The United States of America As Represented by the United States

The United States of America As Represented by the Department of Energy

The University of Chicago

Uchicago Argonne, LLC

John P Ackerman

Terry R Johnson

William E Miller

James E Battles

R Dean Pierce

Dennis W Dees

James L Willit

Mark A Williamson

Zygmunt Tomczuk

John E Young, Jr

Kimio Kinoshita

Stanley A Zwick

Donald F Fischer

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO, is added to a solution of UCldissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

Uranium dioxide electrolysis

A method of electrochemically reducing a metal oxide to the metal in an electrochemical cell is disclosed along with the cell. Each of the anode and cathode operate at their respective maximum reaction rates. An electrolyte and an anode at which oxygen can be evolved, and a cathode including a metal oxide to be reduced are included as is a third electrode with independent power supplies connecting the anode and the third electrode and the cathode and the third electrode.

Three-electrode metal oxide reduction cell

The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

Process to remove rare earth from IFR electrolyte

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The Cu--Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg Cl.sub.2 to transfer Mg values from the transport salt to the Cu--Mg alloy while transuranium actinide and rare earth fission product metals transfer from the Cu--Mg alloy to the transport salt. Then the transport salt is mixed with a Mg--Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg--Zn alloy.

Salt transport extraction of transuranium elements from lwr fuel

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.

Magnesium transport extraction of transuranium elements from LWR fuel

A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

Uranium chloride extraction of transuranium elements from LWR fuel

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

Plutonium recovery from spent reactor fuel by uranium displacement

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

Electrorefining process and apparatus for recovery of uranium and a

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

Process to separate transuranic elements from nuclear waste

A method is disclosed for forming a core for use in a solid oxide fuel cell that electrochemically combines fuel and oxidant for generating galvanic output. The core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support consisting instead only of the active anode, cathode, electrolyte and interconnect materials. Each electrolyte wall consists of cathode and anode materials sandwiching electrolyte material therebetween, and each interconnect wall consists of the cathode and anode materials sandwiching interconnect material therebetween. The electrolyte and interconnect walls define a plurality of substantially parallel core passageways alternately having respectively the inside faces thereof with only the anode material or with only the cathode material exposed. In the wall structure, the electrolyte and interconnect materials are only 0.002-0.01 cm thick; and the cathode and anode materials are only 0.002-0.05 cm thick. The method consists of building up the electrolyte and interconnect walls by depositing each material on individually and endwise of the wall itself, where each material deposit is sequentially applied for one cycle; and where the depositing cycle is repeated many times until the material buildup is sufficient to formulate the core. The core is heat cured to become dimensionally and structurally stable.

Method of fabricating a monolithic core for a solid oxide fuel cell

A solid oxide fuel cell for electrochemically combining fuel and oxidant for generating galvanic output, wherein the cell core has an array of electrolyte and interconnect walls that are substantially devoid of any composite inert materials for support. Instead, the core is monolithic, where each electrolyte wall consists of thin layers of cathode and anode materials sandwiching a thin layer of electrolyte material therebetween, and each interconnect wall consists of thin layers of the cathode and anode materials sandwiching a thin layer of interconnect material therebetween. The electrolyte walls are arranged and backfolded between adjacent interconnect walls operable to define a plurality of core passageways alternately arranged where the inside faces thereof have only the anode material or only the cathode material exposed. Means direct the fuel to the anode-exposed core passageways and means direct the oxidant to the cathode-exposed core passageway; and means also direct the galvanic output to an exterior circuit. Each layer of the electrolyte and interconnect materials is of the order of 0.002-0.01 cm thick; and each layer of the cathode and anode materials is of the order of 0.002-0.05 cm thick.

Solid oxide fuel cell having monolithic core

An electrolyte compact for fuel cells includes a particulate support material of lithium aluminate that contains a mixture of alkali metal compounds, such as carbonates or hydroxides, as the active electrolyte material. The porous lithium aluminate support structure is formed by mixing alumina particles with a solution of lithium hydroxide and another alkali metal hydroxide, evaporating the solvent from the solution and heating to a temperature sufficient to react the lithium hydroxide with alumina to form lithium aluminate. Carbonates are formed by reacting the alkali metal hydroxides with carbon dioxide gas in an exothermic reaction which may proceed simultaneously with the formation with the lithium aluminate. The mixture of lithium aluminate and alkali metal in an electrolyte active material is pressed or otherwise processed to form the electrolyte structure for assembly into a fuel cell.

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Downers Grove, IL, United States of America

John P Ackerman