Appendix B

ANL Monthly Highlights of the Electrometallurgical Treatment Program

As a mechanism for keeping the committee informed of its R&D progress, ANL has prepared brief monthly progress reports. These “Monthly Highlights” of the Electrometallurgical Treatment Program have been submitted to DOE, which then provided them to the committee. The period covered by the monthly reports included in this appendix extends from September, 1997 to April, 1998.

Electrometallurgical Treatment Program Monthly Highlights for September 1997

Status of Fuel Conditioning Facility (FCF): A Proposed modification to the facility's Technical Safety Requirements was reviewed by the Department of Energy Chicago Operations. During this review, several areas that required additional clarification were identified and Argonne provided the necessary information. The review is complete and final approval is expected shortly. This modification will allow the transfer of items that weigh up to 2200 pounds into and out of the argon cell while process operations continue. Presently, unrestricted transfers are allowed to 250 pounds. When this change is approved, more processing time will be available. The facility semiannual shutdown for stack monitor and criticality monitor calibrations was started. During this shutdown period process operations will be suspended; however, process equipment maintenance will be performed. Also, the Mark-V electrorefiner was transferred into the argon cell and equipment assembly was started. This activity met the project milestone for starting Mk-V electrorefiner installation. An additional can of EBR-II driver fuel (two driver assemblies) were retrieved from the Radioactive Waste and Scrap Facility (RSWF) and transferred to FCF. Currently, the driver fuel for the next eight electrorefiner loads (four months of operations) is stored in the facility.

FCF Driver Treatment Operations: One electrorefiner load of fuel (two assemblies) was chopped and loaded into fuel dissolution baskets. A total of 32 fuel assemblies have been chopped.

The Mark-IV electrorefiner was operated with two anode-cathode pairs in parallel. Two cathodes, 5.2 kg and 3.3 kg, were produced and harvested. This run confirmed the August highlight observation that the cell operating voltage at a constant current is lower when two cathodes are operating than when one anode-cathode pair is operating. This lower voltage would allow the current and therefore the processing rate to be increased when two anode-cathode pairs are operated.

Two batches of uranium cathodes were processed in the cathode processor. The uranium product visually looked very clean and the coating on the process crucible performed very well. Two casting furnace runs were completed and the coating on the casting crucible functioned properly. The repair of the cathode processor vacuum pump has corrected the salt removal problem that was reported in the June and July highlights.

Testing of the ceramic waste process equipment in the newly commissioned glovebox was started. The hot V-mixer, which will be used to sorb the electrorefiner salt into zeolite, has had heater failure problems. The equipment vendor is providing modifications to the heaters. After the initial equipment testing is completed, the ceramic waste process equipment will be operated to establish the processing parameters for production of the ceramic waste form.



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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY Appendix B ANL Monthly Highlights of the Electrometallurgical Treatment Program As a mechanism for keeping the committee informed of its R&D progress, ANL has prepared brief monthly progress reports. These “Monthly Highlights” of the Electrometallurgical Treatment Program have been submitted to DOE, which then provided them to the committee. The period covered by the monthly reports included in this appendix extends from September, 1997 to April, 1998. Electrometallurgical Treatment Program Monthly Highlights for September 1997 Status of Fuel Conditioning Facility (FCF): A Proposed modification to the facility's Technical Safety Requirements was reviewed by the Department of Energy Chicago Operations. During this review, several areas that required additional clarification were identified and Argonne provided the necessary information. The review is complete and final approval is expected shortly. This modification will allow the transfer of items that weigh up to 2200 pounds into and out of the argon cell while process operations continue. Presently, unrestricted transfers are allowed to 250 pounds. When this change is approved, more processing time will be available. The facility semiannual shutdown for stack monitor and criticality monitor calibrations was started. During this shutdown period process operations will be suspended; however, process equipment maintenance will be performed. Also, the Mark-V electrorefiner was transferred into the argon cell and equipment assembly was started. This activity met the project milestone for starting Mk-V electrorefiner installation. An additional can of EBR-II driver fuel (two driver assemblies) were retrieved from the Radioactive Waste and Scrap Facility (RSWF) and transferred to FCF. Currently, the driver fuel for the next eight electrorefiner loads (four months of operations) is stored in the facility. FCF Driver Treatment Operations: One electrorefiner load of fuel (two assemblies) was chopped and loaded into fuel dissolution baskets. A total of 32 fuel assemblies have been chopped. The Mark-IV electrorefiner was operated with two anode-cathode pairs in parallel. Two cathodes, 5.2 kg and 3.3 kg, were produced and harvested. This run confirmed the August highlight observation that the cell operating voltage at a constant current is lower when two cathodes are operating than when one anode-cathode pair is operating. This lower voltage would allow the current and therefore the processing rate to be increased when two anode-cathode pairs are operated. Two batches of uranium cathodes were processed in the cathode processor. The uranium product visually looked very clean and the coating on the process crucible performed very well. Two casting furnace runs were completed and the coating on the casting crucible functioned properly. The repair of the cathode processor vacuum pump has corrected the salt removal problem that was reported in the June and July highlights. Testing of the ceramic waste process equipment in the newly commissioned glovebox was started. The hot V-mixer, which will be used to sorb the electrorefiner salt into zeolite, has had heater failure problems. The equipment vendor is providing modifications to the heaters. After the initial equipment testing is completed, the ceramic waste process equipment will be operated to establish the processing parameters for production of the ceramic waste form.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY Treatment of Metallic Spent Fuels: Uranium electrorefining is the key step in electrometallurgical treatment of spent nuclear fuel. Electrorefining separates pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket requires development of a high-throughput electrorefiner (HTER). Initial testing of the 25-in. diameter HTER with unirradiated N-Reactor fuel has been completed. The HTER was shut down, the anode and cathode assemblies and the uranium collection basket were removed from the vessel, and the quantities of uranium were measured to determine the uranium mass balance in the system. Samples of the fuel remaining in the anode baskets, the granular uranium product in the collection basket, and residual uranium on the cathode tubes were taken and submitted for chemical analysis. Preliminary results indicate that good agreement was achieved between total current passed and the distribution of uranium in the system. Completion of this test represents completion of the milestone, “Demonstrate production-scale HTER with unirradiated N-Reactor fuel.” Further testing of the HTER will be done, using only the inner anode baskets, to determine the effectiveness of the baskets for retaining noble metal fission products. The goal of the next test is to electrotransport at least 98.5% of the uranium for the fuel, while retaining 80% of the zirconium in the anode baskets. Treatment on Oxide Spent Fuels: Oxide spent fuels can be treated by first reducing the oxides to metals, using lithium as a reducing agent in molten LiCl salt. The reduced metals can then be fed to the electrorefiner for treatment. The lithium metal can be regenerated by electrolysis of the lithium oxide so that the lithium and LiCl can be recycled, thus minimizing the quantity of waste. Significant progress was made in preparing for and starting up the next engineering-scale reduction experiment. The final safety review was successfully completed, and preparation of the equipment in the glovebox was completed. Fabrication of the fuel basket, designed to be similar to fuel baskets used in the electrorefiner, was completed, and the UO2 needed for the experiment was transferred into the glovebox and crushed to ~3 mm characteristic dimension. The lithium reducing agent was loaded into porous metal discs. The experiment was initiated by heating the reduction vessel to 650 ºC to melt the LiCl. The lithium-loaded discs were immersed in the molten salt, and the crushed UO2 was loaded into the baskets and placed in the reduction vessel. Samples of the salt were taken at regular intervals and analyzed for Li2O content to monitor the progress of the reduction reaction. Results from this reduction experiment, combined with results from the previous engineering-scale experiments, completes the milestone, “Demonstrate treatment of three zones of TMI-2 core debris,” scheduled for 9/97. Treatment of MSRE Fuel and Flush Salt: The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory in the mid-1960s. Electrometallurgical treatment was one of the options considered for disposal of the fuel and flush salt from the MSRE. The treatment process is essentially an electrochemical titration of the elements from the salt, starting with the most noble elements in the electrochemical series. The small-scale experiment work on this process was completed, and further work was terminated, at the direction of the Department of Energy. The final report on the experimental work has been completed, and the report will soon be distributed. Waste Treatment Process: Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. The waste treatment system consists of high-temperature centrifugal contractors (pyrocontractors) and zeolite ion-exchange columns. Testing of the four-stage pyrocontractor with multiple fission product components was completed, and the final report is being written. A series of tests to examine the effect of salt rate flow on the performance of the zeolite column has been completed. The salt feed to the column in these tests was simulated waste salt of a composition similar to that expected in the electrorefiner. About 1 kg of salt was passed through a bed of ~140 g of dehydrated zeolite beads that had been pre-loaded with LiCl-KCl. The linear velocity of the salt in the column ranged from 0.45 cm/min to 0.7 cm/min in these tests. The effluent salt from the column was white, indicating very low impurity concentrations, and remained white for the duration of the tests. The feed salt was blue due to

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY rare earth chlorides in the simulated waste salt. The effluent salt was sampled for analysis, and the results of the analyses will be used to characterize the performance of the zeolite column. Completion of this work, combined with the completion of the study of temperature effects, reported earlier, meets the requirements of the milestone, “Establish design parameters for zeolite column,” scheduled for September 1997. Waste Form Production and Qualification: Two types of waste forms are produced in electrometallurgical treatment of spent fuels: the ceramic waste form that incorporates the transuranic elements and most of the fission products, and the metal waste form that contains the noble metal fission products in a metal matrix made from the fuel cladding. Studies are being done to optimize the compositions of both the glass-bonded zeolite and the glass-bonded sodalite versions of the ceramic waste form. In the most recent series of tests, glass-bonded sodalite samples were prepared that contained 25, 30, 35, 40, 45, and 50 wt % glass. These samples were subjected to leach testing, using both three-day and 28-day leaching times. The performance of the waste form was measured in terms of cesium release. Cesium is the most difficult fission product to contain. The best performance was achieved with the samples containing 35 wt % glass. The normalized release rate for cesium in those samples was less than 0.2 g/m 2-d. Cesium release was greater from samples having both higher and lower concentrations of glass. The measured cesium release rate was well within the range of acceptable release rates for all of the samples in the 28-day leach tests. These data, combined with earlier data from testing the effects of chloride content and waste form additives, will be used to optimize the waste form composition. These results will contribute to selection of the reference waste form composition, scheduled for October 1997. A test plan for the stainless steel-15 wt % zirconium metal waste form was finalized and issued in September. The plan provides a road map towards establishing a qualification-relevant data base for the metal waste form, including attribute, characterization, accelerated test, and service-condition data for the waste form. The accelerated testing includes high-temperature (200 ºC) testing in deionized water and electrochemically enhanced corrosion tests. These accelerated tests are needed to achieve detectable corrosion rates in a reasonable time scale. Treatment of Aluminum-based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous months. Because of higher priority work, additional experiments on treatment of aluminum-based fuels have been postponed. No work was done in this area during August. Electrometallurgical Treatment Program Monthly Highlights for October 1997 Status of Fuel Conditioning Facility (FCF): The facility's Technical Safety Requirements modification was approved and implemented. This modification allows the transfer of items that weigh up to 2200 pounds into and out of the argon cell while process operations continue. This additional flexibility has helped the final installation of the Mark-V electrorefiner. The installation was completed and the in-cell testing of the equipment has started. The facility semiannual shutdown for stack monitor and critical monitor calibrations was completed. Three additional cans of EBR-II driver fuel (six driver assemblies) were retrieved from the Radioactive Waste and Scrap Facility (RSWF) and transferred to FCF. Currently, the driver fuel for the next nine electrorefiner loads (four months of operations) is stored in the facility.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY FCF Driver Treatment Operations: Two electrorefiner loads of fuel (four assemblies) were chopped and loaded into fuel dissolution baskets. A total of 36 fuel assemblies have been chopped. Since the cadmium pool had accumulated approximately 20 kg uranium, the Mark-IV electrorefiner was operated with two power supplies passing current from the cadmium pool to two individual cathodes in parallel. Two cathodes, 11.7 kg and 10.2 kg, were produced and harvested. This run showed that uranium could be collected at the rate of two hours per kilogram. After this dual power supply run, an electrorefiner load of fuel was processed by direct transport. A 3.8 kg cathode was produced and harvested. This run concluded a series of runs, which processed 12 assemblies of fuel through the electrorefiner during a three month period. The next electrorefiner experiments will provide the initial testing of a new concentric anode cathode module, which will be used for higher throughputs. This initial testing will provide data to confirm the mechanical operability and performance of this equipment. Three batches of uranium cathodes were processed in the cathode processor. The uranium product visually looked very clean and the coating on the process crucible continues to perform very well. Two casting furnace runs were completed and the coating on the casting crucible functioned properly. Treatment of Metallic Spent Fuels: Uranium electrorefining is the key step in electrometallurgical treatment of spent nuclear fuel. Electrorefining separates pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, requires development of a high-throughput electrorefiner (HTER). A new series of tests were started with the 25-in. HTER having the objectives: 1) to demonstrate that the noble metals in the EBR-II spent fuel can be retained in the anode baskets as the uranium is anodically dissolved, and 2) to demonstrate that an improved uranium scraper design can increase the uranium throughput rate by preventing uranium holdup between the anode and cathode. Specifically, the goal was to transport 98.5% of the uranium while retaining 80% of the zirconium in the cladding hulls. The unirradiated fuel, containing 18.5 wt% cladding hulls, uranium, zirconium, molybdenum, ruthenium, palladium, and rhodium, was loaded into the anode baskets. The target throughput rate was equivalent to a current density from the anode to the cathode of 0.07 A/cm2 . This high current density is needed for treatment of the EBR-II blanket fuel. The HTER was operated initially at 0.23 A/cm2 and a 0.45 V limiting potential. After about one hour of operation the potential had increased to the limiting value, so the current was reduced to 0.12 A/cm2. The weight of the fuel segments at the start of the test was 10.185 kg. The theoretical integrated current needed to dissolve all of the uranium was 2473 Amp-hours, and the current passed was 2775 Amp-hours. Samples of the cladding hulls and the uranium product were taken for chemical analysis, but the results are not yet available. This test was the first in a series designed to determine the behavior of noble metal fission products during operation of the electrorefiner, and it contributes to completion of the February 1998 milestone, “Establish Mk-V electrorefining operating parameters.” Treatment of Oxide Spent Fuels: Oxide spent fuels can be treated by first reducing the oxides to metals, using lithium as a reducing agent in molten LiCl salt. The reduced metals can then be fed to the electrorefiner for treatment. The lithium metal can be regenerated by electrolysis of the lithium oxide so that the lithium and LiCl can be recycled, thus minimizing the quantity of waste. The engineering-scale experiment, ES-5, was successfully completed. This experiment was designed to test the reduction of some of the Three Mile Island, unit 2 (TMI-2) core debris, to test equipment design concepts applicable to large-scale reduction systems, and to show complete reduction of UO2 on a practical time scale. The progress of the reduction reaction was monitored by periodic analysis of the LiCl reduction salt for Li2O content. The rate of reduction, as measured by this method, was significantly higher than observed in previous engineering-scale experiments. The increase in reduction rate was achieved by introducing the lithium metal into the melt within a high-surface-area porous stainless steel sponge. This arrangement provided adequate surface area for diffusion of lithium into the salt phase, and thus increased the rate of reduction. Analysis of the experiment components and the reduced metal product will be completed in

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY the following weeks. Completion of this experiment meets the requirements of the milestone, “Demonstrate treatment of three zones of TMI-2 core debris,” scheduled for September 1997. The ES-5 experiment was initiated in September and completed in October. Treatment of MSRE Fuel and Flush Salt: The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory in the mid-1960s. Electrometallurgical treatment was one of the options considered for disposal of the fuel and flush salt from the MSRE. The treatment process is essentially an electrochemical titration of the elements from the salt, starting with the most noble elements in the electrochemical series. The small-scale experiment work on this process was completed, and further work was terminated, at the direction of the Department of Energy. The final report on the experimental work has been completed, and the report is in the publication process. Waste Treatment Process: Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. The waste treatment system consists of high-temperature centrifugal contractors (pyrocontractors) and zeolite ion-exchange columns. Testing of the four-stage pyrocontractor with multiple fission product components was completed, and the final report is being written. Preparations are being made for a series of zeolite column tests at very low flow rates. Previous tests showed an improvement in sorption of rare earths when flow rates were reduced from 1.4 to 0.5 cm/min. Further reduction in flow rate is expected to improve the absorption of alkali and alkaline earth fission products as well. These flow-rate tests are aimed at completion of the milestone, “Establish design parameters for zeolite column,” scheduled for September 1998. Waste Form Production and Qualification: Two types of waste forms are produced in electrometallurgical treatment of spent fuels: the ceramic waste form that incorporates the transuranic elements and most of the fission products, and the metal waste form that contains the noble metal fission products in a metal matrix made from the fuel cladding. A detailed test plan has been developed that will lead to qualification of the metal waste form. The plan includes testing of a variety of waste form compositions that encompass the range expected to be produced during treatment of the EBR-II spent fuel inventory. Production of ingots having the desired range on compositions is preceding on schedule. Ingots numbered SS15Zrl7 (type 316 stainless steel-15 wt% zirconium) and SS05Zr18 (type 316 stainless steel-5Zr-2Nb-1Ru-1Pd) were recently cast, sectioned, and examined by scanning electron microscopy and energy-dispersive spectroscopy. Sample SS15Zrl7 was typical of the stainless steel-15Zr alloy observed earlier, and it contained ferrite, austenite, and the Zr(Fe,Cr,Ni) 2+x intermetallic phase. The microstructure of sample SS05Zr18 contained ferrite and austenite within an intermetallic network composed of ZrFe2-based Laves intermetallics. The noble metals were in solution in these phases, with the exception of a small amount of NbFe2 observed at the ferrite-austenite interfaces. Completion of casting and characterization of this series of waste form alloys is aimed at qualification of the metal waste form for disposal, with an interim milestone, “Issue metal waste form handbook, rev. 0,” schedules for March 1998. A significant improvement in ceramic waste form fabrication has been achieved that effectively eliminates the need for using the hot isostatic press (HIP). A sodalite-glass composite was formed, using granular zeolite and glass powders, and has been successfully sintered to full density without the HIP. Both the salt-loaded zeolite and glass powders ranged in particle size from 74 to 250 microns. These powders were mixed merely by shaking them together for a duration of one minute. The mixed powders were then poured into a graphite mold, tapped gently, and sintered two hours at 850 ºC. A fully dense sodalite-glass composite waste form was produced. The waste forms produced by this simplified method may require a higher glass loading; about 50 wt% glass appears to be required in this process, compared with about 25 wt% required in the HIPing process. However, the process simplification that is achieved may be worth the possibility of increased waste form volume. The results of this test contribute to

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY completion of the milestone, “Issue report on reference ceramic waste form compositions and fabrication conditions,” scheduled for March 1998. Treatment of Aluminum-based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous months. Because of higher priority work, experiments on treatment of aluminum-based fuels were delayed. No work was done in this area during October; however, installation of the engineering-scale aluminum electrorefiner will begin in November. Electrometallurgical Treatment Program Monthly Highlights for November 1997 Status of Fuel Conditioning Facility (FCF): In-cell testing of the Mark-V (Mk-V) electrorefiner is progressing with chemical additions anticipated during early December. Introduction of the Mk-V support equipment continues in parallel with in-cell testing. The transfer of fuel into the facility for the month of November included two (2) driver and eleven (11) blanket assemblies. In-cell stored fuel totals 16 driver assemblies and 13 blanket assemblies. Low enriched uranium product shipped from FCF to storage this month totaled 82 kg. A total of 192 kg of product has been shipped from FCF. FCF Driver Treatment Operations: One electrorefiner load of fuel (two assemblies) was chopped and loaded into anode basket containers. A total of 44 fuel assemblies have been chopped. The initial run (two assemblies) of the new concentric anode cathode module for the Mk-IV electrorefiner was initiated on November 22 with electrotransport of uranium ending on November 29. A number of mechanical operability issues were resolved during this run. The concentric anode cathode module run generated 4.9 kg of product within the product collector with approximately 4.0 kg of product remaining on the cathode. The experience gained with the new concentric anode cathode module will provide data necessary to expedite operation of the Mk-V concentric anode cathode module. Three uranium cathodes batches, consisting of four cathodes, were processed through the cathode processor and casting furnace. These batches resulted in product of high visible quality and good process crucible coating performance. After the casting furnace was modified to increase the surface temperature of the melt, a batch of irradiated cladding hulls was casted into a metal waste form. The resulting ingot was well consolidated and appeared homogeneous. This ingot will be characterized to support the metal waste qualification activity. The following milestones were completed during the month of October: Complete Accelerated Waste Form Glovebox Modifications; Issue MC &A Report on Depleted Uranium Operations and Initial Irradiated Batches; Establish Reference Ceramic Waste Composition. The following milestones were completed during the month of November: Complete Initial Integrated Process Model; Complete Surrogate Demonstration Scale Tests on Baseline Ceramic Waste. Treatment of Metallic Spent Fuels: Uranium electrorefining is the key step in electrometallurgical treatment of spent nuclear fuel. Electrorefining separates pure uranium from the spent fuel, thus reducing

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, requires development of a high-throughput electrorefiner (HTER). The Mark-V HTER that will treat EBR-II blanket fuel in the Fuel Conditioning Facility at Argonne-West will demonstrate a throughput of 150-kg uranium per month. Tests are being done with the 25-in. HTER at Argonne-East to simulate the Mark-V HTER operation. These tests are being done to verify the conditions needed for a throughput rate of 30-kg uranium per day module. The effects of anode basket geometry, electrode operation, cathode scraper design, and operating conditions on the HTER performance will be determined. These tests will also be used to determine anode designs and electrorefining operating conditions that retain simulated noble metal fission products in the anode baskets after the uranium has been transported away from the fuel. The laboratory-scale HTER is also being used in the noble metal retention studies. Tests results so far indicate that the 30-kg/d throughput can be achieved and that the noble metal fission products can be retained in the anode baskets. These tests are aimed at completion of the milestone, “Establish Mk-V electrorefining operating parameters,” scheduled for February 1998. Treatment of Oxide Spent Fuels: Oxide spent fuels can be treated by first reducing the oxides to metals, using lithium as a reducing agent in molten LiCl salt. The reduced metals can then be fed to the electrorefiner for treatment. The lithium metal can be regenerated by electrolysis of the lithium oxide so that the lithium and LiCl can be recycled, thus minimizing the quantity of waste. Preparations for the next engineering-scale experiment (ES-6) are continuing. This experiment will be configured the same as previous engineering-scale experiments, except that the fuel will be held in a MK-V electrorefiner basket, which is thicker than the previous baskets. The results of the ES-6 experiment will be used to evaluate the effect that this thicker basket has on reduction effectiveness and time required. The Mk-V fuel baskets have been obtained from Reactor Engineering Division, and additional preparations have been started, with anticipated start of the experiment in January 1998. The interface between the lithium reduction step and the electrorefining step is important from the standpoint of carry-over of lithium metal or lithium oxide into the electrorefiner. The metal would reduce some of the UCl3, which would then have to be replenished, and the oxide would react with the UCl3 to form UO2. To study this process interface, preparations are being made for a small-scale reduction and electrorefining experiment in the X-141 laboratory. As part of this preparation, the glovebox has been partially decontaminated and refurbished. These reduction/electrorefining experiments are expected to begin in early 1998. Treatment of MSRE Fuel and Flush Salt: The Molten Salt Reactor Experiment (MSRE) was operated at Oak Ridge National Laboratory in the mid-1960s. Electrometallurgical treatment was one of the options considered for disposal of the fuel and flush salt from the MSRE. The treatment process is essentially an electrochemical titration of the elements from the salt, starting with the most noble elements in the electrochemical series. The small-scale experiment work on this process was completed, and further work was terminated, at the direction of the Department of Energy. The final report on the experimental work has been completed, and the report is in the publication process. Waste Treatment Process: Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. The waste treatment system consists of high-temperature centrifugal contractors (pyrocontractors) and zeolite ion-exchange columns. Tests are being done to study the effect of linear velocity of molten salt through the zeolite column on the column efficiency. Previous tests have shown that very low velocities are needed to achieve the desired column performance. An improvement in sorption of rare earths was observed when the flow velocity was reduced from 1.4 to 0.5 cm/min. Sorption of the other cations showed no significant differences. The present tests are aimed at improving the sorption of alkali and alkaline earth cations by lowering the flow velocities further. In the earlier tests, the column was 2.8-cm diameter by 30-cm long. The present tests are being done with a column that is 5-cm diameter, and

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY velocities of 0.4 and 0.2 cm/min were tested successfully. An additional test is planned in which the linear velocity will be reduced to 0.1 cm/min. The chemical analyses from these tests have not yet been completed. The results of these tests will contribute to the completion of the milestone, “Establish design parameters for zeolite column, ” scheduled for September 1998. Waste Form Production and Qualification: Two types of waste forms are produced in electrometallurgical treatment of spent fuels: the ceramic waste form that incorporates the transuranic elements and most of the fission products, and the metal waste form that contains the noble metal fission products in a metal matrix made from the fuel cladding. Samples of ceramic waste form containing plutonium are being prepared. A sample of LiCl-KCl eutectic salt was loaded with PuCl3 to a concentration of 15 mol%. The plutonium-containing salt was blended with zeolite 4A to achieve a salt-to-zeolite ratio of 1:9 on a weight basis (10-g salt absorbed in 90-g zeolite) and a chloride loading of 3.5 chloride ions per unit cell of zeolite (Cl−/u.c.). Several waste form samples were prepared from this blended zeolite to have the same chloride and plutonium loading and varying amounts of glass-57 in the mix. The samples were converted to sodalite by hot uniaxial pressing. They will be used to determine the disposition of plutonium in the waste form and to establish the release rate of plutonium during leaching. The behavior of uranium in the ceramic waste form is also an important issue for waste form qualification. Toward increasing our understanding in this area, three batches of blended zeolite 4A were prepared that contained uranium in the range of 0.05 to 7.5 wt%. The salt-to-zeolite ratio was chosen to give a chloride content of 3.2 ± 0.1 chloride ions per unit cell. The success of the blending operation was measured by the amount of free chloride released during a one-minute wash of blended zeolite. The free chloride was found to be 0.01 to 0.02% for the samples that contained 0.05 and 0.5 wt% uranium. These levels of free chloride are typical of blended zeolite normally prepared for the waste form. The sample with 7.5 wt% uranium exhibited 1.4% free chloride, which is about 100 times higher than the normal amount. This greater free chloride amount indicates that the salt with higher uranium content is sorbed with difficulty. One possible explanation is that the high-uranium salt has different properties, e.g., higher melting point, than the normal salt. Characterization of the high-uranium blended zeolite by X-ray analysis showed that the phases present were zeolite 4A and uranium oxide. Further characterization work is continuing, because the interactions of uranium and transuranium chlorides with the zeolite are important to the waste form properties. This work is aimed at completion of the milestone, “Issue report on disposition of transuranic elements and other highly charged ions in ceramic waste form,” scheduled for completion in June 1998. Treatment of Aluminum-based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous months. Because of higher priority work, experiments on treatment of aluminum-based fuels were delayed. No work was done in this area during November; however, installation of the engineering-scale aluminum electrorefiner will begin as soon as production of UCl3 for startup of the Mk-IV electrorefiner is complete. Electrometallurgical Treatment Program Monthly Highlights for December 1997 Status of Fuel Conditioning Facility (FCF): Two hundred and forty ingots of eutectic salt were added to the Mk-V electrorefiner. The chemical additions and equipment checkout at the 500 ºC operating temperature should be completed by the end of January. Initial testing with depleted uranium plates should begin in early February.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY The out-of cell remote qualification of the blanket element chopper was completed. The equipment should be installed in the argon cell during January. Six driver and six blanket assemblies were transferred into the Fuel Conditioning Facility. In-cell stored fuel totals nineteen driver and seventeen blanket assemblies. FCF Driver Treatment Operations: One electrorefiner load of fuel was chopped and loaded into fuel dissolution baskets. Two driver assemblies were processed in the Mk-IV electrorefiner. This run directly transported the uranium from fuel dissolution baskets to a solid cathode mantrel. The run parameters were based on modeling results and the objective was to increase uranium throughput while reducing zirconium in the cladding hull. A 7 kg uranium deposit was produced with 53 hours of current passing time. After the direct transport run, a solid cathode was produced where the uranium from the cadmium pool was transported to a solid cathode. A 7 kg cathode was collected and harvested. The cathode processor product from two electrorefiner cathodes was cast into a low enriched uranium ingot. The casting furnace continues to operate without any significant problems. After this run, the casting furnace was modified so that its batch size could be increased from a 37 kg ingot to a 55 kg ingot. This modification will support the operations when both blankets and drivers are being processed. Treatment of Metallic Spent Fuels: Uranium electrorefining is the key step in electrometallurgical treatment of spent nuclear fuel. Electrorefining separates pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities of spent fuel, such as the EBR-II blanket, requires development of a high-throughput electrorefiner (HTER). A test was competed with the 25-in. diameter HTER, the objective of which was to determine the degree of retention of the noble metals in the anode baskets while transporting essentially all of the uranium to the cathode. The goal of the test was to transport at least 98.5% of the uranium while retaining at least 80% of the zirconium. The test was interrupted after about 80% of the uranium had been transported for the purpose of analyzing the anode residue. It was found that, within analytical error, all of the zirconium and all of the noble metals remained in the anode after 80% of the uranium had been transported. The test was continued until more than 98.5% of the uranium had been transported, and samples of the fuel segments were submitted for chemical analysis. The results of these analyses are not yet available. This work is aimed at completion of the milestone, “Issue status report on mass balance in the high-throughput electrorefiner,” scheduled for March 1998. Treatment of Oxide Spent Fuels: Oxide spent fuels can be treated by first reducing the oxides to metals, using lithium as a reducing agent in molten LiCl salt. The reduced metals can then be fed to the electrorefiner for treatment. The lithium metal can be regenerated by electrolysis of the lithium oxide so that the lithium and LiCl can be recycled, thus minimizing the quantity of waste. The process interface between the reduction step and the uranium electro-refining step is of some concern, because of interaction of UCl3 with lithium and Li2O that could be carried over with the reduced metal. Experiments are being planned to examine the extent and consequences of these interactions. A small electrorefiner has been constructed to study these issues related to electrorefining the product from the reduction step. The electrorefiner is located in the same glovebox with the reduction vessel, so the reduced product can be transferred directly to the electrorefiner for testing. Treatment of Aluminum-based Fuels: Demonstration of the feasibility of electrometallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY as low-level waste. Good separation of aluminum from uranium was demonstrated in previous months. Because of higher priority work, experiments on treatment of aluminum-based fuels were delayed. No work was done in this area during December; however, installation of the engineering-scale aluminum electrorefiner will begin when production of UCl3 has been completed for startup of the Mk-IV electrorefiner at ANL-W. Waste Treatment Process: Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. The waste treatment system consists of high-temperature centrifugal contractors (pyrocontractors) and zeolite ion-exchange columns. Work on pyrocontractor development has been completed, and the final report will be issued in January. Studies on zeolite ion-exchange columns examined the effect of very low flow rates on retention of fission product elements. Previous tests showed an improvement in sorption of rare earths when the flow velocity was reduced from 1.4 to 0.5 cm/min. Sorption of the other cations showed no significant differences. We are now looking for improvements in retention of alkali and alkaline earth cations by lowering to flow velocity still further. Additional tests were completed at flow velocities of 1.5, 0.41, 0.23, and 0.10 cm/min. A significant improvement in retention of alkali and alkaline earth cations was achieved at the lower flow velocities. Breakthrough of all of the cations increases at the point of rare earth cation breakthrough. There was an observable change in the color of the effluent salt as the tests progressed, which coincided exactly with breakthrough of the rare earths, as was expected. If actinide behavior mimics that of the rare earths, this color change may be a useful rapid qualitative indicator of breakthrough during hot cell operation of the zeolite column. This work is aimed at completion of the milestone, “Establish design parameters for zeolite column,” scheduled for September 1998. Waste Form Production and Qualification: Two types of waste forms are produced in electrometallurgical treatment of spent fuels: the ceramic waste form that incorporates the transuranic elements and most of the fission products, and the metal waste form that contains the noble metal fission products in a metal matrix made from the fuel cladding. A series of immersion corrosion tests based on the Materials Characterization Center standard test, MCC-1, were performed on the metal waste form alloy. The test samples were polished to a 600-grit finish, immersed in J-13 well water in sealed Teflon vessels, and placed in a 90 ºC furnace for periods from one to three years. At the end of the tests, the samples were removed for weighing and analysis of the solutions and the samples. The final weight measurements showed that the one-year samples had a slight weight loss ranging from 0.2 to 1.3 mg. The three-year samples showed a slight weight increase of 0.4 to 0.5 mg (~0.01%). The reasons for these small weight losses and gains are not well understood, and they may be artifacts of the weighing techniques and instruments used. The pH of the final solutions ranged from 7.8 to 8.8, which is similar to the pH of the original solution, which was 8.2. The test solutions are being analyzed to determine elemental leaching behavior. The leach rates represented by these data are extremely low, compared to leach rates of other waste forms. These leach data will become part of the broad database that will be used to qualify the metal waste form for ultimate disposal. Production of 13 simulated metal waste form ingots has been completed to provide samples for qualification testing of the metal waste form. Three of the ingots were prepared to generate data on baseline composition and to study the effect of zirconium content deviation from the baseline 15 wt% zirconium composition. Three were prepared to study the effect of cooling rate variations on the microstructure and leaching behavior. Two ingots were prepared to simulate the alloying behavior of severely oxidized stainless steel, and one was produced without zirconium to study noble metal distribution and leaching behavior of stainless steel. This work was done to supply samples for a variety of waste form testing methods specified in the Metal Waste Form Test Plan, which was designed to qualify the metal waste form for ultimate disposal.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY The behavior of the transuranium elements in the ceramic waste form is an important issue for waste form qualification. Toward increasing our understanding in this area, zeolite 4A was blended with plutonium-containing salt, and the salt-loaded zeolite was examined by X-ray diffraction. Three important observations were made from these tests: 1) the zeolite framework structure does not decompose in the presence of plutonium in the salt, 2) plutonium species that are incorporated into the zeolite lattice do not appear in the X-ray diffraction patterns, i.e., they do not form a new phase, and 3) residual water in the “anhydrous” zeolite reacts with the PuCl3 to form PuO2 or PuOCl, depending on the plutonium-to-water ratio. The amount of oxide or oxychloride produced is controlled by the amount of water remaining in the anhydrous zeolite. These tests are being done to complete the milestone, “Issue report on disposition of transuranic elements and other highly charged ions in ceramic waste form,” scheduled for completion in June 1998. Electrometallurgical Treatment Program Monthly Highlights for January 1998 The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main work breakdown structure elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. WBS 1.0 Treatment Operations: Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility and the Hot Fuel Examination Facility. The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel. WBS 1.1 Driver Treatment: The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mk-IV electrorefiner will reach three weight percent. At the end of January, forty-seven driver assemblies had been chopped and forty-four driver assemblies had been introduced to the electrorefiner. During January, three driver assemblies were chopped and two assemblies were processed in the electrorefiner. The electrorefiner testing concentrated on concentric anode cathode module which is the key component for high throughput operations. Several mechanical and process issues were identified and this knowledge will be used to design modifications for the Mk-V electrorefiner. Since the Mk-V electrorefiner is ready for testing, future concentric anode cathode module testing will be conducted in the Mk-V, which is covered in the WBS 1.2 Blanket Treatment, A seventeen kilogram charge of solid cathodes was run through the cathode processor. This run produced a well-consolidated ingot that will be converted to a low enriched uranium product in the casting furnace. This run is the initial test of equipment modifications that will increase the normal batch sizes for 10 kg uranium to 17 kg uranium. These modifications will increase the throughput capacity by seventy percent. WBS 1.2 Blanket Treatment: The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mk-V electrorefiner. Operating instructions and operator training are proceeding in parallel with the equipment qualification and installation that is discussed in WBS 2 activities. The blanket operations are planned to start in March 1998. WBS 1.3 Metal Waste: Three batches (two assemblies) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. Two successful

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY An Argonne readiness assessment for blanket operations was initiated and is scheduled for completion in April 1998. Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl to effect this conversion, and the Li2O that is produced dissolves in the molten salt. The salt and lithium metal are recycled by electrochemical deposition of lithium at a cathode and evolution of oxygen at an anode. It is important to maximize the rate of this electrochemical process to achieve good throughput and reasonable equipment sizes. A series of electrochemical experiments have been completed to establish the range of suitable operating voltages for the platinum anode. If the operating voltage is too high, chlorine is produced and the platinum is dissolved as the chloride. If the voltage is too low, the throughput is not satisfactory. In these experiments, the Li2O concentration in the LiCl was varied between 3 wt% and 0 wt%. The initial results show that when Li2O is present in the salt, the process can be operated at voltages approaching 3.7 V without damage to the anode. This voltage is higher than the theoretical decomposition potential for LiCl. Other, less expensive and equally effective anode materials, such as doped SnO 2, are also being studied for this application. This work is directed toward completion of the milestone, “Complete engineering-scale demonstration of integrated reduction and electrowinning process,” scheduled for September 1998. Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electro-metallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous work. The work in this area during February included preparation of the engineering-scale aluminum electrorefiner and the electrolyte salt for beginning aluminum electrorefining in March. Electrometallurgical Treatment Program Monthly Highlights for March 1998 The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main work breakdown structure elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. WBS 1.0 Treatment Operations: Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility and the Hot Fuel Examination Facility. The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel. WBS 1.1 Driver Treatment: The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mk-IV electrorefiner will reach three weight percent. At the end of March, fifty-three driver assemblies had been chopped and fifty-two driver assemblies had been introduced to the electrorefiner. During March, two driver assemblies were chopped. In the electrorefiner, two cathodes were produced by transporting uranium from the cadmium pool to a single cathode mandrel. These two runs studied collection efficiency and product purity when the uranium is being recovered from the cadmium pool. After recovering the uranium from the cadmium pool, the electrorefiner was brought to an oxidizing

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY condition and samples taken for material balances. Afterwards, a batch of four driver assemblies were introduced to the electrorefiner. This batch is being operated in the proposed configuration for the repeatability demonstration. Four driver assemblies in two sets of anode baskets are electrotransported to one solid cathode mandrel. After approximately one half the current is passed, the mandrel and deposit is removed and a second mandrel is inserted. While the first deposit is being harvested, the second mandrel collects the remaining uranium in anode baskets. At the end of March, the first cathode had been removed and electrotransport to the second mandrel was progressing. This new run configuration is called dual anodes with serial cathodes. A batch of cathodes was run through the cathode processor. This product will be processed into a low enriched uranium product. WBS 1.2 Blanket Treatment: The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mk-V electrorefiner. Operating instructions and operator training are proceeding in parallel with the equipment qualification and installation that is discussed in WBS 2 activities. Due to the minor checkout problems with the equipment, and overhead handling system problems, the start of the blanket operations is now forecast for May 1998. WBS 1.3 Metal Waste: Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. The first full batch of injection cast samples from the irradiated cladding hulls was prepared and shipped to the analytical chemistry laboratory. The chemical analyses will support the metal waste qualification activities (WBS 6.0). WBS 1.4 Ceramic Waste Operation with Irradiated Materials: After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the Hot Fuel Examination Facility where the salt and fission products will be immobilized in ceramic waste samples. This activity is not scheduled to begin until February 1999. WBS 1.5 Facility Operations: Four driver assemblies were received from the Radioactive Scrap and Waste Facility. At the end of the reporting period, 17 driver and 27 blanket assemblies (25 irradiated, 2 unirradiated) were stored in the Fuel Conditioning Facility air cell. The two unirradiated blanket assemblies will be used for initial start-up testing of the blanket equipment before irradiated blanket treatment. The semi-annual Technical Safety Requirements equipment calibrations were completed in mid-March and full process operations resumed. WBS 2.0 Equipment and Facility Modifications: This work element covers the engineering design. fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities. For blanket operations, a new blanket chopper and the Mk-V electrorefiner are needed to treat the larger blanket fuel. The blanket chopper started testing by chopping steel rods. The Mk-V electrorefiner power supplies were tested. Preparation for depleted uranium plate operations was started. Blanket disassembly equipment installation was completed. WBS 3.0 Treatment Process Development: During the demonstration, new process improvements and investigations of problem areas will be investigated in existing developmental and engineering laboratories. The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of large quantities

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY of spent fuel, such as the EBR-II blanket, requires use of a high-throughput electrorefiner (HTER). A recent test with the 25-in HTER was done to study the behavior of noble metal fission products. The operating conditions used in this test were: a maximum cutoff voltage of 0.45V, an average current density of 100 mA/cm2 , and a rotation speed for the anode baskets of 40 rpm. Chemical analysis of the samples after the test showed that the concentrations of noble metals in the uranium product were: 80 ppm Zr, <0.2 ppm Mo, <0.3 ppm Ru, <1 ppm Pd, and 0.08 ppm Rh. The cladding hulls contained: 1.2% of the initial U, 94.3% of the Zr, 88.2% of the Mo, 93.8% of the Ru, 100% of the Pd, and 75% of the Rh. These results verify that greater than 98.5% of the uranium can be separated from the fuel while retaining greater than 80% of the zirconium in the anode baskets, which was the objective of this noble-metal-retention test. Another series of tests was done to determine the effects of current density anode assembly rotation speed, and cathode scraper design on electrorefiner performance. The performance indicators were scraper effectiveness, ability to sustain operation for a long period of time, complete removal of uranium from the fuel cladding, and good characteristics of the uranium product. These tests showed that long-term, sustained operation can be achieved at an average current density of 100 mA/cm2 . Post-test examination showed that a dense uranium layer had formed, between the scrapers and the stainless steel cathode tubes, that caused difficulty in separating the anode assembly from the cathode assembly. These results suggest that it will be necessary to strip the dense uranium layer off the cathode tubes by reversing the cell current for a short time at the end of a long period of sustained operation. This stripping operation can be done in a number of ways, and the next series of experiments will examine a variety of stripping techniques, as well as confirm the recent test results. A parametric study is planned with the improved concentric anode cathode module for the Mark-V electrorefiner that will be used for treating EBR-II blanket fuel in the Fuel Conditioning Facility (FCF). The improvements in the Mark V design were based on test results with the 25-in. HTER. These results led to improvements in scraper design, more space between the anode baskets, ceramic covers over the cathode tube support bars, and increased distance between the fuel in the anode baskets and the cathode tube support bars. This work is being done in support of operation of the Mark-V electrorefiner in the FCF. In support of cathode processor operations, a series of tests were conducted with different operating scenarios that could accelerate the equipment cooldown. These tests showed that it is possible to decrease the total batch processing time from 36 hours to 24 hours. WBS 4.0 Process Modeling and Analysis: These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations. A draft report on mass balances was reviewed by an outside specialist in the material accountancy area. These comments will be incorporated into the final version of this report. WBS 5.0 Metal Waste Treatment Development: The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on larger casting furnace for inventory operations. Work has been essentially completed on development of the metal waste form composition and preparation of waste form samples for characterization and qualification. This task is now concentrated

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY on development of crucible materials for melting the metal waste form. Yttria crucibles have served well in melting small batches of waste form alloys; however, yttria is subject to cracking by thermal shock so that each crucible was good for only one melt/freeze cycle. Beryllia has been ruled out as a potential crucible material for the waste form alloy, because it reacts substantially with zirconium. A contract has been placed with Thermal Technologies, Inc. of California to develop crucibles that are capable of multiple melt/freeze cycles in the stainless steel-zirconium system. A very large array of candidate crucible materials have been tested, and the selection has been narrowed down to various hafnium oxide and hafnium nitride compounds that appear very promising. Thermal Technologies, Inc. is now developing mid-size crucibles for testing at ANL with the stainless steel-zirconium waste form alloy. WBS 6.0 Metal Waste Qualification Testing: The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling. The leach behavior of the metal waste form, stainless steel - 15 wt% Zr (SS-15Zr), is being studied, using the MCC-1 procedure at 90 ºC and accelerated testing at 200 ºC. The 90 ºC tests resulted in negligible corrosion, i.e., only specimens exposed to the solution for 1029 days showed any visible evidence of surface tarnish. All the other specimens retained shiny surfaces with minor discoloration. There was no evidence of preferential attack at the as-cast porosity visible on the specimen surfaces. The final weight measurements were with the error range of the balance. The accelerated tests at 200 ºC in deionized water showed elemental losses that were also very small. Comparison of the 200 ºC data with data from the 90 ºC tests indicate that the leach behavior of the alloy was similar under both conditions. The higher temperature apparently accelerates the corrosion rate without altering the corrosion mechanism. The higher temperature test may be of value as a product consistency test for the metal waste form, during production operations. WBS 7.0 Ceramic Waste Treatment Development: The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the Hot Fuel Examination Facility with salts from the Mk-IV electrorefiner. The necessary processes, materials, and demonstration equipment are being developed to implement these processes in the Hot Fuel Examination Facility. Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. A series of tests have been done to determine the range of stability of the zeolite structure. We had observed that, under some conditions, zeolite will convert to the more stable LiAlSiO4 that has no alpha or beta cages to retain salt. Three samples were tested: one with 0.19 wt% water was heated with salt to 550 ºC for 24 hours; the second with 0.19 wt% water was heated without salt to 550 ºC for 24 hours, then heated with salt for an additional 24 hours; and the third with several percent water was heated with salt for 24 hours. Samples one and three retained the zeolite structure, while sample two was converted to LiAlSiO4. These results suggest that the pre-heating may drive off residual water, which de-stabilizes the zeolite structure. Subsequent heating with salt then completes the conversion to LiAlSiO4. The good news is that, when this LiAlSiO4 is mixed with glass frit and heated to 750 ºC, it converts to sodalite, which is the reference ceramic waste form.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY A new type of zeolite, zeolite X, is being examined as a possible improved waste form material. Like zeolite A, zeolite X contains beta cages, and it has a much larger central cavity (12 angstroms) with larger (7 angstroms) pore openings. This larger cavity should have a higher capacity for salt and should improve the ion exchange kinetics. Unlike zeolite A, the salt is readily washed out of the zeolite X structure, making it unsuitable as a waste form candidate. However, zeolite X is converted to sodalite on heating, so the reference waste form would remain the same as that produced from zeolite A. WBS 8.0 Ceramic Waste Qualification Testing: The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing. Qualification testing has started on samples of the reference glass bonded sodalite. WBS 9.0 Repository Performance Assessment Modeling: A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the wastes to be generated from the process will perform acceptably when ultimately disposed in a geologic repository. An initial ceramic waste dissolution model was reviewed for its applicability for the ceramic waste. This model will be revised to add a zeolite degradation and release mechanism rather than a simple glass dissolution model. WBS 10.0 Environmental and Safety Support Tasks: These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities. An Argonne readiness assessment for blanket operations was initiated and is scheduled for competion in April 1998. Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl to effect this conversion, and the Li2O that is produced dissolves in the molten salt. The salt and lithium metal are recycled by electrochemical deposition of lithium at a cathode and evolution of oxygen at an anode. It is important to maximize the rate of this electrochemical process to achieve good throughput and reasonable equipment sizes. A series of electrochemical experiments have been completed to establish the range of suitable operating voltages for the platinum anode. If the operating voltage is too high, chlorine is produced and the platinum is dissolved as the chloride. If the voltage is too low, the throughput is not satisfactory. In these experiments, the Li2O concentration in the LiCl was varied between 3 wt% and 0 wt%. The initial results show that when Li2O is present in the salt, the process can be operated at voltages approaching 3.7 V without damage to the anode. This voltage is higher than the theoretical decomposition potential for LiCl. Other, less expensive and equally effective anode materials, such as doped SnO 2, are also being studied for this application. This work is directed toward completion of the milestone, “Complete engineering-scale demonstration of integrated reduction and electrowinning process,” scheduled for September 1998. Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electro-metallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been initiated in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Good separation of aluminum from uranium was demonstrated in previous work. The

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY work in this area during February included preparation of the engineering-scale aluminum electrorefiner and the electrolyte salt for beginning aluminum electrorefining in March. Electrometallurgical Treatment Program Monthly Highlights for April 1998 The electrometallurgical program has two components: the EBR-II Spent Fuel Treatment Demonstration project and its adaptation to a variety of DOE spent fuel types. The monthly highlights are divided between the main work breakdown structure elements for the project plus two additional tasks: treatment of oxide spent fuels and treatment of aluminum-based fuels. The technical highlights and milestone status have been combined to provide an overall picture of the program. The attached milestone schedule shows the original scheduled dates and forecasted (f) dates. The original milestone target dates were established in the May 7, 1998, Work Breakdown Structure. For WBS 1.0 and 2.0 milestones, forecasted milestones are tied directly to the project schedule. The other milestones are updated monthly based on progress. WBS 1.0 Treatment Operations: Electrometallurgical treatment technology will convert the highly enriched uranium and the reactive bond sodium in EBR-II fuel into low enriched uranium product, ceramic waste and metal waste. This work element involves the demonstration equipment operations in the Fuel Conditioning Facility (FCF) and the Hot Fuel Examination Facility (HFEF). The process steps will be operated in an integrated manner to demonstrate the economic and technical feasibility of the process with spent irradiated fuel. WBS 1.1 Driver Treatment: The demonstration will treat 100 driver assemblies so that the processes can be demonstrated in an integrated system and fission product loading in the Mk-IV electrorefiner will reach three weight percent. At the end of April, 58 driver assemblies had been chopped and 56 driver assemblies had been introduced to the electrorefiner. During April, five driver assemblies were chopped. In the electrorefiner, the second cathode of the dual anode serial cathodes (see March Highlights) was removed. This batch was finished by depleting the cadmium pool with a deposition run. The second batch (four driver assemblies) was processed with the dual anode series cathodes configuration. This run also produced three cathodes. The significant achievement was that eight driver assemblies were processed within thirty-one days. This rate is twice the demonstration rate of four assemblies per month. During April, 47.5 kg of cathodes were processed through the cathode processor and converted to 39 kg of 50% enriched uranium product. The casting furnace produced 103 kg low enriched uranium product (<20% 235U). These process rates are equivalent to processing the electrorefiner product from greater than eight driver assemblies per month which is twice the demonstration goal rate. WBS 1.2 Blanket Treatment: The EBR-II blanket assemblies contain 47 kg of uranium each and will demonstrate high throughput rates in the Mk-V electrorefiner. The internal readiness assessment for blanket operations was completed and the DOE review is scheduled for the first week in May. The Mk-V electrorefiner concentric anode-cathode module (ACM) was tested with a 10 kg depleted uranium charge. After approximately eight hours operation, the ACM stopped rotation because a uranium deposit jammed between the scraper and cathode tube. Different process conditions were tested that could remove and control this uranium deposit; however, sustained operations could not be achieved. This run was terminated so efforts could be focused on the readiness review items for blanket operations. Also, a new ACM is expected to be received during the first week of May. The new ACM changes the direction of rotation so that uranium is removed from the cathode tube where a large area is available for settling

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY into the product collection basket. This new design was based on the operating experience in the glovebox high throughput electrorefiner (WBS 3.0). Due to these process problems, the blanket treatment is now forecast to start in June 1998. The overall strategy for blanket operations was reexamined to reflect our experience from initial testing. Blanket operations are planned to progress at one fuel assembly (two ACMs) per month for seven months, two assemblies (four ACMs) per month for four months and 3.2 assemblies (six ACMs) per month for three months. Although the schedule shows this can be completed in a shorter time (by May 31, 1999), process improvements will need to be implemented to meet this schedule. Since the time is so limited, the decision was made to test an identical ACM (WBS 3.0) in the glovebox facility at ANL-E. This concurrent testing should accelerate implementation and testing of improvements. WBS 1.3 Metal Waste: Three batches (two assemblies each) of irradiated cladding from driver treatment operations will be converted into typical metal waste forms for waste qualification. A metal waste ingot was sampled by using a core drilling technique. These samples will be compared to the injection cast sample so that a representative sampling method can be selected for the metal waste qualification activities (WBS 6.0). WBS 1.4 Ceramic Waste Operation with Irradiated Materials: After 100 driver assemblies are treated in the Mk-IV electrorefiner, a portion of the salt will be transferred to the HFEF where the salt and fission products will be immobilized in ceramic waste samples. This activity is not scheduled to begin until February 1999. WBS 1.5 Facility Operations: Ten driver assemblies were received from the Radioactive Scrap and Waste Facility. This driver fuel receipt rate is equivalent to what would be required for the proposed inventory operations. At the end of the reporting period, 21 driver and 27 blanket assemblies (25 irradiated, 2 unirradiated) were stored in the FCF air cell. The two unirradiated blanket assemblies will be used for initial start-up testing of the blanket equipment before irradiated blanket treatment. WBS 2.0 Equipment and Facility Modifications: This work element covers the engineering design, fabrication, assembly and testing activities that are required to implement new process equipment, equipment improvements or facility modifications that support operations or development activities. For blanket operations, a new blanket chopper and the Mk-V electrorefiner are needed to treat the larger blanket fuel. The engineering packages for the blanket operations equipment were completed. The equipment is ready for irradiated operations as soon as five minor open items are completed. WBS 3.0 Treatment Process Development: During the demonstration, new process improvements and investigations of problem areas will be investigated in existing developmental and engineering laboratories. The key step in electrometallurgical treatment of spent nuclear fuel is electrorefining to separate pure uranium from the spent fuel, thus reducing the volume of high level waste. Treatment of uranium-zirconium alloy fuel, such as the EBR-II driver fuel, can result in accumulation of zirconium in the electrorefiner. Experimental work is being done in the laboratory-scale electrorefiner to explore methods for removal of this zirconium from the full-scale (Mark-IV) electrorefiner in the FCF. During one recent zirconium electrotransport test, an unusual voltage pattern was observed when the cell was placed on open circuit. The cell voltage, which is normally negative, moved to a positive value on open circuit, then drifted toward a negative value. The cause of this voltage change has not been determined. Zirconium electrotransport was continued in the test until the −50 V cutoff potential was reached. Visual examination showed that all of the zirconium had been removed from the anode compartment. There was

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY a discrepancy between the current passed and the amount of zirconium transported, based on the reaction Zr+4 ⇒ Zr0 . It appears that the oxidation state of zirconium in the salt phase is less than +4 and/or that the material deposited at the cathode remains partially oxidized. Additional experiments are underway to further examine the complex chemistry of the zirconium system. Treatment of large quantities of spent fuel, such as the EBR-II blanket, requires use of a high-throughput electrorefiner (HTER). The 25-in HTER was loaded with unirradiated N-Reactor fuel segments, and two series of uranium electrorefining experiments were completed. In the first series, 74.62 kg uranium was dissolved from the anode baskets, 64.78 kg uranium was found in the collection basket, and 9.84 kg was carried out of the HTER into the containment crucible. In the second series, 5.95 kg uranium was dissolved, 5.45 kg was collected in the basket, and 0.50 kg was carried into the containment crucible. The uranium “losses” to the crucible were caused by salt circulation through the flow vents in the outer cathode tube. These vents will be plugged to eliminate these “losses” in future tests. Reporting of these results met the milestone, “Issue status report on HTER mass balance,” scheduled for March 1998. Based on earlier results with the 25-in HTER, using unirradiated fuel, the scrapers were redesigned to allow the uranium dendrites to fall through the open spaces between the anode baskets. This design change allowed achievement of current densities that were high enough to produce a throughput of 30 kg/d with the 25-in unit. Post-test examination of this electrorefiner showed that it will be necessary to conduct periodic stripping (reversing the polarity of the electrorefiner to remove dense uranium deposit from the cathode) to achieve sustained operation at the 30 kg/d rate. The post-test examination also showed that it is possible to remove essentially all of the dense uranium deposit from the cathode tubes in preparation for the next batch of spent fuel. Experiments designed to investigate and refine these current-reversal techniques are planned for the next series of tests with the 25-in HTER. WBS 4.0 Process Modeling and Analysis: These activities develop and apply models to improve the understanding of various process steps; to help in design of equipment and selection of process variables; to evaluate the data on performance of the engineering-scale equipment; to provide support in planning of test campaigns; and to conduct operations. The FCF material control and accountancy program was reviewed as part of the DOE Safeguards and Security Inspection of ANL-W. An overall rating of satisfactory was assigned with no findings. The status report of the anode-cathode modeling should be issued in May 1998. WBS 5.0 Metal Waste Treatment Development: The noble metal fission products and undissolved cladding hulls are immobilized into a stainless steel-zirconium alloy for geologic repository disposal. In support of waste qualification activities, small samples of the metal waste are being produced so they can be characterized to establish the performance. Also, equipment is being developed and tested in various laboratories to support design efforts on larger casting furnace for inventory operations. Work has been essentially completed on development of the metal waste form composition and preparation of waste form samples for characterization and qualification. This task is now concentrated on development of crucible materials for melting the metal waste form. Yttria crucibles have served well in melting small batches of waste form alloys; however, yttria is subject to cracking by thermal shock so that each crucible was good for only one melt/freeze cycle. Beryllia has been ruled out as a potential crucible material for the waste form alloy, because it reacts substantially with zirconium. Alternative materials, including hafnium oxides and nitrides, have been tested in contact with the molten waste form alloy at 1600 ºC. These, and other, test samples are being analyzed by SEM to determine the extent of any reactions at the ceramic/metal interface.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY WBS 6.0 Metal Waste Qualification Testing: The metal waste form attributes and fission product release mechanisms and rates are being quantified to support repository performance modeling. Testing on the leaching characteristics of the waste form continues. WBS 7.0 Ceramic Waste Treatment Development: The electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to immobilize fission products and transuranium elements for disposal. The necessary processes, materials and demonstration equipment are being developed and tested so these waste treatment processes can be demonstrated in the HFEF with salts from the Mk-IV electrorefiner. The necessary processes, materials, and demonstration equipment are being developed to implement these processes in the Hot Fuel Examination Facility. Electrolyte salt is periodically removed from the electrorefiner and passed through a waste treatment system to remove fission products and transuranium elements for disposal. The electrolyte salt can be passed through a zeolite column to remove fission products and transuranium elements, which will allow the salt to be recycled to the electrorefiner. A variety of zeolite types have been tested for this column application, including zeolite A and chabazite. Another type of zeolite, zeolite X, is now being tested as a potential improvement over zeolite A. Zeolite X has a slightly larger alpha cage, which could improve the kinetics of ion exchange, yet when it is converted to sodalite, it would be equivalent to sodalite prepared from zeolite A. In these early tests, dehydrated zeolite X beads were contacted at 550 ºC with molten simulated waste salt that was high in neodymium. The beads were immersed in the salt for 24 hours, then separated from the salt and cooled to room temperature. Post-test analysis showed that the uptake of neodymium was comparable to that for zeolite A. Additional testing of zeolite X is planned, using salt that simulated electrorefiner conditions after treatment of 300 EBR-II driver fuel assemblies. Development of the ceramic waste form required determination of the minimum glass content that would be acceptable for the glass-bonded zeolite waste form. An internal milestone for this determination was met by distribution of a memo recommending a glass content of 25 wt%. This recommendation was based on a series of samples that were prepared with glass loadings ranging from 0 to 40 wt%. The criterion used for this determination was cesium release from the waste form during 3-day, MCC-1 type leach tests of these samples. The samples were prepared, using the reference fabrication method and composition. The cesium content of the simulated waste salt was representative of that expected in the electrorefiner salt after treating 300 driver assemblies. Cesium loss from the 20 wt% glass samples was comparable to that from the 25 wt% and higher samples; however, as a safety margin, the minimum glass loading of 25 wt% was recommended for the reference glass-sodalite waste form. Ceramic waste form samples containing 238Pu will be fabricated for use in assessing the consequences of alpha damage. Samples containing 239Pu have been fabricated to help establish the optimum fabrication conditions for use with the 238Pu-containing samples. Unlike the reference ceramic waste forms, the plutonium-bearing samples were prepared by hot uniaxial pressing (HUP). Successful fabrication was achieved by pressing the salt-loaded zeolite at 775 ºC for three hours. Samples were prepared both at ANL-W and ANL-E, and some of the samples were exchanged for characterization. In general, chloride release and overall mass loss was higher for these HUPed samples than for the reference ceramic waste form. Despite these deficiencies, plutonium retention was quite good. It is too early to draw any conclusions from these data. The most significant result is that successful preparation of plutonium-bearing waste forms has been achieved at both sites. With these encouraging results, ANL-W is one step closer to preparation of the 238Pu-bearing test samples. WBS 8.0 Ceramic Waste Qualification Testing: The ceramic waste form is being characterized so that its performance in different repository conditions and scenarios can be assessed. This work characterizes hot

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY uniaxial pressing samples and laboratory scale and demonstration scale samples from hot isostatic pressing. The goal of the ceramic waste form qualification task is to evaluate the applicability of standard durability tests and to provide behavior testing and associated activities in support of qualification of this waste for disposition in a high-level waste repository. The qualification testing will investigate the long-term corrosion behavior of the ceramic waste form with emphasis on repository relevance. A report on Ceramic Waste Form test methods was issued in accordance with DOE Project milestones. This report primarily discusses testing methods that have been developed over the past few years. Since the decision to allow zeolite A to convert to sodalite during the hot isostatic pressing process was made in October 1997, most of the previous work was performed with zeolite 4A or 5A containing materials. However, the conclusions from the test method development work should be completely applicable to the qualification testing of the reference ceramic waste form for the EBR-II demonstration project. The main purpose of this report is to outline the approach and methods that will be used for the qualification testing to determine the waste forms' overall attributes, initial and accelerated corrosion behavior, actual “service condition” behavior, and the expected long-term behavior. The latter may be based, in part, on “natural analog” comparisons. The ultimate goal of this testing is to provide appropriate information such that the effects of the ceramic waste form behavior and radionuclide release on the overall repository performance can be determined. Corrosion mechanisms and mechanisms of radionuclide loss will be inferred from the test results. Because a link may need to be established between long-term repository-relevant behavior and results of short-term consistency tests, the results will also be used to develop predictive behavior models and to evaluate the capability of the waste form to meet yet-to-be-determined radionuclide containment requirements for periods of 10,000 years or more. WBS 9.0 Repository Performance Assessment Modeling: A significant element in establishing the viability of electrometallurgical treatment technology is a defensible assessment which shows that the wastes to be generated from the process will perform acceptably when ultimately disposed in a geologic repository. The models for both ceramic and metal waste form continue to be developed. WBS 10.0 Environmental and Safety Support Tasks: These tasks provide the necessary safety analysis support for the electrometallurgical treatment demonstration activities. An Argonne readiness assessment for blanket operations was competed in April 1998. Treatment of Oxide Spent Fuels: Oxide spent fuels cannot be electrorefined directly but must first be converted to metals. Lithium metal is used as the reducing agent in molten LiCl to effect this conversion, and the Li2O that is produced dissolves in the molten salt. The salt and lithium metal are recycled by electrochemical deposition of lithium at a cathode and evolution of oxygen at an anode. Engineering-scale test, ES-8, designed to demonstrate lithium electrowinning on a large scale, is scheduled for completion in early May. Preparations for this experiment, including installation of the off-gas piping system and set-up of the experimental vessel and the anode, were completed. The off-gas system is designed to direct oxygen that is evolved from the anode during decomposition of Li2O, out of the glovebox to a HEPA-filtered vent. The experiment was started the last few days of April, and the electrowinning process was proceeding well. Results from the experiment will be reported after it is completed in May.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: SPRING 1998 STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY The rate of reduction of uranium oxide by lithium has been found to be a function of the lithium/salt contact area to salt volume ratio. Three laboratory-scale experiments have been completed in a series of experiments designed to investigate the effect of this ratio on the reduction rate. The results of these experiments show that the rate of reduction increases with increasing lithium/salt contact area up to a certain level, then no further increase in reduction rate is seen above that level. This optimum ration will be carefully defined and then used in future engineering-scale experiments. Treatment of Aluminum-Based Fuels: Demonstration of the feasibility of electro-metallurgical treatment of aluminum alloy spent fuels, such as foreign and domestic research reactor fuels, has been done in laboratory-scale experiments. The key step in treatment of this fuel is electrorefining of the aluminum, which represents about 90% of the spent fuel volume, and which can, after electrorefining, be discarded as low-level waste. Preparation of the engineering-scale aluminum electrorefiner and the electrolyte salt for beginning aluminum electrorefining was disrupted by a decision to use the same furnace well for testing the prototype Mark-V HTER. A small storage well in the J-118 glovebox is being converted to a furnace well for testing the engineering-scale aluminum electrorefiner.