste containers and waste packaging operations equipment to keep current with the evolving national program. K. M. Goff, ANL-E, spoke about ceramic waste demonstration scale tests. An overview was given of the ceramic waste demonstration equipment, including the mill/classifier, the heated V-mixer, and the hot isostatic press. Testing with sodalite in the HIP cycle found it to be an excellent waste form. The HIP can design showed a low amount of cracking of the waste form in testing. Other performance testing results, including analysis of salt content in HIP cans from various V-mixer runs and sodalite HIP runs, are favorable. Accelerated alpha decay tests with Pu-238 should be starting within the month. J. P. Ackerman, ANL-E, spoke on waste form development and qualification testing. A variety of tests were performed on the ceramic waste form in support of the demonstration project. Short-term (up to one week) thermal stability tests of the ceramic waste form in storage are complete, and appear to indicate that the allowable temperature limit is > 875 K (600 °C). Longer-term tests are under way. Sample fabrication tests include “cold” reference waste forms fabrication, completed in February 1998, uranium-bearing waste forms fabrication, completed in April 1998, and plutonium-bearing HUP specimens, with completion scheduled for August 1998.

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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 Among the parametric studies, glass content studies looked at a range of 0 to 50 wt % glasses. Three-day Cs loss was used as an indicator of quality. 25 wt % glass is recommended for production. Another parametric study investigated zeolite dryness requirements. The nominal requirement was set at 0.3 wt % water. Early results very tentatively indicate no major effects up to 1 wt %. A study of the blending temperature set the nominal temperature at 775 K (500 °C). A range from 675 K (400 °C) to 725 K (450 °C) will be investigated. These tests are to be completed in April 1999. For particle size studies several particle size combinations will be evaluated. Tests are scheduled for completion in November 1999. In the uranium studies, there was a concern that UCl3 reacts to destroy the zeolite. Thermodynamic analyses suggested that reaction is possible, and early scoping studies indicated possible reaction. There is reaction; however, UCl3 appears to react with water, rather than the zeolite. For salt disposal, after zeolite drying, adequate amounts of water should remain to react with all UCl3 in the electrorefiner salt. In plutonium studies, there was concern that PuCl3 might react like UCl3. This concern appears to be unwarranted. PuCl3 reacts with water to form a pouch. Zeolite powders were exposed to varying amounts of PuCl3 and heated to 775 K (500 °C). When examined using X-ray diffraction, no change in the zeolite pattern was seen, no zeolite degradation products were observed, and intensity variations and background effects were all attributable to the salt. Pu-bearing HUP waste forms showed no evidence of degradation. Waste forms are being investigated with a full range of Pu levels. Zeolite column development parallels the demonstration project. Tests with the laboratory-scale column are ongoing. Kinetics and equilibrium tests are under way. An in-cell scale test column is being fabricated. Advanced fabrication is being pursued, and ANL is looking at “sintering ” as a possible future alternative to HIP. Powder preparation is identical for HIP and “sintering.” Loading and powder identification are easier in a simple setter than in a HIP can. Welding, evacuation, sealing, and the HIP thermal/pressure cycle can be eliminated. “Sintering” requires only heating and removal from the setter-tunnel furnace and would give a semi-continuous operation. For the same cycle time, throughput of the consolidation step could be more than 15 times greater. For the metal waste form, advanced characterization is performed by extensive neutron diffraction studies. Advanced characterization provides detailed structural information on several phases, including lattice parameters. For the stainless steel 15 wt % zirconium metal waste form, the development phase is essentially complete. It has been found to have excellent corrosion resistance, it is mechanically strong, its thermal properties are typical of metals, and basic metallurgical characterization has been completed. A metal waste form handbook has been developed. It is a compilation of development work since mid-1993. It is intended as a resource guide for future work. It contains data from ANL-E and ANL-W. The revision 0 draft is complete and internal review is under way. Revision 1 is planned for June 1999 and is to include test matrix data. The program objectives for ceramic waste form testing are to meet requirements for waste form acceptance, and to expedite placement of the waste form in a repository. For the demonstration, the goal is to provide enough corrosion information to assess acceptance of the waste forms by DOE. Dennis D. Keiser, ANL-W, presented material on irradiated metal waste results. The metal waste form comprises stainless steel (316, 304, D9, and HT9) cladding hulls, zirconium (5 to 20 wt %), noble-metal fission products (Tc, Pd, Ru, etc.), and small amounts of actinides (U, Pu, and Np levels up to 5 wt %). Cladding hulls are run through the cathode processor to volatilize

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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 the adhering LiCl-KCl salt. Metal waste form ingots are cast in a vacuum induction furnace in the fuel conditioning facility (FCF) at 1600 °C using an yttria crucible. The yttria crucible is loaded with a 0.25-in. thick type-316SS disk on the bottom (1337 g), followed by 604 g of zirconium, 3312 g of cladding hulls, and two 0.0625-in. thick type-316SS disks (376 g). The cladding hulls were compacted. The used lid is composed of tantalum disks and zirconia felt. The furnace is run at 1600 °C for 3 h. Samples are injection-cast and core-drilled. MWF casting of CFMWF06 consisted of an SS-11Zr ingot that weighed approximately 5.5 kg. Two 1/8-in. thick SS disks were on the bottom, weighing 1150 g. Two assemblies worth of cladding hulls were in the middle (3957 g). One 1/16-in. thick SS disk was on top (287 g). Finally, there was 117 g of zirconium. There was no stirring step, no compression of hulls. A 20-g sample was injection-cast, and a larger yttria crucible (approximately 3 in. higher) was used. Other FCF MWF castings are planned, at 1600 °C with a holding time of one hour, and at 1600 °C using blanket-cladding hulls. For characterization of irradiated MWF ingots, samples are core-drilled from ingots in the Hot Fuel Examination Facility (HFEF) and the FCF. Samples are injection-cast from ingots in the FCF. Chemical analysis of core-drilled and injection-cast samples is performed in the Analytical Laboratory. The core-drilled and injection-cast samples are ground, polished, and mounted in the Containment Box in the hot fuel examination facility. Optical metallography and SEM analysis are performed in the HFEF. Initial observed microstructures for irradiated alloys are similar to ones for alloys doped with noble metals. Stephen G. Johnson, ANL-E, spoke about metal waste qualification testing. The purpose of the metal waste form test matrix is to provide samples that bracket the target alloy composition with realistic variations in constituents and to subject those variations to test methods or techniques that will provide a performance gauge for the sensitivity of the waste form to these variations. This information will allow for the design of a robust metal waste form process. Samples for the metal waste form test matrix contained varied constituents, including zirconium (5-20 wt %), noble metals (0-2 wt %), uranium (0-5 wt %), tranuranics (0-10 wt %), and enhanced alloy. Among the attributes examined in the matrix are the microstructure, phases, thermal expansion, thermal diffusivity, thermal conductivity, specific heat, density, tensile strength, compressive strength, and impact strength. Characterization is performed by immersion tests and pulsed flow monolithic tests. Accelerated tests include vapor hydration and electrochemical tests. For service condition testing, thermal aging, drip test, and materials interaction are used. Pulsed flow tests of SS/15Zr/2Tc showed an extremely low loss of Tc. Pulsed flow tests of SS/5,15,20Zr/2U showed extremely low loss of U. Pulsed flow testing of SS/15 Zr/5 U showed low loss of U. Pulsed flow testing of Super Alloy-I showed a low loss of Tc, and reasonable reproducibility. In the vapor hydration test, alteration is accelerated by exposure to water vapor at 200 °C. Alteration layers are typically 1 micrometer or less in 56-day tests. Excellent durability is seen, even with uranium and plutonium. Glass alterations from 1 micrometer to complete destruction are observed. For electrochemical testing, corrosion rate measurement by linear polarization method is planned. Testing will begin in July 1998. High-temperature immersion tests are being developed as a “quick” product consistency test for the metal waste form. Preliminary data shows that the test accelerates corrosion without altering the mechanism. Additional accelerated tests include an elevated temperature test using SJ-13 water.

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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 Eddie Gay, ANL-E, spoke on electrorefiner development testing. Observations from testing the original 25-in.-diameter HTER found uranium hold-up between the faces of the anode baskets and the cathode tubes. Scraper failure was caused by buildup of dense uranium deposits on the cathode tubes. The cathode scrapers were modified so that 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. Noble-metal retention tests in the 25-in.-diameter HTER met the objective to electrotransport 98.5% of the uranium from the fuel segments with 80% retention of zirconium in the cladding hulls. Operating conditions for the run of the HTER-1 were a 500 °C operating temperature, 40-rpm rotation speed for the anode drive, a cutoff voltage of 0.45 V, no stripping, improved cathode scrapers, and 0.05 A/cm2 current density. Alternative methods to control the dense uranium deposit on the cathode tube include stripping, dissolution, and replacement of the cathode scraper with a uranium cutter. Operating conditions for the stripper steel rods include an 8-in.-diameter HTER, a 20-rpm rotation speed for the stripper rods, 0.6-milliohm resistance, 50-to 200-Å stripping current, a cutoff voltage of 0.45 V, 500 °C operating temperature, and separating the electrodes at 500 °C after stripping. The uranium product from the stripping rod test had the following characteristics: weight of the product on steel rods of 2.2 kg, and a product composition of 77.0 wt % uranium, about 22 wt % salt, 0.2 wt % cadmium, and |lessthan|0.1 wt % zirconium. Advantages of the dissolution method to remove the dense uranium deposit on the cathode tubes include no stripper cathode, the number of steps and time to remove the uranium from the cathode tubes are minimized, and no additional cathode products have to be harvested. For the dissolution of the dense uranium deposit on the cathode tubes, the anode baskets are emptied, the current density is 0.05 A/cm2, the anode assembly is run at 25- to 40-rpm rotation speed, the voltage cutoff is 1.0 V, and the operating temperature is 500 °C. The near-term plan to manage the dense phase deposits is stripping and investigation of the effect of anode-cathode spacing. Back-up options include the dissolution method and the uranium cutter. Plans for scraper testing include the previously completed test of a scraper with a negative back-rake angle in the 25-in.-diameter HTER. A scraper with a zero back-rake angle will be tested in the Mk-V HTER. In addition, there are plans to test uranium cutters in FY 99. Mk-V HTER activities that have been completed include modification of the original Mk-V HTER for tests in the Chemical Technology Division glovebox. UCL 3 has been prepared for the Mk-V HTER to be tested in the CMT. A purified LiCl-KCl eutectic has been established for above the Mk-V HTER. Finally, MK-V HTER testing in the CMT glovebox has been initiated. Planned near-term tests include continued investigation of the uranium dissolution method to remove dense uranium deposits from the cathode tubes. Parametric tests with the Mk-V and 25-in.-diameter HTERs are also planned. The parametric tests will include observing the effects of current density, anode basket spacing, the separation between the anode baskets and the cathode tubes, separation between the scraper and the cathode tubes, and the effect of rotation speed of the anode drive on the uranium throughput rate. Robert D. Mariani, ANL-W, spoke about the irradiated fuel electrorefining experience. Electrorefining activities through June 1999 will include characterization of the electrorefining process in order to develop uranium product specifications (Success Criterion 2-1). Characterization of fuel dissolution in order to develop metal waste specifications will also be performed (Success Criterion 2-2). The electrorefiner (ER) operating conditions will be specified (Success Criterion 1-1). Twelve driver assemblies will be treated over 3 months 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 fixed operating conditions in the Mark-IV ER (Success Criterion 1 and 1-1). Finally, three blanket assemblies will be electrorefined over 3 months with fixed operating conditions in the Mk-V ER (Success Criterion 1 and 1-1). Significant accomplishments include the electrorefining of four driver assemblies in the Mk-IV in 16 days (Success Criterion 1), and eight driver assemblies were electrorefined in the Mark-IV in 31 calendar days. Process specifications have been drafted (Success Criterion 1-1). Cathode products have been characterized in relation to operating conditions (Success Criterion 2-1). Fuel dissolution has been characterized in relation to operating conditions (Success Criteria 2-1 and 2-2). The Mark-V ER equipment check-out is complete, and initial experiments with the MK-V anode-cathode module are under way. For the Mark-IV ER power supply and electrode configurations, direct transport has the chopped fuel segments in fuel dissolution baskets at the anode, and the steel mandrel at the cathode. For deposition, the dissolved uranium in the cadmium pool is at the anode, and the steel mandrel is at the cathode. The anode-cathode module for the Mark-IV electrorefiner allows higher current, permits collection of fallen product, and eliminates the need for the cadmium pool. R. K. Ahluwalia, ANL-W, presented a talk on electrorefiner model development. This included the GPEC code development and applications, which require gathering of a thermodynamic base for electrorefining, co-dissolution and co-deposition of uranium and zirconium in the direct transport tests, transport phenomena in the deposition tests, and removal of zirconium from the Mark-IV ER. Computational fluid dynamics issues in the ACMs include flow visualization in the Mark-V mock-up and ohmic resistance in different ER configurations. The Oracle database for electrorefining requires trend analyses for deposit morphology, zirconium transport, and current density effects on ACM operation. Future work will include development of a drawdown model, extension of an anodic dissolution model to include U-NM intermetallics, dissolution of blanket fuel, multi-node representation of anode baskets, application of GPEC to ACM geometry, and anode and cathode mass transfer coefficients for ACM configuration. JUNE 26, 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 Consideration of Reviewers' Comments of Committee Report 11:45 Meeting Planning 12:00 Adjournment The committee met from 8:00 a.m. to 12:00 noon at the conference facilities at the Shilo Inn in Idaho Falls, Idaho. The chairman followed the agenda listed above.

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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