. "B ANL Monthly Highlights of the Electrometallurgical Treatment Program." Electrometallurgical Techniques for DOE Spent Fuel Treatment: Spring 1998 Status Report on Argonne National Laboratory's R & D Activity. Washington, DC: The National Academies Press, 1998.
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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