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 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.

Reaction chamber and methods of extraction of metal compounds, specifically tools for uranium hexafluoride (UF.sub.6) conversion into uranium dioxide (UO.sub.2) ceramic powder (up to 5% enrichment of U.sup.235) by applying a method of reductive pyrohydrolysis. In one aspect, the reaction chamber is a shell with upper and lower heads, comprising upper filtration area, equipped with metalceramic filters, regenerating nitrogen, the first reaction zone for conversion of uranium hexafluoride into uranyl fluoride, the second reaction zone with gas-distribution grid for building up fluidization layer for reduction of uranyl fluoride to uranium dioxide with a nozzle of steam, and hydrogen and nitrogen supply. On the side walls of the first reaction zone of the reaction chamber shell there are two nozzles located symmetrically for uranium hexafluoride, hydrogen and water steam supply. The chamber is equipped with a device for discharge of powder.

Reaction chamber and methods of extraction of metal compounds, specifically tools for uranium hexafluoride (UF.sub.6) conversion into uranium dioxide (UO.sub.2) ceramic powder (up to 5% enrichment of U.sup.235) by applying a method of reductive pyrohydrolysis. In one aspect, the reaction chamber is a shell with upper and lower heads, comprising upper filtration area, equipped with metalceramic filters, regenerating nitrogen, the first reaction zone for conversion of uranium hexafluoride into uranyl fluoride, the second reaction zone with gas-distribution grid for building up fluidization layer for reduction of uranyl fluoride to uranium dioxide with a nozzle of steam, and hydrogen and nitrogen supply. On the side walls of the first reaction zone of the reaction chamber shell there are two nozzles located symmetrically for uranium hexafluoride, hydrogen and water steam supply. The chamber is equipped with a device for discharge of powder.

A thermochemical method for storing and releasing thermal energy by means of a compound in solid form of formula AO.sub.xB.sub.y.zH.sub.2O, in which: A is an element selected from uranium (U) and thorium (Th); O is the element oxygen; B is an anion or an oxoanion; x is a number comprised between 0 and 4; y is a number comprised between 0 and 2; z is a number greater than 0 and less than 10; it being understood that at least one of x and y is different from 0 and that the compound of formula Th(SO.sub.4).sub.2.xH.sub.2O is excluded.

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.

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 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.

Described are methods for the recovery of uranium from uranium hexafluoride dissolved directly into ionic liquids.