23results about "Uranium oxides/hydroxides" patented technology

The wisdom utilization of nuclear energy occupies a core and irreplaceable position in consolidating national defense security and transforming an energy system to greenization. While the nuclear energy technology is vigorously developed, how to properly manage and efficiently recycle depleted uranium (especially 238U, although the depleted uranium does not cause direct radioactive hazards to human bodies) generated in the nuclear energy application process becomes a critical technical challenge faced by the field of environmental protection and sustainable development. Particularly, U3O8 prepared by depleting UF6 is used as a raw material, UO4. 2H2O prepared by a simple dissolution and re-precipitation method is used as a catalyst, and aniline is catalyzed and oxidized to synthesize diphenyl diazene or derivatives thereof. According to the invention, on one hand, a mode of recycling the depleted uranium waste is created, a brand new conversion way is found for the waste which is traditionally considered as a difficult problem, and on the other hand, a bridge from aniline and derivatives thereof to a high value-added chemical, namely diphenyl diazene or derivatives thereof, is skillfully erected. Potential threats of the depleted uranium waste to the environment can be effectively reduced, green transformation and resource utilization of the waste are achieved, meanwhile, the chemical industry can be promoted to proceed towards the more efficient and environment-friendly direction, and economic benefits are remarkably improved for the chemical industry.

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%.

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.

Metal oxide gel particles having a desired particle size are produced by preparing a low temperature aqueous metal nitrate solution containing hexamethylenetetramine as a feed solution; flowing the feed solution through a first conduit and exiting the first conduit as a first stream at a first flow rate to contact a high temperature non-aqueous drive fluid. The drive fluid flows through a second conduit at a second flow rate. Shear between the first stream and the drive fluid breaks the first stream into metal nitrate solution particles and decomposition of the hexamethylenetetramine converts the metal nitrate solution particles into metal oxide gel particles. A sensor device directed at the flow of the metal oxide gel particles within the stream of the drive fluid is used to optically measure the size of the metal oxide gel particles. The sensor measures the transmission of light absorbed by the metal oxide gel particles or the drive fluid so that as the metal oxide gel particles pass through the optical sensor, the transmission of light by the drive fluid changes over time. If the measured particle size is not approximately equal to the desired particle size, the particle size is corrected by adjusting the ratio of the first flow rate to the total flow rate, where the total flow rate is the sum of the first flow rate and the second flow rate.

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 invention discloses a high-activity uranium trioxide production method and equipment for treating large-flux uranyl nitrate, and the method comprises the following steps: introducing fuel gas into a combustion chamber through a gas supply system, and enabling the fuel gas to be subjected to combustion reaction to generate high-temperature gas; high-temperature gas is introduced into the reaction chamber, a uranyl nitrate solution is conveyed into the reaction chamber, the uranyl nitrate solution makes contact with the high-temperature gas, then a thermal decomposition reaction is conducted, and powder mixed gas is obtained; powder mixed gas in the reaction chamber is conveyed to a powder receiving hopper through an L discharging valve and filtered through a filtering device, then the gas is introduced into a tail gas purification system to recycle nitric acid, nitric oxide in tail gas is converted into nitrogen and then discharged, the powder is led out to a collecting device after being settled, and a uranium trioxide finished product is obtained. The uranium trioxide product prepared by the method is large in specific surface area, high in activity, large in treatment flux, high in tail gas treatment efficiency and good in effect, and can solve the problems that high-flux uranyl nitrate cannot be treated and nitric oxide severely exceeds the standard in the prior art.

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%.

To provide an improved method for recovering uranium.SOLUTION: The present disclosure describes a method of recovering uranium including a continuous ion exchange (CIX) process including a single cycle or a dual cycle CIX process and at least a gradient elution or resin crowding process. The present disclosure also describes an apparatus including a single cycle or dual cycle CIX system and a gradient elution and / or resin crowding system. The present disclosure describes apparatus and process for recovering uranium from a source that involve a combination of one or more CIX processes and at least a gradient elution or resin crowding process to enhance the purity of the obtained uranium. In embodiments, the apparatus and process for recovering uranium described herein include a CIX apparatus and a gradient elution and / or resin crowding system.SELECTED DRAWING: None

A crucible for preparing ceramic particles from metal oxide gel particles, the crucible comprising a tubular graphite shell having an open end, an inner surface, and a seat in the inner surface proximate the open end. A sleeve is lined on the inner surface of the tubular shell. The sleeve has at least one open end and is made of a metal that is chemically inert to metal oxide gel particles. A graphite outer cover removably covers at least one open end of the tubular housing. An inner cap made of a chemically inert metal fits the seat in the inner surface of the tubular housing and is pressed into the seat by the outer cap against the open end of the sleeve. The crucible can be used for preparing ceramic particles from uranium oxide gel particles, and the sleeve and the inner cover can be made of molybdenum, tungsten or alloy thereof.

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.

With the sustainable development of nuclear energy technology, how to efficiently utilize and recycle depleted uranium (mainly 238U) has become a key technical problem of environmental protection and sustainable development. In order to solve the problem, the U3O8 catalyst is prepared by taking uranyl nitrate hexahydrate as a raw material and adopting a simple process of combining dissolution, precipitation and roasting, and is successfully applied to a reaction system for synthesizing azobenzene compounds through catalytic oxidation of aniline and derivatives thereof. According to the method, green conversion and high-value utilization of the depleted uranium waste are achieved, new functional attributes are given to the depleted uranium waste, an efficient catalytic path for conversion from aniline to high-value-added fine chemicals is constructed, and good environmental benefits and economic potential are shown.

Methods of producing peroxides of uranium may include preparing an aqueous solution comprising uranyl ions (UO2+2) in a vessel, bubbling ozone (O3) through the aqueous solution comprising the uranyl ions, and recovering the peroxide of uranium. The peroxide of uranium may comprise at least one of studtite (UO4·4(H2O) or metastudtite (UO4·2(H2O).

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.

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).

Uranium as an indispensable strategic resource in the nuclear energy field of China has self-evident importance. However, along with continuous expansion and deepening of the nuclear energy application field, how to safely and efficiently treat a large amount of depleted uranium (especially 238U which does not cause direct radiation harm to human bodies but still has toxicity similar to that of common heavy metals) generated in the nuclear concentration process has become an urgent subject to be solved in the strategy of environmental protection and sustainable development. The invention aims to explore a new mode for utilizing depleted uranium (238U, which has no radiation injury to human bodies), and particularly, U3O8 prepared by depleting UF6 is used as a raw material, and UO4. 2H2O is prepared by a simple dissolution and reprecipitation method. UO4. 2H2O is used as a catalyst, an organic solvent is used as a reaction solvent, hydrogen peroxide is used as an environment-friendly oxidizing agent, and aniline or derivatives thereof can be catalyzed to be selectively oxidized to generate nitrosobenzene or derivatives thereof. According to the implementation of the method, breakthrough progress is achieved in the field of nuclear waste management, on one hand, the problem of reutilization of the depleted uranium waste is innovatively solved, and a novel and efficient method is provided for resource utilization of the nuclear waste. And on the other hand, by ingeniously converting aniline and derivatives thereof into nitrosobenzene or derivatives thereof with higher economic value, the idea of resource recycling is practiced, gorgeous transformation from wastes to high-added-value products is realized, and the comprehensive utilization rate of resources is greatly improved.

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.

Reasonable utilization of nuclear energy takes an important position in national defense safety consolidation and energy system green transformation processes in China. However, proper management and resource utilization of depleted uranium (especially 238U which does not cause direct radiation threat to a human body) generated along with nuclear energy application become a technical problem to be solved urgently in the fields of environmental protection and sustainable development. Particularly, U3O8 prepared by depleting UF6 is used as a raw material, UO4. 2H2O prepared by a simple dissolution and re-precipitation method is used as a catalyst, and aniline is catalytically oxidized to synthesize 1-diphenyl oxide diazene or derivatives thereof. On the other hand, a bridge from aniline and derivatives thereof to a high value-added chemical namely 1-diphenyl oxide diazene or derivatives thereof is ingeniously erected, and the method shows extraordinary value in environmental, economic and social benefits. The method has extremely important significance from the aspects of environmental friendliness and sustainable development and from the perspective of synthesis of products with higher added values.