Abstract: Methods of producing plutonium-238 are disclosed. One method includes dissolving neptunium-237 in a nitric acid solution to produce a neptunium target solution, subjecting the neptunium target solution to neutron radiation to produce plutonium-238, and removing the plutonium-238 from the neptunium target solution. A second method includes exposing a solution of neptunium-237 to neutron radiation to produce plutonium-238, complexing the plutonium-238 with an organophosphorus compound, and separating the plutonium-238/organophosphorus compound complex from the solution of neptunium-237. A third method includes dissolving neptunium-237 to form a neptunium-237 target solution, exposing the neptunium-237 to thermal neutrons to produce plutonium-238, utilizing an organophosphorus compound to complex the plutonium-238 and the organophosphorus compound, extracting the plutonium-238/organophosphorus compound complex from the irradiated neptunium target solution, and recovering the plutonium-238.

Abstract: Embodiments of a method and a system for recovering a metal, such as uranium, from a metal-containing material are disclosed. The metal-containing material is exposed to an extractant containing a liquid or supercritical-fluid solvent and an acid-base complex including an oxidizing agent and a complexing agent. Batches of the metal-containing material are moved through a series of stations while the extractant is moved through the stations in the opposite direction. After the extraction step, the metal is separated from the solvent, the complexing agent and/or other metals by exposing the extract to a stripping agent in a countercurrent stripping column. The complexing agent and the solvent exit the column and are separated from each other by reducing the pressure. The recovered complexing agent is recharged with fresh oxidizing agent and recombined with fresh or recovered solvent to form a recovered extractant, which is distributed through the extraction stations.

Abstract: A method for separating and recovering trivalent and tetravalent actinoids in a simple and less costly manner without using an organophosphorus compound is provided. This method selectively separates and recovers the tetravalent actinoid plutonium Pu (IV) and the trivalent actinoids americium Am (III) and curium Cm (III) from trivalent lanthanoids Ln (III), etc. with the use of an extractant having a functional group with neutral multidentate ligand activity which is a hybrid donor type organic compound having both of donor atoms, i.e., an oxygen atom and a nitrogen atom.

Abstract: Embodiments of a method and a system for recovering a metal, such as uranium, from a metal-containing material are disclosed. The metal-containing material is exposed to an extractant containing a liquid or supercritical-fluid solvent and an acid-base complex including an oxidizing agent and a complexing agent. Batches of the metal-containing material are moved through a series of stations while the extractant is moved through the stations in the opposite direction. After the extraction step, the metal is separated from the solvent, the complexing agent and/or other metals by exposing the extract to a stripping agent in a countercurrent stripping column. The complexing agent and the solvent exit the column and are separated from each other by reducing the pressure. The recovered complexing agent is recharged with fresh oxidizing agent and recombined with fresh or recovered solvent to form a recovered extractant, which is distributed through the extraction stations.

Abstract: A treatment apparatus (10) includes a liquid reactant metal containment vessel (11) for containing a first liquid reactant metal and isolating the reactant metal from the atmosphere. A release chamber (14) is adapted to receive the first liquid reactant metal from the containment vessel (11) and a submerging arrangement (21) is adapted to dunk or submerge a container (46) of feed material into the liquid reactant metal and move the container to a release location within the release chamber (14). Relatively light materials rising from the submerged container (46), including unreacted feed material, intermediate reaction products, and perhaps final reaction products collect in a collection area (60) having an upper surface defined by an upper surface of the release chamber (14).

Abstract: A method for producing ultra-pure Pu-238 is provided. The method comprises the steps of short-term irradiating Am-241 targets with a high, thermal neutron flux greater than 6.5×1014 neutrons cm?2 s?1 and more preferably greater than 1×1015 neutrons cm?2 s?1 for a predetermined period of time preferably from 20 days to 30 days and more preferably approximately 25 days to convert a substantial fraction of the Am-241 to Cm-242, e.g., approximately 0.555 g of Cm-242 per g of Am-241 charged, thereafter promptly chemically separating the produced Cm-242, preferably within 10 days to 20 days of the irradiation cycle of the Am-241 targets and more preferably within about 15 days of the irradiation cycle of the Am-241 targets, and recovering the ultra-pure Pu-238 decay product of the separated Cm-242.

Abstract: A treatment apparatus (10) includes a liquid reactant metal containment vessel (11) for containing a first liquid reactant metal and isolating the reactant metal from the atmosphere. A release chamber (14) is adapted to receive the first liquid reactant metal from the containment vessel (11) and a submerging arrangement (21) is adapted to dunk or submerge a container (46) of feed material into the liquid reactant metal and move the container to a release location within the release chamber (14). Relatively light materials rising from the submerged container (46), including unreacted feed material, intermediate reaction products, and perhaps final reaction products collect in a collection area (60) having an upper surface defined by an upper surface of the release chamber (14).

Abstract: The present invention relates to a process to dissolve plutonium or a plutonium alloy, by placing it in contact with an aqueous dissolution mixture, wherein said dissolution mixture comprises nitric acid, a carboxylic acid with complexing properties with respect to plutonium, and a compound comprising at least one —NH2 radical such as urea. The invention also relates to a process to convert plutonium or a plutonium alloy into plutonium oxide and to manufacture nuclear fuel from said oxide. The invention particularly applies to the dismantling of plutonium contained in nuclear weapons with a view to its use in civilian nuclear reactors, particularly in the form of MOX fuel.

Abstract: A method is described for the chemical reduction of plutonium oxides to plutonium metal by the use of pure lithium metal. Lithium metal is used to reduce plutonium oxide to alpha plutonium metal (alpha-Pu). The lithium oxide by-product is reclaimed by sublimation and converted to the chloride salt, and after electrolysis, is removed as lithium metal. Zinc may be used as a solvent metal to improve thermodynamics of the reduction reaction at lower temperatures. Lithium metal reduction enables plutonium oxide reduction without the production of huge quantities of CaO--CaCl.sub.2 residues normally produced in conventional direct oxide reduction processes.

Abstract: 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.