27results about "Uranium fluorides" patented technology

The invention relates to the technical field of uranium hexafluoride recovery, in particular to a two-stage recovery uranium hexafluoride recovery system. The uranium hexafluoride recovery system comprises a carbon reactor, an inlet of which is connected with a feeding pipeline through an electromagnetic valve and an outlet of which is connected with a condensing device; the condensing device comprises a first outlet, a second outlet and at least one condenser, and the condenser is used for condensing uranium hexafluoride into a solid state or a liquid state; the first outlet is connected with a recovery pipeline, and the second outlet is connected with a tail gas treatment system; the recovery pipeline is provided with heating medium heat tracing and is used for heating the solid uranium hexafluoride in the recovery pipeline to be in a gas state or a liquid state; and the recovery container is arranged in the cooling container filled with the cooling medium and is used for cooling the uranium hexafluoride entering the recovery container into a solid state for storage. According to the uranium hexafluoride recovery system with the two-stage recovery function, uranium hexafluoride can be safely and efficiently recovered.

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 UF6, a source of inert gas and a source of water vapor. The tubular injector (9) is obtained by additive manufacturing.

The invention belongs to the technical field of uranium conversion production processes, and particularly relates to a vertical fluorination reactor protection device which comprises a reactor body, connecting flanges welded to the two ends of the reactor body and connecting plates installed on the sides, away from the reactor body, of the connecting flanges through bolts. The protective assembly is mounted between the connecting flange and the connecting plate; the protection assembly is arranged, the Monel webs are arranged between the connecting flange and the connecting plate, the Monel webs are connected through the fixing plate, the weakest mortise and tenon joint structure of the connecting flange is provided with the Monel webs, the production cost is greatly saved, the corrosion rate of the furnace body connecting flange is reduced, and the service life of the furnace body connecting flange is prolonged. The maintenance frequency is reduced, the manual maintenance workload is reduced, the irradiation receiving time of maintenance personnel in a production field is effectively shortened, and meanwhile secondary pollution to the environment during maintenance is reduced.

A baffled moving bed gas-solid reaction device, a method of use, and a hydrofluorination production system belong to the technical field of uranium conversion production. The present invention is to improve the fluidity of uranium dioxide powder in the equipment so that uranium dioxide can fully contact with hydrogen fluoride in the tail gas of the uranium dioxide hydrofluorination process, thereby reducing the generation rate of fluorine-containing waste. The present invention includes a bed body, a moving center axis and a baffle, the bed body is provided with a feed pipe and an air inlet pipe, the top of the bed body is provided with an air outlet pipe, the bottom of the bed body is provided with a discharge port, a moving center axis is provided in the bed body, and a baffle is arranged on the moving center axis. The present invention has an ingenious structural design, a compact structure, complete functions, is easy to install and maintain, can effectively realize the reuse of hydrogen fluoride in the tail gas of the uranium dioxide hydrofluorination process, and is suitable for promotion and use.

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

The invention relate to a method for producing a powder comprising UN2 and / or U2N3 and / or UN, the method comprising the steps of: a) heating a reaction zone in a vessel containing an NH3 atmosphere to a temperature in the range of 700 – 1200 °C to drive a reaction between UF6 and NH3, the pressure in the vessel being from 1 up to 45 atm; b) continuously injecting gaseous UF6 as a process gas into the reaction zone, optionally co-injecting with a shield gas around the UF6; c) continuously injecting gaseous NH3 as a process gas into the reaction zone, the amount of NH3 controlled to be in stochiometric excess for the reaction UF6 + 2 NH3 -> UN2 + 6 HF; d) reacting the process gases in the reaction zone, thereby forming the powder and gaseous HF; e) continuously, evacuating the powder and process gases including HF and excessive NH3 from the reaction zone; f) separating the powder from the process gases. The invention also relates to a reaction vessel having walls comprising a ceramic and / or a metallic material that preferably have low or zero catalytic activity for ammonia dissociation.

Provided is a process which allows uranium and molybdenum fluorides to be efficiently separated, said process comprising a step of providing a mixture containing MoF6 and UF6; a step of reducing the UF6 to UF5 in the gas phase or in a liquid phase; and a step of separating the UF5 and the MoF6 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.

Described herein are methods for recovering uranium tetrafluoride (UF4) from uranium hexafluoride (UF6) by directly dissolving UF6 in ionic liquids and recovering UF4, which can be processed to obtain UO2 (s) or uranium metal.

The invention belongs to the technical field of nuclear power, and particularly relates to a device and method for preparing UF4. The invention provides a device and a process method for directly preparing uranium tetrafluoride from a uranyl nitrate solution. The uranium tetrafluoride preparation device combining the characteristics of centrifugal spray drying and a horizontal fluorination reactor is designed for the first time. The uranium tetrafluoride is directly prepared from the uranyl nitrate solution in the same reactor through denitration, reduction and fluorination. Therefore, the technical problems that in an existing small-amount high-abundance UF4 preparation process, the material input amount of a dry process is large, the nuclear critical safety pressure of a wet process is large, and the efficiency is low can be effectively solved, and a safe and efficient preparation device and an operation process method are provided for hundreds of grams to several kilograms of high-abundance materials in scientific research and production.

A uranium tetrafluoride preparation system based on a reducing hydrofluorination reaction includes a gas supply mechanism, a three-stage fluidized bed mechanism, and a tail gas treatment mechanism. The gas supply mechanism includes an ammonium fluoride heating tank and a gas separation tank. The gas separation tank has a mixed gas inlet, an ammonia outlet A, and a hydrogen fluoride outlet at its top. The mixed gas inlet of the gas separation tank is connected to the mixed gas outlet of the ammonium fluoride heating tank. The three-stage fluidized bed mechanism includes a fluidized bed A, an intermediate container X, a fluidized bed B, an intermediate container Y, and a fluidized bed C. The tail gas treatment mechanism includes a condenser, a crystallization tank, a plate filter, a microwave dryer, and a scrubbing tower. This invention uses a three-stage fluidized bed system to reduce and hydrofluorinate uranium oxide to prepare uranium tetrafluoride. Compared with the existing "uranium purification-uranium conversion" process, it enables continuous production control and significantly shortens the process flow.

A novel semi-batch process is provided for converting high-assay low-enriched uranium (HALEU) from the uranium hexafluoride state to uranium dioxide and other chemical states useful as a feed for nuclear fuel in a nuclear reactor. This semi-batch process enables the use of a sufficiently small apparatus and a sufficiently low production rate to meet the nuclear criticality safety constraints for HALEU, and enables the safe, reliable and economical production of HALEU feed for nuclear fuel at a nominal capacity of up to approximately 20 MTU (metric tons of uranium metal) per year per denitration reactor.

Method and system for detecting a uranium hexafluoride leak. The detection method allows for the detection of the presence of uranium hexafluoride (UF6) in a gas stream (G). The detection method comprises humidifying the gas stream by circulating the gas stream (G) through contact with a copolymer fluoropolymer based on sulfonate tetrafluoroethylene, in particular Nafion, and then detecting the presence of hydrogen fluoride (HF) in the gas stream (G) using a sensor. Figure for the abstract: Figure 1

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 method for preparing nuclear pure grade uranium hexafluoride, which comprises the following steps: (1) adding fluoride containing molybdenum, tungsten, vanadium, titanium, chromium and antimony and crude uranium hexafluoride of hydrogen fluoride impurities into a distillation kettle, and carrying out pressurized distillation to obtain a gaseous mixture; (2) introducing the gaseous mixture obtained in the step (1) into a calcium fluoride adsorption column, and removing hydrogen fluoride in the mixture through selective adsorption of calcium fluoride; and (3) carrying out single-tower pressurized distillation on the gaseous mixture treated in the step (2) to obtain nuclear pure grade uranium hexafluoride. The method solves the problems of low uranium hexafluoride purity, high energy consumption and high cost of the existing uranium purification conversion technology in the conversion technology of the high-impurity-content natural uranium concentrate, realizes cost reduction and efficiency improvement of uranium hexafluoride purification on the premise of ensuring that uranium hexafluoride reaches the nuclear purity grade standard, and has a wide application prospect. Rapid popularization of a dry uranium purification and conversion process is boosted, and the technical level of uranium purification and conversion in China is further improved.

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 invention relates to the use of natural uranium hexafluoride UF6 as a propellant in an electric propulsion system.

The invention provides a device and method for preparing uranium hexafluoride based on uranium-containing molten salt, the uranium-containing molten salt contains uranium dioxide and / or triuranium octaoxide, and the device comprises a reaction cavity used for containing the uranium-containing molten salt; the gas inlet is communicated with the reaction cavity and is used for introducing fluorinated gas containing chlorine trifluoride into the reaction cavity, the chlorine trifluoride is subjected to fluorination reaction with each of uranium dioxide and / or triuranium octaoxide, and the fluorination reaction generates reaction gas containing uranium hexafluoride; the gas outlet is communicated with the reaction cavity and is used for discharging reaction gas; and the separation unit is used for separating uranium hexafluoride from the reaction gas.

Provided is a process which allows uranium and molybdenum fluorides to be efficiently separated, said process comprising a step of providing a mixture containing MoF6 and UF6; a step of reducing the UF6 to UF5 in the gas phase or in a liquid phase; and a step of separating the UF5 and the MoF6 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.

The invention relates to chemistry and metallurgy. A method for producing uranium tetrafluoride includes obtaining a solution of tetravalent uranium, precipitating uranium-containing raw material from said solution to produce a sludge, filtering said sludge, washing the residue, and drying and calcining same. The solution of tetravalent uranium is obtained using, as a starting material, a solution of a uranium-rich eluate formed as a result of the sorption concentration of uranium from pregnant solutions during the in-situ leaching of uranium, said solution is pre-acidified with sulphuric acid to a concentration of 200-250 g / dm3, and the acidified solution is passed through a layer of zinc, or uranium is electrochemically reduced using a uranium-rich eluate pre-acidified with sulphuric acid to a concentration of 40-45 g / dm3. The uranium-containing raw material is precipitated in the form of ammonium uranium pentafluoride by the introduction of ammonium bifluoride in three stages every 15 minutes at 90-100°C under vigorous stirring, and washing is carried out using process water at a liquid to solid ratio of 3:1. Drying is carried out at 210-250°C for one hour, and calcining is carried out at 600-650°C for one hour. The invention makes it possible to precipitate crystalline hydrates of ammonium uranium pentafluoride directly from solutions of uranium-rich eluates, thus reducing consumption of chemical reagents and electrical energy.

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 discloses a spent fuel dry reprocessing method based on plasma, the method includes the following steps: S1, processing a spent fuel after removal of a cladding thereof into a spent fuel powder, and sending the spent fuel powder into a plasma reactor; S2, fully mixing and reacting the spent fuel powder with a plasma containing F atoms, to generate a product; S3, rapidly cooling the product, to generate a gas-solid two-phase flow and a solid-state product; S4, filtering the gas-solid two-phase flow to remove a solid-state product therein, then carrying out condensation, adsorption, desorption and distillation, to obtain a mixed gas; S5, performing a recovery process to the mixed gas to generate a corresponding oxide or metal. Through reaction of the F atoms of the plasma with U, Pu and other components in the spent fuel in the present disclosure, compared with a flame fluorination process, the plasma has effects of a wider temperature range and a stronger activity of a fluorinating agent, a complete conversion of Pu to PuF6 is realized, the problem of separation of Pu from a fluoride of a fission product is solved, the separation of U and Pu from the fission product is promoted, a recovery rate is improved, and a process is simplified.

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 application provides a device and a method for preparing uranium hexafluoride based on a uranium-containing molten salt, the uranium-containing molten salt containing uranium dioxide and / or triuranium octaoxide, the device comprising: a reaction cavity for containing the uranium-containing molten salt; a gas inlet in communication with the reaction cavity, for introducing a fluorination gas containing chlorotrifluoride into the reaction cavity, the chlorotrifluoride being fluorinated with each of the uranium dioxide and / or the triuranium octaoxide, the fluorination reaction generating a reaction gas comprising uranium hexafluoride; a gas outlet in communication with the reaction cavity, for discharging the reaction gas; and a separation unit for separating the uranium hexafluoride from the reaction gas.

The invention belongs to the technical field of uranium concentration, and particularly relates to a device for purifying light impurities. The device comprises a membrane component mounted before a material enters a material receiving container; the membrane component comprises a shell and separation membrane filaments arranged in the shell; the separation membrane wire is used for separating UF6 and light impurities; an inlet of the membrane component is connected with the material, a product outlet is connected with a material receiving container, and an impurity outlet is connected with an impurity collecting device; the impurity collecting device is used for purifying tail gas. The process flow for purifying the light impurities in the product is simplified, the workload is reduced, the working efficiency is improved, continuous, stable and efficient separation of the materials and the light impurities in the uranium concentration production product is achieved, and a foundation is laid for follow-up optimization of the layout of a material supply and taking plant, reduction of the engineering construction cost and the operation cost and improvement of the system automation level.

The invention relates to a method for producing uranium dinitride or lower nitrides comprising the steps of: a) providing a reactor made of steel; b) providing to the reactor an NH3 atmosphere heated to 200-800 °C at a pressure of 100-600 bar; c) providing a powder comprising uranium tetrafluoride and / or tetravalent ammonium uranium fluorides to the reactor for a time period sufficient to react the uranium tetrafluoride and / or tetravalent ammonium uranium fluorides to uranium dinitride or lower nitrides; and d) optionally heating the powder in step c) in a vacuum or inert atmosphere at temperatures between 1000 – 1300 °C up to 10 hours for conversion of the powder to uranium mononitride.

The present invention belongs to the technical field of uranium hexafluoride production from natural uranium, and specifically relates to a device and method for preparing uranium hexafluoride from natural uranium by low-temperature plasma chemical reaction. The integrated device for preparing uranium hexafluoride by low-temperature plasma fluorination is developed, including: a gas supply system, a low-temperature plasma fluorination reaction system, a material collection system, a vacuum system and an exhaust gas treatment system. The method includes the following steps: preparation of the gas supply system, generation and regulation of low-temperature plasma, preparation of uranium hexafluoride by low-temperature plasma fluorination, system atmosphere replacement, unloading weighing, vacuum purification and calculation of uranium hexafluoride product yield. The device and method provided by the present invention effectively promote the ionization of fluorine gas through argon dielectric barrier discharge technology, providing a new technical solution for broadening the preparation path of uranium hexafluoride.