A method of synthesizing alkali uranium fluorophosphate crystals. The method includes combining a uranium-based feedstock with a mineralizer solution. The mineralizer solution includes an alkali nutrient, a phosphate, and a fluoride. The feedstock and mineralizer solution are pressurized and a thermal gradient applied thereto such that a first portion of the feedstock and the mineralizer solution is heated to a temperature that is greater than a temperature of a second portion of the feedstock and the mineralizer solution. Uranium nutrient enters the mineralizer solution from the feedstock in the first portion and uranium nutrient precipitates to spontaneously form crystals in the second portion.

A method of synthesizing alkali uranium fluorophosphate crystals. The method includes combining a uranium-based feedstock with a mineralizer solution. The mineralizer solution includes an alkali nutrient, a phosphate, and a fluoride. The feedstock and mineralizer solution are pressurized and a thermal gradient applied thereto such that a first portion of the feedstock and the mineralizer solution is heated to a temperature that is greater than a temperature of a second portion of the feedstock and the mineralizer solution. Uranium nutrient enters the mineralizer solution from the feedstock in the first portion and uranium nutrient precipitates to spontaneously form crystals in the second portion.

A method for treating process distillate heavies produced during uranium fluoride purification is described. The heavies contain primarily uranium hexafluoride, UF.sub.6, and molybdenum oxytetrafluoride, MoOF.sub.4. The uranium hexafluoride is removed via distillation at reduced pressure leaving essentially MoOF.sub.4 containing <0.1% of residual uranium hexafluoride. This mixture is hydrolyzed in water, then treated with a solution of sodium hydroxide until a pH of at least 7.5 is reached. The precipitated sodium diuranate and sodium fluoride are removed by filtration. The filtrate is reacted with calcium chloride to precipitate the molybdenum values as calcium molybdate containing trace quantities of calcium fluoride.

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 10 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 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 10 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.

Provided is a process which allows uranium and molybdenum fluorides to be efficiently separated, said process comprising a step of providing a mixture containing MoF.sub.6 and UF.sub.6; a step of reducing the UF.sub.6 to UF.sub.5 in the gas phase or in a liquid phase; and a step of separating the UF.sub.5 and the MoF.sub.6 or a conversion product thereof which may be obtained by further converting the molybdenum fluoride to another molybdenum compound. In a further aspect, a process for the fluorination of metals or semimetals is provided.

Provided is a process which allows uranium and molybdenum fluorides to be efficiently separated, said process comprising a step of providing a mixture containing MoF.sub.6 and UF.sub.6; a step of reducing the UF.sub.6 to UF.sub.5 in the gas phase or in a liquid phase; and a step of separating the UF.sub.5 and the MoF.sub.6 or a conversion product thereof which may be obtained by further converting the molybdenum fluoride to another molybdenum compound. In a further aspect, a process for the fluorination of metals or semimetals is provided.

A reactor (1) for hydrolysis of uranium hexafluoride comprises a tubular injector (9) comprising first (11), second (13) and third (15) concentric fluid circulation ducts intended to be connected respectively with a source of UF.sub.6, a source of inert gas and a source of water vapor. The tubular injector (9) is obtained by additive manufacturing.

Described herein are methods for recovering uranium tetrafluoride (UF.sub.4) from uranium hexafluoride (UF.sub.6) by directly dissolving UF.sub.6 in ionic liquids and recovering UF.sub.4, which can be processed to obtain UO.sub.2 (s) or uranium metal.

Described herein are methods for recovering uranium tetrafluoride (UF.sub.4) from uranium hexafluoride (UF.sub.6) by directly dissolving UF.sub.6 in ionic liquids and recovering UF.sub.4, which can be processed to obtain UO.sub.2 (s) or uranium metal.