A process for dissolving nuclear fuel, in particular irradiated nuclear fuel, comprising immersion of the nuclear fuel in a nitric acid solution. This dissolution process further comprises mechanical milling of the nuclear fuel, this mechanical milling being performed in the nitric acid solution during the immersion. The disclosure also relates to the use of a mill equipped with mechanical milling structure to implement the dissolution process.

Provided are methods to separate an isotope from a first solution including uranium. The methods may include (a) cleaning the first solution to form a second solution including the uranium and a third solution including the isotope; (b) oxidizing the third solution to form an oxidized isotope; and (c) separating the oxidized isotope.

The present invention relates to a method for recovering plutonium from spent radioactive fuel. In one embodiment, the method comprises steps of extracting tetravalent plutonium from an aqueous solution of the spent radioactive fuel using a first organic solvent comprising tributyl phosphate; reducing tetravalent plutonium to trivalent plutonium by adding to an organic phase a second organic solvent comprising dimethylhydroxylamine; and stripping plutonium into the aqueous phase for recycling by adding an aqueous dilute acid solution into an organic phase. The method can significantly improve the efficiency of recovering plutonium from spent radioactive fuel compared with HAN stripping, and at the same time, can avoid the problems resulting from U(IV) reduction and extraction.

The present invention relates to a method for recovering plutonium from spent radioactive fuel. In one embodiment, the method comprises steps of extracting tetravalent plutonium from an aqueous solution of the spent radioactive fuel using a first organic solvent comprising tributyl phosphate; reducing tetravalent plutonium to trivalent plutonium by adding to an organic phase a second organic solvent comprising dimethylhydroxylamine; and stripping plutonium into the aqueous phase for recycling by adding an aqueous dilute acid solution into an organic phase. The method can significantly improve the efficiency of recovering plutonium from spent radioactive fuel compared with HAN stripping, and at the same time, can avoid the problems resulting from U(IV) reduction and extraction.

A method of producing uranium halides is disclosed in which chlorine gas is introduced into a liquid uranium-nickel alloy. NaCl salt is surrounding the crucible containing the liquid uranium-nickel alloy, producing a eutectic mixture of NaClUCl.sub.3. Upon chlorination, the metal halide dissolves in the matrix salt forming a solution. Adding the reactant metal, uranium to the nickel, the alloy is able to remain molten throughout processing. The liquid metal alloy may be removed from the salt bath, while the halide gas continues to enter the system through the sparge until the desired composition of NaClUCl.sub.3UCl.sub.4 is achieved. The method and system can be used to produce other metal halide salts such as actinide, lanthanide or transition metal halides contained in a matrix salt consisting of alkali and/or alkaline earth halides.

A method of producing uranium halides is disclosed in which chlorine gas is introduced into a liquid uranium-nickel alloy. NaCl salt is surrounding the crucible containing the liquid uranium-nickel alloy, producing a eutectic mixture of NaClUCl.sub.3. Upon chlorination, the metal halide dissolves in the matrix salt forming a solution. Adding the reactant metal, uranium to the nickel, the alloy is able to remain molten throughout processing. The liquid metal alloy may be removed from the salt bath, while the halide gas continues to enter the system through the sparge until the desired composition of NaClUCl.sub.3UCl.sub.4 is achieved. The method and system can be used to produce other metal halide salts such as actinide, lanthanide or transition metal halides contained in a matrix salt consisting of alkali and/or alkaline earth halides.

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.

A single integrated system for recycling used nuclear fuel (UNF) emerging from a reactor has a decladding vessel separating fuel pellets from nuclear fuel rods via oxidation to produce U.sub.3O.sub.8. A fluorination vessel is coupled to the decladding vessel to remove hexafluorides from the U.sub.3O.sub.8 produced by the decladding vessel. An electrowinning vessel is coupled to the fluorination vessel removing plutonium and actinides via electrowinning.

A single integrated system for recycling used nuclear fuel (UNF) emerging from a reactor has a decladding vessel separating fuel pellets from nuclear fuel rods via oxidation to produce U.sub.3O.sub.8. A fluorination vessel is coupled to the decladding vessel to remove hexafluorides from the U.sub.3O.sub.8 produced by the decladding vessel. An electrowinning vessel is coupled to the fluorination vessel removing plutonium and actinides via electrowinning.

A separation apparatus for separating a supply of high-level nuclear waste (HLW), where the HL nuclear waste is separated into high-mass and low-mass portions. The high-and-low mass portions of the HLW have respective atomic masses that are above and below an atomic mass cut-off point of the separation apparatus. The separation apparatus includes first and second ICP torches that are respectively mounted to and within an apparatus housing. The apparatus housing defines a cylindrical separation chamber and includes first and second magnetic elements which generate a magnetic field along the length of the separation chamber, and a plurality concentric ring electrodes which generate an electric field that is perpendicular to, and which crosses the magnetic field. The supply of HLW is subject to a mass separation process within the separation chamber using the set of crossed electric and magnetic fields.