12results about "Uranium oxides/hydroxides" patented technology

A method and system (300) for separating thorium from uranium and its decay products is provided. The method includes combining a nuclear fuel feedstock (102) containing thorium and uranium with a first acid (104, 204) to form a first solution. The first solution is contacted with an ion exchange resin (308) that is selective for thorium or uranium. The thorium or uranium is at least partially removed from the first solution by binding the thorium or uranium to the ion exchange resin (308), thereby forming a second solution (106, 206). The second solution (106, 206) is combined with oxalic acid to precipitate the uranium or thorium from the second solution, forming a precipitate (110, 210). The precipitate (110, 210) is separated from the second solution (106, 206).

The invention relates to a process for preparing a powder comprising one or more oxides selected from uranium oxide UO2, plutonium oxide PuO2 and minor actinide oxides, the minor actinides being selected from americium, neptunium and curium, comprising the steps of: a) cryogenic granulation of an aqueous solution comprising cations selected from uranium-based cations, plutonium-based cations and minor actinide-based cations; b) freeze-drying of the granules obtained in a); etc.; calcination of the granules from b). Applications: manufacture of nuclear fuels or blankets loaded with minor actinide(s).

A crucible used to form ceramic particles from metal oxide gel particles includes a cylindrical graphite housing having a seat at an open end, on its inner surface, and on the inner surface near the open end. A sleeve is positioned along the inner surface of the cylindrical housing. This sleeve has an open end and is made of a metal that is chemically inert to the metal oxide gel particles. An outer cap made of graphite detachably covers the open end of the cylindrical housing. An inner cap made of a chemically inert metal fits into the seat on the inner surface of the cylindrical housing and is pressed against the open end of the sleeve by the outer cap. This crucible can be used to form ceramic particles from uranium oxide gel particles, and the sleeve and inner cap can be made of molybdenum, tungsten, or an alloy thereof.

The present invention relates to a method for obtaining a bath of molten chloride salts comprising at least one actinide in which the oxidation state and / or degree of oxidation of said at least one actinide are predetermined according to the intended application, comprising the following steps: (a) establishing, for said at least one actinide, its potential-oxoacidity diagram at a given temperature T in a bath of molten chloride salts, (b) from the potential-oxoacidity diagram established in step (a), defining the potential-oxoacidity domain in which said at least one actinide is in the predetermined oxidation state and / or degree of oxidation, (c) selecting at least three different compounds which are different from said at least one actinide setting, at the temperature T, the potential and oxoacidity of the bath of molten chloride salts in the domain defined in step (b), it being possible for one of said at least three compounds to be the chloride ion contained in the bath of molten chloride salts, then (d) introducing, into a bath of molten chloride salts comprising said at least one actinide at the temperature T, said at least three compounds selected in step (c). The present invention relates to the use of various mixtures or compositions in the preparation of a nuclear fuel, during its operation or else during its reprocessing.

The rational use of nuclear energy plays a vital role in the process of consolidating national defense and security and green transformation of energy system in China. However, the proper management and resource utilization of depleted uranium (especially 238 U, which does not pose a direct radiation threat to human body) generated from the application of nuclear energy have become a technical problem to be solved in the field of environmental protection and sustainable development. In particular, we use U3O8 prepared from depleted UF6 as raw material, and use UO4·2H2O prepared by simple dissolution and precipitation as catalyst to prepare 1-oxidized diphenyldiazene or its derivatives by catalytic oxidation of aniline. On the one hand, the invention creates a way to reuse depleted uranium waste, and finds a new way to transform this traditionally regarded as a problem waste, on the other hand, it ingeniously builds a bridge from aniline and its derivatives to high value-added chemicals-1-oxidized diphenyldiazene or its derivatives, which shows extraordinary value in environmental, economic and social benefits. Whether from the perspective of environmental friendliness and sustainable development, or from the perspective of synthesizing higher value-added products, it has extremely important significance.

The group of inventions relates to a nuclear fuel material and represents a high-temperature dense composite nuclear fuel material and a method of its production. The high-temperature dense composite nuclear fuel material comprises a radiation inert ceramic matrix in which nuclear fuel particles are distributed. The matrix is made of a powder of a silicon carbide based material. The nuclear fuel particles are particles of an oxygen-free nuclear fuel. The method of producing the high-temperature dense composite nuclear fuel material includes preparation of a mixture of nuclear fuel particles and a powder of the radiation inert ceramic matrix, molding of the mixture by pressing, and sintering of the molded mixture. The molded mixture is sintered by hot pressing. The group of inventions makes it possible to produce the high-temperature dense composite nuclear fuel material which, after annealing in vacuum at 0.63Tmelt for 10 hours, exhibits a loss in mass of not more than 2%.

The wisdom utilization of nuclear energy plays a core and irreplaceable role in consolidating national defense security and transforming energy system to green. While the nuclear energy technology is booming, how to properly manage and efficiently resource the depleted uranium (especially 238 U, although it does not constitute direct radioactive hazard to human body) generated in the process of nuclear energy application has become an urgent and key technical challenge in the field of environmental protection and sustainable development. In particular, we use U3O8 prepared from depleted UF6 as raw material, and use UO4·2H2O prepared by simple dissolution and precipitation as catalyst to catalyze the synthesis of diphenyldiazene or its derivatives from aniline: one aspect of the invention creates a way to reuse depleted uranium waste, finding a new way to transform this traditionally difficult waste, and on the other hand, it ingeniously builds a bridge from aniline and its derivatives to high value-added chemicals-diphenyldiazene or its derivatives. Not only can it effectively reduce the potential threat of depleted uranium waste to the environment, realize the green transformation and resource utilization of waste, but also can promote the chemical industry to a more efficient and environmentally friendly direction, and bring significant economic benefits to the chemical industry.

A crucible used for forming ceramic particles from metal oxide gel particles includes a tubular graphite housing having an open end, an inner surface, and a seat in the inner surface near the open end. A sleeve lines the inner surface of the tubular housing. The sleeve has at an open end and is formed from a metal which is chemically inert to the metal oxide gel particles. A graphite outer cap removably covers the open end of the tubular housing. An inner cap formed from the chemically inert metal fits into the seat in the inner surface of the tubular housing, and is pressed into the seat against the open end of the sleeve by the outer cap. The crucible may be used for forming ceramic particles from uranium oxide gel particles, and the sleeve and the inner cap may be formed from molybdenum, tungsten, or an alloy thereof.

The present invention relates to a method for obtaining a chloride molten salt bath comprising at least one actinide in which the state and / or degree of oxidation of said at least one actinide are predetermined, comprising the following steps: (a) establishing, for said at least one actinide, its potential-oxoacidity diagram at a given temperature T in a chloride molten salt bath, (b) from the potential-oxoacidity diagram established in step (a), defining the potential-oxoacidity range in which said at least one actinide is in the predetermined state and / or degree of oxidation, (c) selecting at least three different compounds for fixing, at temperature T, the potential and oxoacidity of the chloride molten salt bath in the range defined in step (b), one of said at least three compounds being the chloride ion contained in the chloride molten salt bath.then (d) introduce into a bath of molten chloride salts comprising said at least one actinide at temperature T said at least three compounds selected in step (c). The present invention relates to the use of different mixtures or compositions in the preparation of nuclear fuel, during its operation or during its reprocessing.

The invention provides a denitration UO3 continuous multi-stage hydration activation system and a denitration UO3 continuous multi-stage hydration activation method, belongs to the technical field of uranium purification and conversion, and is used for solving the problems that the specific surface area of activated UO3 obtained by an existing denitration UO3 activation method is remarkably different, and the activity needs to be further improved. The system comprises a first-stage hydration reactor, a second-stage hydration reactor, a third-stage hydration reactor and a fourth-stage hydration reactor which are sequentially connected in series, the rotary drum vacuum filter is communicated with an outlet of the fourth-stage hydration reactor; the dryer is communicated with an outlet of the rotary drum vacuum filter; the desalted water tank is communicated with the first-stage hydration reactor; the dryer comprises a first drying chamber at the temperature of 200-280 DEG C, a second drying chamber at the temperature of 290-370 DEG C, a first dewatering chamber at the temperature of 380-420 DEG C and a second dewatering chamber at the temperature of 380-480 DEG C which are sequentially arranged in the material flowing direction. Denitrated UO3 activated by the system is uniform in specific surface area and high in activity.

A solution of acid deficient uranyl nitrate has a formula of UO2(OH)y(NO3)2-y, where y ranges from 0.1 to 0.5. The solution is prepared by placing UxOz in aqueous nitric acid to produce a uranium solution, wherein x is 1 to 3 and z is 2 to 8; placing the uranium solution under a pressure greater than atmospheric pressure in a sealed reaction chamber; and heating the uranium solution to a desired temperature of between 150°C and 250°C by applying microwave energy to the uranium solution. The uranium solution is maintained at the desired temperature under a pressure of from 5 atmospheres to 40 atmospheres for a hold time of 15 minutes to 6 hours to produce the desired acid deficient uranyl nitrate.

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.