A method of processing a nuclear material for use as a nuclear fuel in a nuclear reactor is disclosed herein. The nuclear material includes a complex isotope vector including a plurality of isotopes including a targeted isotope and a non-targeted isotope. The method can include: determining a wavelength of electromagnetic radiation based, at least in part, on the targeted isotope; emitting a beam of electromagnetic radiation including the determined wavelength towards the nuclear material; separating, via the emitted beam of electromagnetic radiation, the nuclear material into a first stream and a second stream; enriching, via the emitted beam of electromagnetic radiation, a concentration of the targeted isotope to a predetermined concentration; and dispositioning, via a sensitivity to the determined wavelength, the enriched concentration of the targeted isotope to the first stream of the nuclear material; and dispositioning, via a lack of sensitivity to the determined wavelength, the non-targeted isotope to the second stream of the nuclear material.

Disclosed are anodes for an electrochemical reduction system, such as for the electrochemical reduction of oxides in systems using molten salt electrolytes. The anodes comprise a rod or plate formed of and include at least one alloy of at least one transition metal and at least one platinum group metal. The alloy anodes may be less expensive than anodes formed solely from platinum group metals and may exhibit less material attrition than anodes formed solely from transition metals. Related methods and electrochemical reduction systems are also disclosed.

A method for reprocessing nitride spent nuclear fuel in molten salts comprises chlorinating the fuel in a melt of a mixture of alkali and/or alkaline earth metal chlorides containing cadmium dichloride. The chlorination is carried out in an apparatus for reprocessing nitride spent nuclear fuel using an inert gas atmosphere The apparatus has a heated zone containing a reactor with molten chlorides and nitride spent nuclear fuel submerged therein, and also a cold zone arranged under the reactor. In the chlorination process, the zone of the apparatus containing the reactor is heated to a temperature greater than 700 C., the nitride spent nuclear fuel is kept in the melt until fully chlorinated. The cold zone of the apparatus is used for crystallizing metallic cadmium which forms during the chlorination.

A method for reprocessing nitride spent nuclear fuel in molten salts comprises chlorinating the fuel in a melt of a mixture of alkali and/or alkaline earth metal chlorides containing cadmium dichloride. The chlorination is carried out in an apparatus for reprocessing nitride spent nuclear fuel using an inert gas atmosphere The apparatus has a heated zone containing a reactor with molten chlorides and nitride spent nuclear fuel submerged therein, and also a cold zone arranged under the reactor. In the chlorination process, the zone of the apparatus containing the reactor is heated to a temperature greater than 700 C., the nitride spent nuclear fuel is kept in the melt until fully chlorinated. The cold zone of the apparatus is used for crystallizing metallic cadmium which forms during the chlorination.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

A material (e.g., an alloy) comprises molybdenum, rhenium, and at least one element selected from the group consisting of tellurium, iodine, selenium, chromium, nickel, copper, titanium, zirconium, tungsten, vanadium, and niobium. Methods of forming the material (e.g., the alloy) comprise mixing molybdenum powder, rhenium powder, and a powder comprising at least one element selected from the group consisting of tellurium, iodine, selenium, chromium, nickel, copper, titanium, zirconium, tungsten, vanadium, and niobium. The mixed powders may be coalesced to form the material (e.g., the alloy).

An in-line dissolved gas removal membrane-based apparatus for removing dissolved hydrogen and fission gases from the letdown stream from a reactor coolant system.

A process for recovering uranium from components contaminated with uranium oxide includes providing a cleaning apparatus with a cleaning solution for dissolving the uranium oxide of the components, carrying out a cleaning process by introducing a batch of components into the cleaning apparatus, and carrying out a measurement for determining the uranium content of the components. The cleaning and the measuring are repeated if a limit value for the uranium content is exceeded. The components are discharged from the process if the uranium content falls below a limit value. The cleaning is carried out on a plurality of successive batches of components until a control measurement indicates an unsatisfactory cleaning action of the cleaning solution. The uranium oxide dissolved in the cleaning solution is recovered after indication of the unsatisfactory cleaning action.

A method for stripping U(VI) and an An(IV) from an organic solution including tri-n-butyl phosphate in an organic diluent, the solution containing U(VI) and the An(IV) present as U(VI) nitrate and An(IV) nitrate at concentrations such that the U(VI) nitrate concentration is higher than the An(IV) nitrate concentration, and the sum of the U(VI) nitrate and An(IV) nitrate concentrations is ?55 g/L. The method includes contacting the organic solution and an aqueous solution of nitric and oxalic acids, the oxalic acid concentration in the aqueous solution and the O/A volume ratio selected so that the oxalic acid is deficient with respect to the stoichiometric conditions of a complete precipitation of U(VI) and actinide(IV), to obtain a precipitate containing the actinide(IV) in oxalate form and a fraction of the U(VI) in oxalate form with a U(VI)/actinide(IV) mass ratio of between 0.5 and 5; and separating the precipitate from the organic and aqueous solutions.