85 results about "Uranium oxide" patented technology

The invention discloses an evolution simulation method and system for uranium dioxide heavy structure behaviors, and relates to the technical field of crystal evolution, and the method comprises the following steps: in a current time step, solving a pore-containing polycrystalline system evolution phase field model to obtain a vacancy concentration field variable and a grain orientation sequence parameter of the current time step; correcting the pore migration speed through the temperature distribution of the next time step; correcting grain orientation sequence parameters of the different air holes based on grain orientation and generation positions of columnar crystals generated by the different air holes; iteration is carried out on the next time step based on the corrected pore migration speed and grain orientation sequence parameters; and carrying out visualization processing on the vacancy concentration field variable and the grain orientation sequence parameter after iteration is finished to obtain the morphology evolution process of the pores and the grains in the uranium dioxide heavy structure process. According to the invention, a columnar crystal generation mechanism in a uranium dioxide heavy structure process is reflected, and simulation of an evolution process of generating columnar crystals with different orientations by migration of one or more pores is realized.

The embodiment of the invention discloses a measuring device and method for the fluorination rate of uranium dioxide. The measuring device comprises a pipeline assembly, a reaction container, a first gas source, a second gas source, a collecting container, a vacuum device, a switch assembly and a heating furnace. The pipeline assembly comprises a first pipeline unit, a second pipeline unit and a third pipeline unit. The reaction container comprises a container body and a container end cover, the container body is provided with a reaction cavity and an opening, and the reaction cavity is used for containing uranium dioxide. The first gas source is communicated with the reaction cavity and is used for supplying fluorine gas; the second gas source is communicated with the reaction cavity and is used for supplying inert gas. The collecting container is communicated with the reaction cavity and is used for collecting uranium hexafluoride. The vacuum device is arranged on the third pipeline unit and can be used for vacuumizing the collecting container and the reaction container. The switch assembly at least can be used for selectively connecting or disconnecting the first pipeline unit, the second pipeline unit and the third pipeline unit.

The invention provides a method for preparing a uranium dioxide single crystal by using a molten salt method and the obtained uranium dioxide single crystal, aiming at solving the problem that the growth of different crystal faces cannot be controlled when the uranium dioxide single crystal is prepared by using an existing fluxing agent method. The method comprises the following steps: in the presence of a reducing agent and a surface active substance, heating U3O8 powder in molten salt to a heat preservation temperature, preserving heat, and then cooling and growing to obtain the uranium dioxide single crystal, wherein the surface active substance is SiO2 (silicon dioxide), GeO2 (germanium dioxide), V2O5 (vanadium pentoxide) or Na2WO4 (titanium dioxide). According to the method, U3O8 is taken as a raw material, a molten salt fluxing agent is combined, the reducing agent is used for reducing U3O8 and growing the UO2 crystal, and the preferred crystal face of UO2 in the nucleation and growth processes is effectively controlled by selecting a proper surface active substance and changing the surface energy of UO2 in the nucleation and growth processes, so that the UO2 single crystal with the preferred crystal face and an ordered structure is obtained.

The invention provides a recovery treatment method of fluoride slag in uranium hexafluoride preparation, belongs to the technical field of nuclear industry waste slag treatment, and solves the technical problems that a product obtained by an existing fluoride slag treatment method is high in metal impurity content and low in uranium recovery rate, secondary fluoride slag is easily generated, and the product cannot be returned to a uranium purification and conversion production line for reuse. The method comprises the following steps: at the temperature of 400-600 DEG C, contacting the fluorinated slag with excessive water vapor to carry out high-temperature hydrolysis reaction, so that UF4 and uranium intermediate fluoride in the fluorinated slag generate uranium oxides U3O8 and UO3 with the fluorine ion content of less than 1mg / g, and filtering to obtain uranium oxides; mixing uranium oxide with a nitric acid solution, dissolving and reacting to obtain a uranyl nitrate solution; the pH of the uranyl nitrate solution is adjusted to 3-4, hydrogen peroxide is added for a precipitation reaction, solid-liquid separation is conducted, and uranium peroxide solid precipitates are obtained. The method is high in uranium recovery rate, low in product impurity content and capable of returning to a uranium purification and conversion production line for reuse.

The invention provides an AEM electrolytic cell for reducing uranyl, and an EE-ER system and method for recovering uranium from offshore spent fuel wastewater. The EE-ER system comprises an AEM electrolytic bath, an overflow blending tank, an acid pool, an ammonia tank, a filter press and a high-temperature calcining furnace which are sequentially connected, a cathode of the AEM electrolytic bath is ZrN / CF, an overflow water outlet of the overflow blending tank is connected with a water treatment system, and a sludge outlet of the overflow blending tank is connected with the acid pool. ZrN / CF is selected as a cathode of the AEM electrolytic cell, high-activity sites are provided to promote electrochemical deposition of uranium in the spent fuel, and 100% degradation of uranium ions can be achieved within two hours. Moreover, the whole EE-ER system realizes online and continuous treatment at a pollution source, the use of an organic solvent is avoided, the purity of the recycled uranium dioxide can exceed 98%, and the recycled uranium dioxide can be directly reused as a nuclear reactor fuel.

The application provides an AEM electrolytic cell for reducing uranyl, an EE-ER system and method for recovering uranium in offshore spent fuel wastewater. The EE-ER system for recovering uranium in offshore spent fuel wastewater comprises an AEM electrolytic cell, an overflow distribution tank, an acid tank, an ammonia tank, a filter press and a high-temperature calcining furnace connected in sequence, wherein the cathode of the AEM electrolytic cell is ZrN / CF, the overflow outlet of the overflow distribution tank is connected with a water treatment system, and the sludge outlet of the overflow distribution tank is connected with the acid tank. The AEM electrolytic cell of the application selects ZrN / CF as the cathode, provides high active sites to promote the electrochemical deposition of uranium in spent fuel, and can achieve 100% degradation of uranium ions within two hours. Moreover, the whole EE-ER system realizes online and continuous treatment at the pollution source, avoids the use of organic solvents, and the purity of the recovered uranium dioxide can exceed 98%, which can be directly reused as nuclear reactor fuel.

The application relates to a preparation method of uranium oxide nanoparticles, and belongs to the technical field of nanometer material preparation, which comprises the following steps: 1, preparation of a precursor solution; 2, template impregnation and drying to obtain a uranium oxide precursor blank; and 3, ceramic treatment, grinding and screening to obtain the uranium oxide nanoparticles. The preparation method of the uranium oxide nanoparticles is simple in experiment process, convenient in operation control and suitable for large-scale production of the nanoparticles. The prepared uranium oxide nanoparticles are good in dispersity, high in phase purity, adjustable in average particle size in the range of 30-500 nm, and can meet the application in the fields of catalysis and nuclear fuel preparation technology.

The invention discloses a uranium oxide solid catalyst, a preparation method and application. The preparation method comprises the following steps: preparing a uranium peroxide precursor by taking hydrogen peroxide as a precipitator, preparing uranium oxide solid alkali through dipping and roasting treatment of alkali precursors such as potassium nitrate, and preparing the high-dispersion-state alkalinity-adjustable uranium oxide solid alkali with a large specific surface area by taking mesoporous silica as a carrier and adopting an ammonia water induced internal hydrolysis method and alkali precursor dipping and roasting treatment. The reaction which cannot be catalyzed by a traditional catalyst is realized.

The invention relates to a process for converting CO2 into methane, wherein hydrogen is contacted with a gaseous feed comprising CO2 in at least one methanation reactor comprising a catalyst bed at a catalyst bed temperature of 160°C to 550°C at a pressure of 0.1 MPa to 1 MPa, wherein the gas hourly space velocity is 10 m 3 / kg / hour to 50 m 3 / kg / hour, and wherein the H2 / CO2 molar ratio is 1 to 8, and wherein the catalyst comprises Ni metal deposited on a substrate made of uranium oxide having the formula UO 2+x , wherein x is 0.01 to 0.6, and wherein the mass content of nickel is 5% to 40% of nickel metal relative to the total mass of the catalyst.

The embodiment of the invention relates to the technical field of binary compounds of nitrogen and metal, in particular to a preparation method of uranium nitride pellet fuel and the uranium nitride pellet fuel. The invention provides a preparation method of uranium nitride pellet fuel, which comprises the following steps: treating molybdenum-niobium metal powder to coat a zirconium coating on the molybdenum-niobium metal powder; preparing uranium oxide particles with uniform particle size; the uranium oxide particles react with ammonia gas, and uranium nitride particles are obtained; molybdenum-niobium alloy powder and uranium nitride particles are mixed, and the molybdenum-niobium alloy powder and the uranium nitride particles are evenly distributed in the obtained mixture; sintering the mixture to enable the obtained pellet to form a compact structure; and molding the pellet to obtain the uranium nitride pellet fuel. The invention further provides the uranium nitride pellet fuel. According to the method for preparing the fuel, on the premise that efficient combustion of the fuel can be guaranteed, the damage resistance of the fuel under the intense radiation condition can be improved, irradiation swelling of the fuel is reduced, the low release property of the fuel is guaranteed, and the application requirement in a space nuclear reactor is fully met.

An installation for the conversion of uranium hexafluoride (UF6) to uranium dioxide (UO2) comprises a hydrolysis reactor (4) for the conversion of UF6 into uranium oxyfluoride powder (UO2F2), a pyrohydrolysis furnace (6) for converting the UO2F2 powder supplied by the reactor (4) into UO2 powder, a supply device (8) comprising reagent injection ducts (10) for the injection of UF6, water vapor or H2, and a control system (16) designed to control the supply device (8) so as to supply at least one of the reagent injection ducts (10) with a neutral gas during a shut-down or start-up phase of the conversion installation.

The invention provides a large-grain uranium dioxide fuel pellet and a preparation method thereof. The preparation method of the large-grain uranium dioxide fuel pellet comprises the following steps: heating uranium dioxide powder in an oxidizing atmosphere to obtain uranium dioxide powder with increased point defect concentration; the uranium dioxide powder with the increased point defect concentration is subjected to external field auxiliary sintering, and the large-grain uranium dioxide fuel pellet is obtained. According to the method, uranium dioxide powder is firstly heated in an oxidizing atmosphere, in the process, the point defect concentration is increased by increasing the oxygen-uranium ratio of the uranium dioxide powder, then the uranium dioxide powder with the increased point defect concentration is subjected to external field auxiliary sintering, under the action of an electric field, charged defects in the powder can be migrated in an accelerated mode, and therefore the uranium dioxide powder with the increased point defect concentration is obtained. After sintering is finished, the point defect concentration can be greatly reduced, and the fuel pellet with the characteristics of large grains and low defect concentration is obtained.

A storage container for uranium oxide powder, comprising: - a drum having a central axis; - an insert comprising several tubes extending in respective directions parallel to the central axis, fastened to the drum and located inside the drum; - neutron absorbing bodies housed inside the tubes.

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.

The application relates to a preparation method of a uranium dioxide-carbon composite porous material, which comprises the following steps: 1) preparing a resorcinol-formaldehyde sol; 2) storing the resorcinol-formaldehyde sol in a constant-temperature sealed state, and performing air drying to obtain a resorcinol-formaldehyde aerogel; 3) performing high-temperature treatment on the resorcinol-formaldehyde aerogel in an argon atmosphere to obtain a porous carbon material; 4) preparing a uranium oxide precursor solution; 5) immersing the porous carbon material in the uranium oxide precursor solution, and then performing drying in air, and repeating the above steps for multiple times to obtain a porous material blank; and 6) performing high-temperature treatment on the porous material blank in a low-pressure environment to obtain the uranium dioxide-carbon composite porous material. The preparation process is simple, and no expensive equipment is needed. The uranium dioxide-carbon composite porous material has high phase purity, good structural stability, and the density, specific surface area and conductivity of the material can be controlled in a large range according to design requirements, so that the application of the material in the fields of catalysis and thermoelectricity can be met.

The invention provides a UAsTe single crystal and a preparation method thereof. The preparation method comprises the following steps: mixing a 238U raw material, an As raw material and a Te raw material according to an atomic ratio of 1: 0.8: (0.75-0.95), putting the mixture into the bottom of a gas-phase transportation pipe, and putting I2 as a gas-phase transportation agent into the bottom of the gas-phase transportation pipe; then, a quartz sealing piece is placed in the gas-phase conveying pipe for sealing, vacuumizing and heat sealing treatment; and putting the gas-phase conveying pipe into a double-temperature-zone tubular furnace, carrying out heat preservation treatment on the gas-phase conveying pipe after heating treatment, observing disappearance of light green in the conveying quartz pipe in the technological process, and finally cooling to room temperature to obtain the UAsTe single crystal. Compared with the prior art in which the stoichiometric ratio of UAsTe is 1: 1: 1, the method has the advantages that the raw material ratio is accurately regulated and controlled, and the double-temperature-zone gas-phase transportation parameters are optimized, so that the generation of toxic by-products such as uranium oxide and arsenic oxide is remarkably inhibited. According to the preparation method disclosed by the invention, the purity of the single crystal is improved by more than 30%, and meanwhile, stable growth of millimeter-scale large-size crystals is realized.

The embodiment of the application relates to the technical field of binary compounds of nitrogen and metal, and particularly relates to a preparation method of uranium nitride particle fuel and the uranium nitride particle fuel. The application provides a preparation method of uranium nitride particle fuel, which comprises the following steps: treating molybdenum-niobium metal powder, so that the molybdenum-niobium alloy powder is coated with a zirconium coating; preparing uranium oxide particles with uniform particle size; reacting the uranium oxide particles with ammonia to obtain uranium nitride particles; mixing the molybdenum-niobium alloy powder and the uranium nitride particles, so that the two are uniformly distributed in the obtained mixture; sintering the mixture, so that the obtained compact is formed into a dense structure; and forming the compact to obtain the uranium nitride particle fuel. The application also provides a uranium nitride particle fuel. The fuel prepared by the method of the application can improve the damage resistance of the fuel under strong radiation conditions, reduce the irradiation swelling of the fuel, ensure the low release of the fuel, and fully meet the application requirements in space nuclear reactors.

The invention discloses a composite uranium dioxide microsphere and a preparation method thereof, and belongs to the technical field of nuclear fuel materials. According to the method disclosed by the invention, the use of a traditional doped element precursor is abandoned, so that the problems of decomposition, oxidation and component control of the precursor are fundamentally avoided, and the process flow is remarkably simplified; through combination of oxygen-controlled calcination and reduction sintering, oxidation of the dopant is effectively prevented, and it is ensured that the dopant stably exists in the UO2 microspheres in a high-thermal-conductivity phase. The prepared composite uranium dioxide microspheres have the excellent characteristics of uniform second phase distribution, strong interface bonding, high thermal conductivity and the like on the premise of keeping high uranium density. The invention provides an innovative solution which is simple in process and remarkable in effect for improving the thermal conductivity of the nuclear fuel, and is suitable for an advanced nuclear reactor fuel system.

A preparation method of low-density U3O8 foam ceramics, which method comprises the following steps: 1) preparing a uranium oxide precursor solution; 2) immersing a foam template in the uranium oxide precursor solution prepared in step 1), and then drying in air; 3) repeating the above steps 1 to 10 times to obtain a uranium oxide foam green body; 4) performing high-temperature treatment on the uranium oxide foam green body obtained in step 3) under a low-pressure environment to obtain a uranium oxide / carbon composite foam ceramic; 5) performing high-temperature sintering on the product obtained in step 4) in an air atmosphere to obtain U3O8 foam ceramics. The present invention can obtain a low-density U3O8 foam ceramic bulk material with a complete structure, high material phase purity, a large regulation range of density and foam pore size, a maximum porosity of the material reaching ~95%, and can meet its applications in fields such as catalysis and laser fusion engineering technology.

The application discloses a method for preparing uranium tetrafluoride based on a reduction hydrogen fluoride reaction, and is applied to a system for preparing uranium tetrafluoride based on a reduction hydrogen fluoride reaction; steps are as follows: 1, thermal decomposition of ammonium fluoride; 2, separation of ammonia and hydrogen fluoride in mixed gas; 3, reduction hydrogen fluoride reaction of uranium oxide; 4, recycling of tail gas. The method has the advantages that: the uranium oxide is reduced and hydrogen fluoridated by adopting a three-stage fluidized bed structure to prepare uranium tetrafluoride, compared with an existing 'uranium purification-uranium conversion' process, continuous control of production can be realized, and the process flow is greatly shortened.

To provide a fusion-driven subcritical reactor that is more efficient and safer. [Solution] A fusion-driven subcritical reactor is a nuclear reactor that starts and operates by amplifying neutrons obtained from nuclear fusion and supplying the amplified neutrons to a reactor that holds subcritical nuclear fuel by external neutrons. It consists of a main neutron source that generates DD neutrons through a nuclear fusion reaction, a cascade member that surrounds the main neutron source, separates fusion / fission neutrons, multiplies the fast fission neutrons generated by nuclear fission, and slows down the fast fission neutrons, and a cascade member that surrounds the cascade member and is mainly made of thorium oxide ( 232 It is equipped with a subcritical main fission blanket made of uranium dioxide (ThO2), and a neutron flattening blanket that has a neutron source located around the outer periphery of the main fission blanket to improve the containment of fission neutrons.

A storage container for uranium oxide powder, comprising a drum having a central axis; an insert comprising several tubes extending in respective directions parallel to the central axis, fastened to the drum and located inside the drum; and neutron absorbing bodies housed inside the tubes.

The invention provides a uranium dioxide pellet as well as a liquid-phase sintering method and application thereof, and relates to application of oxide ceramic in preparation of the uranium dioxide pellet through liquid-phase sintering. And the melting point of the liquid eutectic of the oxide ceramic and uranium dioxide is 1200-1650 DEG C. According to the invention, the low-melting-point oxide ceramic is added in the preparation of the uranium dioxide pellet for liquid-phase sintering, so that the sintering temperature can be reduced while the grain size is maintained; and a higher grain size is obtained at the same sintering temperature, so that more excellent oxidation resistance and creep resistance are obtained.

The present invention proposes a uranium dioxide powder stabilization system, a post-processing system, and a stabilization treatment method. The uranium dioxide powder stabilization system includes a treatment unit, a feed unit, and a stabilized air intake unit. The feed unit is connected to the treatment unit, through which uranium dioxide powder enters the treatment unit. The stabilized air intake unit is connected to the treatment unit, and the stabilized air intake unit introduces an oxygen-containing gas mixture into the treatment unit to stabilize the uranium dioxide powder in the treatment unit. The system is characterized in that it also includes an isolation unit connected to the feed unit. The isolation unit controls the opening and closing of the feed unit and prevents hydrogen and / or water vapor in the feed unit from entering the treatment unit when closed. The uranium dioxide powder stabilization system proposed in this application, by providing an isolation unit, prevents hydrogen and water vapor in the preparation unit from entering the treatment unit, thereby improving the oxygen-uranium ratio of uranium dioxide and achieving continuous treatment.

A uranium recovery device after treatment of uranium-containing fluorine oil waste liquid comprises a water storage unit, and the water storage unit comprises a liquid outlet and a liquid inlet; the plasma torch comprises a nozzle, and the nozzle is arranged in the water storage unit; the plasma torch is externally connected with an air inlet unit; the waste liquid feeding unit is connected with the nozzle; the waste liquid feeding unit inputs uranium-containing fluorine oil waste liquid into a reaction area of the nozzle, and the uranium-containing fluorine oil waste liquid and thermal plasma jet flow generated by the plasma torch are subjected to decomposition reaction; the liquid outlet is externally connected with an inlet of a heat exchange unit, and an outlet of the heat exchange unit is connected with a filtering unit; the filtering unit is used for filtering liquid in the water storage unit after heat exchange of the heat exchange unit and collecting uranium oxide solid particles; and the filtering unit is also used for circulating the filtered liquid back into the water storage unit through the liquid inlet.

The invention discloses a finite element method-based pure uranium thin layer liquid film corrosion one-dimensional model method and device, and the method comprises the steps: constructing a one-dimensional model according to the relative humidity of a target environment and the liquid film thickness under the relative humidity; presetting basic parameters of the one-dimensional model; according to the basic parameters, establishing a reaction process of a gas-liquid interface, a liquid phase, a liquid-solid interface and uranium oxide precipitation in the one-dimensional model; carrying out current distribution initialization on a reaction process of a liquid phase and a liquid-solid interface to obtain an initial current potential distribution condition; taking the initial current potential distribution condition as an initial value, and performing transient solution on each reaction process to obtain a target current potential distribution condition; and according to the parameter value list of the liquid film thickness and the target current potential distribution condition, obtaining a corrosion result of each reaction process. The method can improve the simulation precision and efficiency of the corrosion one-dimensional model, and can be widely applied to the technical field of computer simulation.

The invention discloses a rapid analysis method for determining the content of < 99 > Tc in uranium trioxide powder. The rapid analysis method comprises the following steps: S1, drawing a standard curve; s2, dissolving the sample; s3, sample price adjustment; s4, column separation; s5, ICP-MS measurement is carried out; s6, data processing; according to the method, trace < 99 > Tc is extracted from a high-uranium sample, other radioactive elements interfering technetium-99 measurement are completely removed, separation of a uranium matrix can be completed through one-time column loading and one-time elution under low acidity, and recovery of 99% Tc can be completed through one-time elution. The sample pretreatment method is simple in process operation and easy to master; personnel operation is reduced, and the risk of personnel exposure is reduced.

The invention provides a method for preparing a UO2 fuel pellet with a high fuel consumption structure, which comprises the following steps: (1) preparing a UO2 solution: weighing uranyl and hydrazine hydrate, and adding into water to form a uniform solution; (2) preparing UO2 powder: heating the solution to obtain a uranium dioxide suspension, then washing the suspension with ethanol, centrifuging, and naturally airing the precipitate to obtain the UO2 powder; and (3) core block preparation: mixing the UO2 powder and a pore forming agent, and molding through a hot pressing process.

The present invention relates to an oxide nuclear fuel sintered body and a method for manufacturing the same. In a matrix of a uranium dioxide sintered body used as nuclear fuel for a nuclear power plant, microscopic plate-like precipitates are uniformly precipitated into the nuclear fuel sintered body structure to form annular precipitate clusters distributed therein. More specifically, the present invention aims to provide an oxide nuclear fuel sintered body and a method for manufacturing the same. In order to improve the thermal and physical properties of the uranium dioxide nuclear fuel sintered body, microscopic plate-like precipitates or annular two-dimensional precipitate clusters are uniformly precipitated and distributed within the nuclear fuel sintered body structure to achieve a high creep strain rate and improved thermal conductivity of the sintered body, thereby reducing sintered body-cladding interaction damage and the core temperature of the nuclear fuel in the event of an accident, thereby significantly improving the safety of the nuclear reactor.

Cryogenic granulation, in the conventional way, is carried out on an aqueous suspension comprising one or more oxide powders, which is a preparation that can be a source of splashing and does not completely eliminate the risks.SOLUTION: The invention relates to a process for the manufacture of a powder comprising one or more oxides chosen from uranium oxide UO2, plutonium oxide PuO2 and oxides of minor actinides, the minor actinides being chosen from americium, neptunium and curium, comprising the following steps: a) cryogenic granulation of an aqueous solution comprising cations chosen from uranium-based cations, plutonium-based cations and minor actinide-based cations; b) freeze-drying of the granules obtained in step a); and c) calcining of the granules obtained from step b). An application of the invention is the manufacture of nuclear fuels or blankets filled with minor actinides.SELECTED DRAWING: None