A system is provided that determines optimal movements of fuel assemblies in a nuclear reactor, such as a traveling wave reactor (TWR). Such a system may be capable of modeling core operations and fuel moves in parallel to determine optimal fuel cycle moves responsive to one or more constraints, including, but not limited to core criticality and location of a deflagration wave within an operating reactor core. According to one embodiment, the optimal solution may be determined using a branch search to simulate possible fuel moves.

A system is provided that determines optimal movements of fuel assemblies in a nuclear reactor, such as a traveling wave reactor (TWR). Such a system may be capable of modeling core operations and fuel moves in parallel to determine optimal fuel cycle moves responsive to one or more constraints, including, but not limited to core criticality and location of a deflagration wave within an operating reactor core. According to one embodiment, the optimal solution may be determined using a branch search to simulate possible fuel moves.

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.

A rod transfer assembly has an outer rotating plug. A pick-up arm assembly extends from the outer rotating plug and includes a pivoting arm. An inner rotating plug is disposed off-center from and within the outer rotating plug and is rotatable independent of a rotation of the outer rotating plug. An access port rotating plug is disposed off-center from and within the inner rotating plug and is rotatable independent of rotation of the outer and inner rotating plugs. A pull arm extends from the access port rotating plug.

A rod transfer assembly has an outer rotating plug. A pick-up arm assembly extends from the outer rotating plug and includes a pivoting arm. An inner rotating plug is disposed off-center from and within the outer rotating plug and is rotatable independent of a rotation of the outer rotating plug. An access port rotating plug is disposed off-center from and within the inner rotating plug and is rotatable independent of rotation of the outer and inner rotating plugs. A pull arm extends from the access port rotating plug.

Systems and methods for producing acid deficient uranyl nitrate from a dilute uranyl nitrate solution are disclosed. In one form, the present disclosure provides a system comprising a feed evaporation system and a diffusion dialysis system. The feed evaporation system is configured to receive a feed stream and to boil off water, under vacuum, from the feed stream to produce a concentrated uranyl nitrate solution and a distilled water product. The diffusion dialysis system is configured to counter flow the concentrated uranyl nitrate solution and the distilled water product across a plurality of membrane vessels to promote nitrate migration from the concentrated uranyl nitrate solution to the distilled water, and to produce a dialysate stream and a recycle acid stream. The feed stream may include a product of a solvent extraction process used to recycle spent nuclear fuel and/or a recovery stream from other fuel fabrication activities.

Systems and methods for producing acid deficient uranyl nitrate from a dilute uranyl nitrate solution are disclosed. In one form, the present disclosure provides a system comprising a feed evaporation system and a diffusion dialysis system. The feed evaporation system is configured to receive a feed stream and to boil off water, under vacuum, from the feed stream to produce a concentrated uranyl nitrate solution and a distilled water product. The diffusion dialysis system is configured to counter flow the concentrated uranyl nitrate solution and the distilled water product across a plurality of membrane vessels to promote nitrate migration from the concentrated uranyl nitrate solution to the distilled water, and to produce a dialysate stream and a recycle acid stream. The feed stream may include a product of a solvent extraction process used to recycle spent nuclear fuel and/or a recovery stream from other fuel fabrication activities.

The present invention relates to a method of adjusting the oxoacidity of a molten metal hydroxide salt, the method comprising the steps of: estimating a target concentration of at least one of H.sub.2O, O.sup.2, and OH in a molten salt of a metal hydroxide; providing an oxoacidity control component; and contacting the oxoacidity control component with the molten salt of a metal hydroxide to adjust the oxoacidity of the molten salt of a metal hydroxide. The method allows better utilisation of the available temperature range for a molten salt of a metal hydroxide by reducing the corrosive nature of the metal hydroxide.

The present invention is directed to a fission product extraction system operable to capture and extract fission products from a flow of irradiated fueled molten salt of a molten salt reactor. The example extraction systems described herein utilize electroless deposition to chemically plate fission products onto a metallic structure. The metallic structure may be partially coated with a reducing agent, such as beryllium to provide an electron source for the fission products. The metallic structure may be a component of an extraction system designed to facilitate submersion of the metallic structure into the flow of molten salt. The extraction system may also be designed to facilitate removal of the metallic structure without requiring a shut down or slowdown of the reactor system.