Abstract: Uranic and transuranic metals and metal oxides are first dissolved in ozone compositions. The resulting solution in ozone can be further dissolved in ionic liquids to form a second solution. The metals in the second solution are then electrochemically deposited from the second solutions as room temperature ionic liquid (RTIL), tri-methyl-n-butyl ammonium n-bis(trifluoromethansulfonylimide) [Me3NnBu][TFSI] providing an alternative non-aqueous system for the extraction and reclamation of actinides from reprocessed fuel materials. Deposition of U metal is achieved using TFSI complexes of U(III) and U(IV) containing the anion common to the RTIL. TFSI complexes of uranium were produced to ensure solubility of the species in the ionic liquid. The methods provide a first measure of the thermodynamic properties of U metal deposition using Uranium complexes with different oxidation states from RTIL solution at room temperature.

Abstract: An in situ method for detecting alpha particles contained in a liquid medium, which uses a system which includes a counter-electrode and an alpha particle detector including a substrate made of an intrinsic semiconductor material sandwiched between two electrical contacts, wherein the contact intended to be in contact with the liquid medium is made of boron-doped diamond. By forming a particular electrolyte 8 and by causing a current to flow between counter-electrode and the boron-doped diamond contact in contact with the liquid medium, actinides or polonium present in the liquid medium may be concentrated on the boron-doped diamond contact, and by this means the detection limit of the alpha emitters may be lowered.

Abstract: A high capacity anode preparation apparatus is provided which allows for the processing of raw anodes at production rates of up to, or exceeding, 600 anodes per hour. The processed anodes are suitable for use in the electrorefining of various metal materials, but in particular, in the electrorefining of copper. The apparatus is preferably part of a system which utilizes high speed industrial robots to supply, and remove, anodes to or from the apparatus, and provides the anodes in a horizontal orientation. The apparatus is equipped with a variety of treatment stations which are adapted to treat the raw anode while it is held in a horizontal orientation. The horizontal orientation allows the center of gravity for the apparatus to be kept close to the center of gravity for the apparatus, and thus allows the apparatus to rotate more rapidly than prior art device. Faster processing of the raw anodes is provided.

Abstract: The present invention provides a method of simultaneously removing uranium and transuranics from metallic nuclear fuel in an electrorefiner. In the method, a potential difference is established between an anode basket containing the fuel and a solid cathode of the electrorefiner, thereby creating a diffusion layer of uranium and transuranic ions at the solid cathode, a first current density at the anode basket, and a second current density at the solid cathode. The ratio of anode basket area to solid cathode area is selected based on the total concentration of uranium and transuranic metals in a molten halide electrolyte in the electrorefiner and the effective thickness of the diffusion layer at the solid cathode, such that the established first and second current densities result in both codeposition of uranium and transuranic metals on the solid cathode and oxidation of the metallic nuclear fuel in the anode basket.

Abstract: A method for producing metal powder is provided the comprising supplying a molten bath containing a reducing agent, contacting a metal oxide with the molten bath for a time and at a temperature sufficient to reduce the metal in the metal oxide to elemental metal and produce free oxygen; and isolating the elemental metal from the molten bath.

Abstract: A spent nuclear fuel is reprocessed by dissolving a spent nuclear fuel in an aqueous nitric acid solution and separating and recovering nuclides contained in the resulting fuel solution by solvent extraction. A spent nuclear fuel reprocessing method includes: an electrolytic valence adjustment step in which nuclides contained in the fuel solution is electrolytically reduced without removing fission products or minor actinides until valence of plutonium is at a level at which solvent extraction efficiency is low by using the valence of plutonium contained in the fuel solution as a parameter; and a nuclide separation step in which, by using an extraction solvent which extracts uranium contained in the fuel solution, uranium is distributed from the fuel solution subjected to the electrolytic valence adjustment step to the extraction solvent.

Abstract: An electrolytic apparatus for an oxide electrolytic method includes an interior of an electrolytic vessel, a common cathode and two types of anodes different in shape and arrangement, a first electrolysis controller is connected between the cathode and the first anode, and a second electrolysis controller is connected between the cathode and the second anode. The electrolytic processing of the substance in the electrolytic vessel is carried out such that a pair of the cathode and one of the anodes is used for main electrolysis and a pair of the cathode and the other anode is used for auxiliary electrolysis. By this apparatus, prevention of the ununiform distribution of the electrodeposit, improvement of the processing speed and improvement of the durability of the crucible are achieved, whereby the recycling of spent nuclear fuels based on the nonaqueous reprocessing method is made feasible in a commercial scale.

Abstract: The invention provides a process for dissolving actinic oxides, the process comprising performing the steps of (a) introducing the actinic oxides into a solution of nitric acid; (b) treating the acidic solution in order to substantially remove palladium; and (c) treating with divalent silver. Preferably, the actinic oxides are comprised in spent nuclear fuel. Optionally, the process comprises a second treatment of the acidic solution in order to substantially remove palladium and a second treatment with divalent silver. The steps may be performed on a batchwise or continuous basis. The treatment to remove palladium is preferably carried out by solvent extraction or ion exchange, and provides greatly improved rates of dissolution of oxides of plutonium. The treatment with divalent silver preferably involves the addition of a source of monovalent silver, followed by an electrolysis treatment to generate divalent silver.

Abstract: A method for separating a metal from a composition including the metal involves forming an electrolytic cell in which the anode comprises a composition including the metal. The electrolyte is an ionic liquid. A sufficient potential difference is applied between the anode and the cathode to cause the metal to transfer from the anode to the cathode deposited thereon.

Abstract: This invention is provided for improvement of corrosion-resistant property of a crucible and for promotion of safety in a pyrochemical reprocessing method for the spent nuclear fuel. The spent nuclear fuel is dissolved in a molten salt placed in the crucible. In a pyrochemical reprocessing method, the nuclear fuel is deposited, and the crucible (2) is heated by induction heating. Cooling media (5, 6) are supplied to cool down, and a molten salt layer (7) is maintained by keeping balance between the heating and the cooling, and a solidified salt layer (8) is formed on inner wall surface of the crucible.

Abstract: A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

Abstract: Disclosed are a method of reducing spent oxides nuclear fuel to nuclear-fuel metal, in which metal oxides are reduced to metals using an electrochemical reduction device with LiCl—Li2O salt as an electrolyte, a cathode electrode assembly used in the method, and a reduction device including the cathode electrode assembly. The method is advantageous in that the process of reducing the spent oxide nuclear fuel to the nuclear-fuel metal and another process of recovering Li are united to simplify the whole processes, direct use of high oxidative Li metals is excluded to secure safety, and conversion efficiency of the spent oxide nuclear fuel is 99% or more.

Abstract: Purified plutonium and gallium are efficiently recovered from a solid plutonium-gallium (Pu—Ga) alloy by using an electrorefining process. The solid Pu—Ga alloy is the cell anode, preferably placed in a moving basket within the electrolyte. As the surface of the Pu—Ga anode is depleted in plutonium by the electrotransport of the plutonium to a cathode, the temperature of the electrolyte is sufficient to liquify the surface, preferably at about 500° C., resulting in a liquid anode layer substantially comprised of gallium. The gallium drips from the liquified surface and is collected below the anode within the electrochemical cell. The transported plutonium is collected on the cathode surface and is recovered.

Abstract: A spent reactor fuel processing method is provided for recovering at least any one of metallic nuclear fuel materials, in which the reactor fuel is composed by covering the metallic nuclear fuel material with a cladding tube made of alloy and having a melting point lower than that of the metallic nuclear fuel material and end plugs made of alloy are mounted to both ends thereof. The processing method comprises a cladding tube smelting separation process, a molten salt electrorefining process and a salt evaporation separation process for recovering metallic uranium, uranium and plutonium, or uranium, plutonium and transuranium elements.

Abstract: A method of treating a body of conducting material electrolytically which comprises the steps of placing the material to be treated in an aqueous oxidizing electrolyte, passing an electric current through the electrolyte and the material to be treated and sparging the electrolyte with gas. The material to be treated is placed in a basket comprising a conducting frame having an insulating container fitted therein, the basket being insertable in and removable from the electrolyte as desired. The insulating container is removable from the conducting frame, the frame having retaining means for receiving the insulating container. The container includes a base having perforations to allow electrolyte to contact the material to be treated. The electrolyte is sparged with gas in the region where the material to be treated is in contact with it, the gas being supplied by a plurality of pipes (21, 22) extending across the electrolyte at a location beneath the perforated base of the insulating container.

Abstract: Apparatus and methods for decontaminating surfaces are disclosed. A housing is configured with first and second channels and first and second fluid pathways in fluid communication therewith, respectively. First and second applicators are positioned within respective first and second channels and electrodes are electrically connected with the applicators. Electric current of a first polarity is supplied to a first applicator via the first electrode, and electric current of a second polarity is supplied to a second applicator via the second electrode. Decontaminating a surface comprises supplying a first fluid to a first applicator, supplying a second fluid to a second applicator, generating an electrical potential between the first and second applicators, and contacting the contaminated surface with the first and second applicators.

Abstract: Processes for in-situ decontamination and recovery of metal from radioactive-contaminated metal which is contained in process equipment, including ancillary systems of process equipment, comprise two basic steps. In the first step, an acid decontamination solution is circulated through the equipment and in contact with the radioactive-contaminated metal for removing the radioactive contaminants and a first surface portion of the metal from the metal-containing equipment. In the second step, an acid digestion solution is circulated through the equipment for removing at least a second portion of the metal which is substantially free of radioactive contaminants. The present methods are particularly suitable for in-situ decontamination and recovery of nickel from radioactive-contaminated nickel diffusion barriers in the cascade converters of uranium gas diffusion plants.

Abstract: A process and apparatus for dissolving a mixed oxide or mixture of oxides of uranium and plutonium. The powder is added with nitric acid to a chamber to dissolve uranium oxide, and the solution is circulated through a circuit of the apparatus with a portion of the solution passing through a filter. At least a portion of the filtered solution containing dissolved uranium oxide is removed from the apparatus, while returning non-filtered circulating solution containing non-dissolved plutonium oxide to the chamber. The removal of solution is then terminated, a monovalent silver salt is added and divalent silver is generated in-situ by electrolysis, the divalent silver causing dissolution of the plutonium oxide.

Abstract: An electrolytic cell, system and method for producing excess heat for use and for deactivation an actinide series metal by electrolysis in an aqueous media. The electrolytic cell includes a non-conductive housing having an inlet and an outlet and spaced apart first and second conductive grids positioned within the housing. A plurality of preferably cross linked polymer non-metallic cores each having an improved conductive exterior metallic surface formed of one or more of the actinides or daughter elements therefrom to form a bed of conductive beads are positioned within the housing in electrical contact with the first grid adjacent the inlet. The non-conductive cores are preferably sulfonated and then evaporation loaded with an actinide salt solution, dried and fired at decomposition temperature of the salt. An electric power source in the system is operably connected across the first and second grid whereby electrical current flows between the grids within the aqueous media flowing through the cell.

Abstract: A cathode-anode arrangement for use in an electrolytic cell is adapted for electrochemically refining spent nuclear fuel from a nuclear reactor and recovering purified uranium for further treatment and possible recycling as a fresh blanket or core fuel in a nuclear reactor. The arrangement includes a plurality of inner anodic dissolution baskets that are each attached to a respective support rod, are submerged in a molten lithium halide salt, and are rotationally displaced. An inner hollow cylindrical-shaped cathode is concentrically disposed about the inner anodic dissolution baskets. Concentrically disposed about the inner cathode in a spaced manner are a plurality of outer anodic dissolution baskets, while an outer hollow cylindrical-shaped is disposed about the outer anodic dissolution baskets. Uranium is transported from the anode baskets and deposited in a uniform cylindrical shape on the inner and outer cathode cylinders by rotating the anode baskets within the molten lithium halide salt.

Abstract: According to the invention, a consumable anode constituted by a metal alloy incorporates between 20 and 70 wt. % iron, between 20 and 40 wt. % cobalt and between 5 and 30 wt. % aluminium. To these basic constituents can optionally be added elements such as nickel and/or titanium and/or copper and/or niobium. The decontamination process involves an electrodissolution of said anode.

Abstract: A combination anode and cathode for an electrorefiner which includes a hollow cathode and an anode positioned inside the hollow cathode such that a portion of the anode is near the cathode. A retaining member is positioned at the bottom of the cathode. Mechanism is included for providing relative movement between the anode and the cathode during deposition of metal on the inside surface of the cathode during operation of the electrorefiner to refine spent nuclear fuel. A method is also disclosed which includes electrical power means selectively connectable to the anode and the hollow cathode for providing electrical power to the cell components, electrically transferring uranium values and plutonium values from the anode to the electrolyte, and electrolytically depositing substantially pure uranium on the hollow cathode. Uranium and plutonium are deposited at a liquid cathode together after the PuCl.sub.3 to UCl.sub.3 ratio is greater than 2:1.