The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

According to embodiments, a light water reactor uranium fuel assembly is capable of reducing heating values of both Am-241 and Cm-244, to reduce the amount of generated vitrified waste without using fast reactors. The light water reactor uranium fuel assembly is a light water reactor uranium fuel assembly to be used in a nuclear fuel cycle that extracts. An americium isotope is extracted at the time of reprocessing of spent fuel to be added to a fuel, in which a weight fraction W (unit: wt %) of americium 241 to be added to a fuel heavy metal is in ranges of W<−0.006e.sup.2+0.12e−0.43 (enrichment: 5 wt % or more), W<−0.000356e+0.00357 (enrichment: 4.2 wt % or more and less than 5.0 wt %) with respect to an average enrichment of uranium 235 e (unit: wt %) of the fuel assembly.

According to embodiments, a light water reactor uranium fuel assembly is capable of reducing heating values of both Am-241 and Cm-244, to reduce the amount of generated vitrified waste without using fast reactors. The light water reactor uranium fuel assembly is a light water reactor uranium fuel assembly to be used in a nuclear fuel cycle that extracts. An americium isotope is extracted at the time of reprocessing of spent fuel to be added to a fuel, in which a weight fraction W (unit: wt %) of americium 241 to be added to a fuel heavy metal is in ranges of W<−0.006e.sup.2+0.12e−0.43 (enrichment: 5 wt % or more), W<−0.000356e+0.00357 (enrichment: 4.2 wt % or more and less than 5.0 wt %) with respect to an average enrichment of uranium 235 e (unit: wt %) of the fuel assembly.

The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

The invention provides a system for collecting metal in an electrorefining process, the system having a hollow cathode; and a container defining an upwardly extending surface adapted to be received by the hollow cathode. An embodiment of the invention provides for metal reduction to occur on laterally facing and medially facing surfaces of the cathode such that electrolyte resides between surfaces of the cathode. Also provided is a metal electrorefining process having the steps of subjecting molten salt containing metal moieties to electrolysis wherein reduced metal accumulates in a cathode-cup construct in a first position; raising the construct to a second position above the molten salt while subjecting the construct to heat from the molten salt; withdrawing the cathode from the construct into a vestibule to the electrorefiner to a third position; and removing the cathode and cup from the electrorefiner to a fourth position.

Spent nuclear fuel is added to an electro-reduction cell, wherein the electro-reduction cell includes a halide salt electrolyte, and anode, and a cathode including an alloy of uranium and a first metal forming a low melting point alloy with uranium, the first metal being one or more of: iron; chromium; nickel; manganese; and cobalt. The spent nuclear fuel is electrochemically reduced at a potential sufficient to reduce plutonium and lanthanides in the spent nuclear fuel, to form a molten alloy of the first metal, uranium and higher actinides present in the spent nuclear fuel. The alloy is extracted from the electro-reduction cell while uranium oxide is present in the electro-reduction cell. The spent nuclear fuel includes uranium oxide and at least 1 mol of lanthanides per tonne of uranium in the spent nuclear fuel, and the electro-reduction cell is operated at a temperature above the melting point of the alloy.

Spent nuclear fuel is added to an electro-reduction cell, wherein the electro-reduction cell includes a halide salt electrolyte, and anode, and a cathode including an alloy of uranium and a first metal forming a low melting point alloy with uranium, the first metal being one or more of: iron; chromium; nickel; manganese; and cobalt. The spent nuclear fuel is electrochemically reduced at a potential sufficient to reduce plutonium and lanthanides in the spent nuclear fuel, to form a molten alloy of the first metal, uranium and higher actinides present in the spent nuclear fuel. The alloy is extracted from the electro-reduction cell while uranium oxide is present in the electro-reduction cell. The spent nuclear fuel includes uranium oxide and at least 1 mol of lanthanides per tonne of uranium in the spent nuclear fuel, and the electro-reduction cell is operated at a temperature above the melting point of the alloy.

A method for converting uranium hexafluoride to uranium dioxide includes steps of hydrolysis of UF.sub.6 to uranium oxyfluoride (UO.sub.2F.sub.2) in a hydrolysis reactor (4) by reaction between gaseous UF.sub.6 and dry water vapour injected into the reactor (4), and pyrohydrolysis of UO.sub.2F.sub.2 to UO.sub.2 in a pyrohydrolysis furnace (6) by reaction of UO.sub.2F.sub.2 with dry water vapour and hydrogen gas (H.sub.2) injected into the furnace (6). The hourly mass flowrate of gaseous UF.sub.6 supplied to the reactor (4) is between 75 and 130 kg/h, the hourly mass flowrate of dry water vapour supplied to the reactor (4) for hydrolysis is between 15 and 30 kg/h, and the temperature inside the reactor (4) is between 150 and 250° C.

A method for converting uranium hexafluoride to uranium dioxide includes steps of hydrolysis of UF.sub.6 to uranium oxyfluoride (UO.sub.2F.sub.2) in a hydrolysis reactor (4) by reaction between gaseous UF.sub.6 and dry water vapour injected into the reactor (4), and pyrohydrolysis of UO.sub.2F.sub.2 to UO.sub.2 in a pyrohydrolysis furnace (6) by reaction of UO.sub.2F.sub.2 with dry water vapour and hydrogen gas (H.sub.2) injected into the furnace (6). The hourly mass flowrate of gaseous UF.sub.6 supplied to the reactor (4) is between 75 and 130 kg/h, the hourly mass flowrate of dry water vapour supplied to the reactor (4) for hydrolysis is between 15 and 30 kg/h, and the temperature inside the reactor (4) is between 150 and 250° C.

A nuclear fuel powder production plant comprises a conversion installation (2) for the conversion of uranium hexafluoride (UF.sub.6) into uranium dioxide (UO.sub.2) having a hydrolysis reactor (4) for the conversion of UF.sub.6 into uranium oxyfluoride powder (UO.sub.2F.sub.2) and a pyrohydrolysis furnace (6) for converting the UO.sub.2F.sub.2 powder into UO.sub.2 powder. The nuclear fuel powder production plant also includes a packaging unit (20) for the UO.sub.2 powder comprising a filling station (22) having a chamber (26) for receiving a container (24) to be filled, a filling duct (28) supplied from the furnace (6) and a suction system (32) comprising a suction ring (34) disposed at the outlet (30) of the filling duct (28) for sucking an annular air flow (A) around a stream (P) of UO.sub.2 powder falling from the outlet (30) from the filling duct (28) into the container (24).