C

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 progress reports approximately monthly. These “Monthly Highlights” of the Electrometallurgical Treatment Program have been submitted to the DOE, which has then provided them to the committee. In the preparation of this report, the committee used the Monthly Highlights for the periods mid-October through mid-November 1995, mid-November through December 1995, January 1996, and February 1996. These documents are included in this appendix.



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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY C 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 progress reports approximately monthly. These “Monthly Highlights” of the Electrometallurgical Treatment Program have been submitted to the DOE, which has then provided them to the committee. In the preparation of this report, the committee used the Monthly Highlights for the periods mid-October through mid-November 1995, mid-November through December 1995, January 1996, and February 1996. These documents are included in this appendix.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY ELECTROMETALLURGICAL TREATMENT PROGRAM MONTHLY HIGHLIGHTS FOR MID-OCTOBER THROUGH MID-NOVEMBER 1995 Status of Fuel Conditioning Facility (FCF): Approval was granted for the FCF safety basis and limited operations of EBR-II spent fuel handling and storage. This approval recognizes the resolution of all Operational Readiness Review (ORR) prestart findings related to the facility safety basis. However, spent fuel treatment or processing activities using irradiated spent fuel are not authorized at this time pending completion of a further National Environmental Policy Act (NEPA) process. The current plan is to prepare an environmental assessment (EA) for a limited electrometallurgical treatment R&D program at FCF. This limited demonstration program would include electrometallurgical treatment of up to 100 EBR-II drivers and 25 blanket assemblies (for high-throughput electrorefiner demonstration). This additional NEPA process will involve public review and comment including a public hearing. Detailed schedule is currently being developed by DOE. FCF Startup Tests with Depleted Uranium: Five uranium cathode deposits were made with the FCF electrorefiner under different operating conditions. The deposits ranged in size from 0.4 kg uranium to 6.0 kg uranium. Some minor modifications will be made to the electrorefiner as a result of information gained during these experiments. Two experiments were run with the FCF cathode processor. In the first experiment, 7 kg of uranium dendrites from the ANL-East engineering-scale electrorefiner were processed to assess the behavior of a high-zirconium product. In the second experiment, 6 kg of FCF electrorefiner dendrites were processed to gain information needed in optimizing the process parameters. Electrorefining Process Development: A high-throughput laboratory-scale electrorefiner, designed to test the high-throughput electrorefining of particles resulting from the reduction of spent oxide fuel, has been placed into operation and is performing better than expectations. Very high current density has been obtained at the low voltage levels (100 mA/cm² at 0.15 V) required for treatment of materials such as the TMI-2 core debris. A shipment of 36 unirradiated N-Reactor fuel elements (842.6 kg natural uranium) was received from Westinghouse Hanford Company during the month. These fuel elements will be used in experiments with the 25-in. diameter high-throughput electrorefiner; the quantity provided by Westinghouse will permit about five batches to be tested with this electrorefiner. Design of major components for the 25-in. high-throughput electrorefiner has been completed and fabrication is in progress. Treatment of Oxide Spent Fuel: The anode for the lithium electrowinning step of the lithium reduction process has been redesigned and fabrication has been completed; the anode is ready for transfer into the glovebox for final assembly. This anode incorporates major design revisions, including the use of a cylindrical-shaped anode, which will have a more uniform current density than the finned anode used previously. This will prevent the formation of the extremely fine oxygen bubbles observed with the first anode. The new anode also provides for a convection-driven flow of salt and oxygen from the anode toward the inside of the inner anode shroud, to aid in the removal of oxygen from the system. Lithium-doped nickel oxide anodes have been tested for application in this process. These anodes were found to perform as well as platinum with very minimal corrosion. It is expected that this relatively cheap and easily prepared anode will eventually replace the platinum anode currently being used for regeneration of metallic lithium. Treatment of MSRE Fuel and Flush Salts: Tests of the bismuth pounder cathode in the chloride electrolyte system have been completed successfully. The pounder concept works exceptionally well with the bismuth cathode. The remaining heavy metals in the laboratory-scale electrorefiner have been oxidized into the chloride salt phase in preparation for removing the salt and the crucible from the

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY electrorefiner. The vessel will be replaced with a vessel suitable for operation with simulated MSRE fuel salt, consisting of a mixture of Zr-Be-Li-U fluorides, which has already been prepared. Mineral Waste Form Development: A zeolite ion exchange column has been operated with simulated waste salt. This was the initial test to determine the operating conditions for an ion exchange column and to establish the feasibility of the process. Approximately 9 kg of waste-salt was loaded into the feed tank and 2.5 kg salt was passed through a 1.25-in. diameter column packed with preloaded (infused with LiCl-KCl salt to drive out residual water and gas) zeolite pellets. Samples of the effluent salt were collected at regular intervals and are now being analyzed. A second run of the zeolite ion exchange column has also been completed, using salt containing a typical distribution of fission product elements. Analytical results showed that the concentration of the fission product elements in the first run increased regularly with time during the run, in good agreement with the predictions of the dynamic model of column performance that is under development. Several successful hot isostatic press (HIP) runs were completed during the past month; all resulted in uniform, highly dense composites. Each run was done with a 1-in. diameter × 1-in. high 304 stainlesssteel can filled with glass frit and blended zeolite powders. Careful powder blending and can loading procedures appear to have been key to the success of these runs. Hot isostatic pressing of glass-bonded zeolite waste form samples has now been extended to 3-in. long cans. The samples in the larger cans appear to be equally dense and uniform. A hot uniaxial press for production of Pu-bearing mineral waste form samples has been installed for testing in the zeolite glovebox. After testing without plutonium additions, the press will be transferred to a glovebox in a plutonium laboratory for testing with plutonium; a range of waste form compositions will be tested to determine optimum Pu concentrations in the waste form. Immobilization of Surplus Plutonium: Samples of zeolite were equilibrated with molten LiCl-KCl salt that contained 19 wt. % plutonium as PuCl3 and 4.7 wt. % cesium as CsCl to determine the plutonium loading that could be achieved. Analyses of the samples indicated a plutonium content of 27.5 wt. % and a cesium content of 4.8 wt. % in the zeolite. This plutonium loading is sufficient to favor the use of glass-bonded zeolite for plutonium immobilization. A hot uniaxial press for production of Pu-bearing mineral waste form samples has been installed for testing in the zeolite glovebox. After testing without plutonium additions, the press will be transferred to a glovebox in a plutonium laboratory for testing with plutonium; a range of waste form compositions will be tested to determine optimum Pu concentrations in the waste form.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY ELECTROMETALLURGICAL TREATMENT PROGRAM MONTHLY HIGHLIGHTS FOR MID-NOVEMBER THROUGH DECEMBER 1995 Status of Fuel Conditioning Facility (FCF): The draft environmental assessment for a limited electrometallurgical treatment research and demonstration at FCF is being reviewed by DOE, and is expected to be available for public review and comment on January 16, 1996. Public meetings are also scheduled in Idaho Falls on February 8, and in Washington, DC on February 13. Preparations were completed for receiving the first irradiated fuel assembly which will be stored in the FCF's air cell. Construction of the pneumatic transfer lines between FCF and HFEF and the analytical lab were completed. This should help expedite sample transfers for analytical work. FCF Startup Tests with Depleted Uranium: Minor modifications were made and preventive maintenance was performed on the electrorefiner. An additional 5 kg of uranium were transferred into the argon cell. This uranium will be used in a series of five tests which are designed to better understand the solid cathode operating parameters. The hot isostatic press (HIP) for mineral waste forms was received from the vendor. The HIP is being installed out-of-cell for some initial testing of large-scale waste forms with nonradioactive materials. Electrorefining Process Development: Baskets made of 6- to 8-mesh steel screen were tested as inserts for the prismatic anodic dissolution baskets in the engineering-scale electrorefiner. The inserts were easily removed from the anodic dissolution baskets at room temperature after contact with the salt in the electrorefiner. This result shows promise for the development of removable anode screen baskets that can be used to retain the cladding hulls and noble metal fission products. Similar tests are planned with inserts for the curved baskets that will be used in the high-throughput electrorefiner. The laboratory-scale high-throughput electrorefiner, designed to treat the granular product from lithium reduction of simulated oxide spent fuel, is provided with the capability to compact the uranium product scraped off the cathode. In the first operation of this compactor, the uranium deposit was found to have a density of 11 g/cm³. This is a factor of three greater than cathode deposits formed on the conventional solid mandrel cathodes and means that: (1) the electrorefiner volume can be reduced, (2) uranium deposit handling is easier for a given batch size, and (3) the throughput rate of the cathode processor can be increased substantially. Chemical analysis results have been obtained for the electrorefining test in which chopped, clad U-Zr-Fs (Fs represents noble metal fission products) was loaded into anodic dissolution baskets having fine mesh steel retainer screens with 200-, 250-, and 325-mesh screens (and one basket with no retainer screen). The results show that 92% of the noble metal fission products were retained in the cladding hulls without the use of a fine mesh retainer screen. Only slightly better retention was obtained with use of the screens. Therefore, the use of retainer screens is not necessary for satisfactory concentration and removal of the noble metal fission products. Treatment of Oxide Spent Fuels: Lithium-doped nickel anodes have been tested for application in the electrochemical decomposition of lithium oxide dissolved in lithium chloride, as part of the process for treatment of oxide spent fuels. These anodes were found to perform as well as platinum, with very minimal corrosion. It is expected that this relatively cheap and easily prepared anode will eventually replace the platinum anode currently being used for regeneration of metallic lithium. Examination of the results of the most recent engineering-scale lithium oxide electrowinning experiment indicates that the design changes made since the first experiment have corrected the problems with oxygen venting. It appears that the system was behaving as expected and was producing metallic lithium for about the first 3 h of the experiment. At that point, the MgO insulator at the bottom of the vessel failed and an effective cathode was formed directly below the anode. The lithium formed at the new

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY cathode location contacted the platinum anode and destroyed it. This problem can be remedied by a redesign of the insulator; such redesign is now in progress. Treatment of MSRE Fuel and Flush Salts: The original laboratory-scale electrorefiner has been removed from Glovebox No. 1 in the G-118 plutonium laboratory and replaced with a Hastelloy-424 crucible for use in the demonstration of MSRE fuel salt treatment. The fluoride salts required to make up the simulated MSRE fuel salt have been loaded into the laboratory-scale electrorefiner glovebox and pretreated to remove trace quantities of water. The next step will be the in situ production of the uranium fluoride fuel constituent. The metal-salt contactor apparatus for experiments with the nitride-enhanced recovery of fission product strontium and yttrium from the MSRE salt has been fabricated and loaded into the plutonium glovebox. Preliminary tests involving the removal of zirconium from molten bismuth via nitride formation were successful. The next series of tests will be metal phase/nitride phase tests, followed by salt phase/metal phase/nitride phase tests. Mineral Waste Form Development: The initial series of zeolite/molten salt ion exchange column tests has yielded valuable information on the effect of temperature on column behavior. Much more rapid exchange is observed at 525°C than at 480°C, based on color changes in the molten salt. The temperature effect is consistent with the slow exchange of +3 ions observed in batch tests and the expected strong temperature sensitivity of the exchange reactions. Two different thermal profiles have been investigated as part of the effort to optimize the hot isostatic pressing operation for production of the mineral waste form. One involved a 4 h hold at 750°C (vs. 1 or 2 h previously used), and the other examined a higher holding temperature (800°), chosen to enhance solid state densification kinetics. Both procedures produced sound samples; destructive examination of the composites is in progress. A number of mineral waste form samples have now been hot isostatically pressed in the new, larger 1-in. diameter × 3-in. long HIP cans, using the same time temperature/pressure profiles used with the 1-in. long samples. Sample preparation appears to have been completely successful. The first sample of the mineral waste form to be prepared by hot isostatic pressing was leach tested for seven days at 900 °C in deionized water. The mass loss was 0.4%, comparable to the values obtained in identical tests with uniaxially-pressed glass-bonded zeolite. The code for predicting behavior of the zeolite column has been improved to provide for a variation of exchange factors with zeolite loading, to account for the variation of activity coefficients of the fission products in the zeolite. For the alkaline earth elements Ba and Sr, the exchange factor depends on both the zeolite loading of the alkaline earth elements and on the loading of lanthanide fission products. Lanthanide-element and cesium exchange factors depend only on the loading of these elements in the zeolite. The alkali metals sodium and potassium have invariant exchange factors. Increasing the capacity of the zeolite blender from 100 to 300 g was found to yield a great reduction in the amount of free salt in the blended zeolite at salt loadings between 7.5 and 9.5 chloride ions per unit cell. A small increase in the amount of free salt (to 0.15 wt %) was observed at 9.5 Cl− per unit cell, and the amount of free salt at 12 Cl− per unit cell was unacceptably high (2 wt % of total solids, 10% of total salt). These results serve to define the preferred operating range for zeolite loading at less than 9.5 chloride ions per unit cell.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY Waste Treatment Process Development: Tests of the four-stage pyrocontactor with cadmium have shown excellent rotor stability up to the motor limit of 3300 rpm. The feed system for the pyrocontactors has been changed from gravity flow to pressure-controlled flow, resulting in an improvement in the stability of metal flared by a factor of six.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY ELECTROMETALLURGICAL TREATMENT PROGRAM MONTHLY HIGHLIGHTS FOR JANUARY 1996 Status of Fuel Conditioning Facility (FCF): The draft environmental assessment (EA) for a limited electrometallurgical treatment research, development, and demonstration at FCF received intense review by DOE. The many comments that resulted were evaluated, discussed and resolved. The draft EA is now scheduled to be available for public review and comment on February 5. Public meetings in Idaho Falls and Washington, DC are now scheduled on February 21 and February 27, respectively. The first irradiated fuel assembly was transferred into the FCF air cell for interim storage in early January. FCF Startup Tests with Depleted Uranium: A cathode processor experiment was completed as part of the effort to determine an optimum time-temperature-pressure operating range that minimizes chemical interaction between the crucible or crucible coating material and the cathode product. The FCF electrorefiner dendrites from Cathodes 10 and 11 (6.2 kg) were processed. Electrorefiner cathode runs resumed after an FCF maintenance and modification outage. Casting furnace consolidation and sampling experiment was successfully conducted involving a 5.6 kg ingot. Electrorefining Process Development: Chemical analysis results have been obtained for the electrorefining test in which chopped, clad U-Zr-Fs fuel was loaded into anodic dissolution baskets having fine mesh steel retainer screens with 200-, 250-, and 325-mesh screens (and one basket with no retainer screen). The results show that 92% of the noble metal fission products are retained in the cladding hulls with no fine mesh retainer screen. Only slightly better retention was obtained with use of the screens. Therefore, the use of retainer screens is not necessary. Treatment of Metal Fuels: An electrorefining test of a simulated Hanford single-pass reactor fuel element, consisting of a uranium slug clad with aluminum and bonded to the cladding with an Al-Si alloy, has been completed. The uranium slug was exposed to the electrolyte salt by slitting the cladding longitudinally in four 5/16-in. strips spaced at 90° intervals around the element. Approximately 70% of the uranium slug was fully exposed in the area stripped, with some Al-Si bond remaining in the balance of the stripped area. The operating current density in the test was initially 0.07 A/cm², based on the uranium surface area exposed to the electrolyte. The cutoff voltage for the test was 0.1 V, based on previous laboratory-scale tests that showed the aluminum cladding was not dissolved at voltages less than 0.15 V. Complete analytical chemistry results for the test should be available by the time of the next report. Treatment of Oxide Spent Fuels: A significant portion of the TMI-2 core debris consists of a mixture of (U, Zr)O2 phases. This ceramic material is very similar to the “corium” material produced in the thermite reaction of uranium, zirconium, Fe2O3 and CrO3 and used in a separate LWR safety program on studies of the interactions between molten core material and concrete. A lithium reduction experiment has been conducted using corium to determine how this material will behave in the lithium reduction process for the treatment of DOE spent oxide fuel such as the TMI-2 core debris. Preliminary results of the test indicate that it was successful; chemical analysis results are pending. An experiment using a liquid bismuth cathode for the regeneration of metallic lithium from the oxide reduction process was performed successfully. A lithium ingot was produced with 88% coulombic efficiency. This process may prove to be the solution to the difficult problem of lithium recovery. Treatment of MSRE Salt: Two salt compositions have been prepared for use in experiments on MSRE fuel salt treatment. The first is a LiF-BeF2 salt, and the second is a LiF-BeF2-ZrF4 salt. Both AlN and Hastelloy-242 crucibles were used in these preparations; neither showed signs of attack by the salts. Six

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY batches of LiF-BeF2-ZrF4 matrix salt, 1 kg each, have now been prepared. The salt ingots were easily removed from the Hastelloy-242 crucible in which they were prepared. The UF4 constituent of the simulated MSRE fuel salt will be prepared in situ using these salts. A nitride bed in a liquid metal cathode is being considered for use in the removal of certain radionuclides from the electrolyte salt used in the treatment of MSRE fuel salt and aluminum-based metal fuels. An experiment was performed to determine if there is nitrogen exchange between bismuth and barium nitride: the results showed that these materials are not compatible and would not be suitable in the proposed application. A second experiment is being prepared to evaluate the compatibility between molten zinc and lithium nitride. Waste Treatment Process Development: Tests of the four-stage pyrocontactor with cadmium have shown excellent rotor stability up to the motor limit of 3300 rpm. The feed system for the pyrocontactors has been changed from gravity flow to pressure-controlled flow, resulting in an improvement m the stability of metal flow rate by a factor of six. In an experiment with the zeolite column apparatus, approximately 3 kg of waste salt were passed through the column at a flow rate of about 50 g/min. The untreated salt contained rare earth fission product elements and had a distinct blue color due to the presence of the rare earth cations. The effluent salt from the column was pure white, even after the full loading of salt had been transferred through the column, indicating that fission product removal was very effective. Mineral Waste Form Development: The 12 hot isostatic press (HIP) samples of the mineral waste form produced during the last week of operations in December have been analyzed to determine leach resistance, bulk density and final crystalline morphology. All samples except one exhibited varying amounts of conversion from zeolite to sodalite, as measured by X-ray diffraction. This conversion is catalyzed by the glass constituent, and the degree of conversion varies with glass composition. The conversion of sodalite is not unfavorable, because the sodalite may be somewhat more leach resistant than the glass-bonded zeolite structure. The three-day leach tests used to evaluate leach resistance of these samples did not discriminate among degrees of conversion to sodalite. Three mineral waste form samples were prepared by hot isostatic pressing at 600°C for 1 h. The three samples contain 1:1 mixtures of blended zeolite (approximately 7.5 chloride ions per unit cell) and one of three different glass compositions. The three samples showed uniform densification upon removal from the press. Core samples were taken for X-ray diffraction analysis of the crystalline phases present. If it proves that the zeolite structure was retained, processing time will be increased in order to enhance the bulk density of the monoliths. Samples of blended zeolite containing 4 chloride ions per unit cell were heated at 700°C for varying periods of time and their structure compared with samples containing 7.5 and 9.5 chloride ions per unit cell. With 4 Cl− per unit cell, the sample was found to contain sodalite, nepheline and a small amount of zeolite. The 7.5 Cl− per unit cell sample contained weak nepheline X-ray spectra, but the 9 Cl− per unit cell sample did not. This supports the idea that thermal stability of the zeolite increases with salt loading. Treatment of Aluminum-based Fuels: A presentation on the proposed electrometallurgical technique for treatment of aluminum-clad spent fuel was given to the DOE Research Reactor SNF Task Team on January 17. Two options were described, one in which the uranium is recovered, denatured and made available for recycle, with the aluminum removed for disposal as a low-level waste. In the second option, the bulk of the aluminum (> 80%) is removed by electrotransport and the remaining aluminum and uranium, together with some fission products, are recovered and sent to a glass melter.

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ELECTROMETALLURGICAL TECHNIQUES FOR DOE SPENT FUEL TREATMENT: A STATUS REPORT ON ARGONNE NATIONAL LABORATORY'S R&D ACTIVITY ELECTROMETALLURGICAL TREATMENT PROGRAM MONTHLY HIGHLIGHTS FOR FEBRUARY 1996 Status of Fuel Conditioning Facility (FCF): Public hearings on the draft environmental assessment were held in Idaho Falls and Washington, DC. The public comment period will continue until March 22. A comment response document will be prepared and should be available in mid-April. Two additional irradiated fuel assemblies were transferred into the FCF air cell during February. FCF Startup Tests with Depleted Uranium: A different zirconia crucible coating material (ZO modified) was tested during a cathode processor experiment at lower temperature and pressure. The initial observations from the January and this experiment indicate that the lower temperature operations will minimize the uranium ingot and crucible coating interaction. Six different cathodes were produced in the electrorefiner. Sufficient electrorefiner experience has been obtained to start depleted uranium-zirconium experiments. The hot isostatic press for the ceramic waste form was installed in the engineering laboratory and its acceptance testing was completed. Three ceramic waste form samples were produced and the samples appeared identical to ceramic waste forms that are produced in the Chemical Technology Division's hot isostatic press. These activities meet the DOE milestone to initiate ceramic waste form experiments at Argonne-West. Electrorefining Process Development: Uranium electrodeposition tests were initiated in the engineering-scale electrorefiner at ANL-E to support the Fuel Conditioning Facility electrorefiner operations. The objective of these tests is to perform parametric studies of salt and cadmium mixing, operating temperature, electrodeposition current density, solid mandrel cathode rotation speed, cutoff voltage, and cathode conditioning, in order to determine their effects on the weight of uranium collected at the solid cathode and the morphology of the uranium deposits. Treatment of MSRE Salt: Preparation of the LiF-BeF2-ZrF4 matrix salt for MSRE fuel salt electrorefining experiments has been completed. A total of eight 1-kg ingots has been prepared. Two separate batches of uranium fluoride have been prepared by the oxidation of uranium with bismuth fluoride, and will be added to the matrix salt. Waste-Treatment Process Development: A new, high-capacity double-cone blender has been placed into service for use in preparation of zeolites for hot-pressing. The smoother walls and lack of baffles in this blender reduce the tendency of the powders to agglomerate. The blender was tested successfully with 500 g of material at a salt:zeolite ratio of 21:79 and a test temperature of 500°C; the powders remained free-flowing. Treatment of Aluminum-Based Fuels: An electrotransport process for the treatment of aluminum-matrix fuel has been developed. In this process, the aluminum fuel is melted and then silicon is added to melt to form silicides of the actinides and rare earths. The molten metal is then cast into ingots which fit into the anodic dissolution baskets of the standard high-throughput electrorefiner. The aluminum is then deposited on the tubular cathodes by electrotransport in a fluoride electrolyte, leaving the uranium in the anode basket. The uranium ingots remaining in the anode baskets can then be incorporated in a glass waste form, or recovered by conventional electrorefining in a chloride electrolyte. Thereafter, the uranium can be blended to low enrichment levels for storage or recycle. This process has the advantage of operation with a proven electrorefiner design with significant throughput capacity. An important potential application of the process is in the treatment of foreign research reactor spent fuel, wherein the high-level waste volumes generated would be approximately two orders of magnitude less than that arising from direct disposal of the spent fuel (i.e., a factor of 100 reduction in waste volume).