39 results about "Uranium oxide" patented technology

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 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.

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 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 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.

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.

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.

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.

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

A method for preparing a powder including 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, including 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 the granules obtained in a); and c) calcining the granules obtained from b). The method can be used to manufacture nuclear fuels or blankets charged with minor actinide(s).

The application relates to the field of molten salt electrochemistry, in particular to application of normal-pressure micro-plasma discharge in preparation of nano uranium dioxide particles. The normal-pressure plasma discharge can be used for synthesizing nano uranium dioxide particles with high crystallinity in molten salt, the prepared uranium dioxide product has good crystallinity, high purity and excellent oxidation resistance, and the product particle size is between 20-200 nm.

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 invention relates to the technical field of nuclear fuel post-treatment, in particular to a method for separating a metal fission product and uranium oxide through cooperation of chlorination reduction and liquid metal extraction. The method for separating the metal fission product from the uranium oxide through cooperation of chlorination reduction and liquid metal extraction comprises the steps that a mixture containing the uranium oxide and the metal fission product and ammonium chloride are placed in a closed container in a separated mode; heating the closed container to 400-600 DEG C, and keeping the temperature for 3-5 hours to obtain a reaction product; washing and drying the reaction product to obtain a mixture containing uranium dioxide and a noble metal simple substance; and mixing the mixture containing uranium dioxide and the noble metal elementary substance with liquid metal, and carrying out extraction separation to obtain uranium dioxide and the noble metal elementary substance. The method provided by the invention is simple in process flow, good in separation selectivity, low in operation cost and environment-friendly, and valuable elements such as uranium, rare earth and precious metal can be efficiently recycled.

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.

The application belongs to the technical field of seawater uranium extraction catalysts, and particularly relates to a senna leaf catalyst preparation method for seawater uranium extraction and application. The preparation method comprises the following steps: S1: mixing senna leaves with a sacrificial agent at a mass ratio of 1:5-1:1 at room temperature for 2-5 minutes to obtain a senna leaf catalyst; wherein the sacrificial agent is a mixed solution of methanol and NaOH solution; S2: mixing the senna leaf catalyst with seawater, and reacting under irradiation of light with a wavelength of 400-780 nm at 0-35 DEG C for 18-24 days to catalytically obtain uranium peroxide tetrahydrate. The scheme has the advantages of sufficient performance, extremely low cost, extremely simple process and scalable production.

The invention discloses a large-grain uranium dioxide pellet batch manufacturing method and a uranium dioxide pellet, and belongs to the field of nuclear power. The batch manufacturing method of the large-grain uranium dioxide pellets comprises the following steps that UO2 powder, Cr2O3 powder, SiO2 powder, Al2O3 powder and U3O8 are provided and mixed to be uniform through ball milling, raw material powder is obtained, according to the weight ratio, the content of the Cr2O3 powder ranges from 800 ppm to 4000 ppm, the content of the SiO2 powder ranges from 100 ppm to 400 ppm, the content of the Al2O3 powder ranges from 200 ppm to 800 ppm, the content of the U3O8 ranges from 10% to 20%, and the content of the Al2O3 powder is larger than that of the SiO2 powder; the raw materials are pressed into a green body with the density of 4.8 g / cm < 3 >-5.1 g / cm < 3 >, the green body is subjected to heat preservation for 4-8 h at the temperature of 1700-1780 DEG C in a wet hydrogen atmosphere, and the large-grain uranium dioxide pellets are obtained.

The present invention relates to a method for preparing a powder comprising one or more oxides selected from uranium oxide UO2, plutonium oxide PuO2 and oxides of minor actinides selected from americium, neptunium and curium, said one or more oxides comprising one or more oxides selected from uranium oxide UO2, plutonium oxide PuO2 and oxides of minor actinides selected from uranium oxide UO2, plutonium oxide PuO2 and curium. The method comprises the steps of: a) subjecting an aqueous solution comprising cations selected from uranium-based cations, plutonium-based cations and minor actinide-based cations to low temperature granulation; b) freeze-drying the particles obtained in the step a); and c) calcining the particles obtained from b). Applications: the manufacture of nuclear fuels or claddings loaded with minor actinides.

The invention belongs to the technical field of uranium product impurity removal, and particularly relates to an impurity removal method for a uranium dioxide product prepared through hydrogenation reduction. According to the method, firstly, a diluted hydrochloric acid solution is used for washing to remove entrained carbonate impurities, further, an oxalic acid solution is used for washing to deeply remove a small amount of vanadium and molybdenum impurity elements in the product, finally, the high-quality uranium product is obtained through washing, drying and calcining, and the sodium content in the product can be reduced to 0.01% or below; and the content of vanadium and molybdenum can be reduced to 0.0002% or below.

The application discloses preparation and application of a uranium separation hydroxyl-rich cobalt oxide nanosheet, and comprises the following steps: Co(NO3)2.6H2O and C 19 H 42 BrN are added into deionized water, stirred uniformly until completely dissolved, then NaBH4 is added, and stirring is conducted at room temperature by using a magnetic stirrer until air bubbles disappear; the obtained mixed solution is centrifuged, the precipitate is collected, washed by anhydrous ethanol and deionized water in sequence, and vacuum dried to obtain the hydroxyl-rich cobalt oxide nanosheet. The hydroxyl-rich cobalt oxide nanosheet prepared by the application is used as an electrocatalyst for extracting uranium from fluorine-containing uranium wastewater; due to the existence of a large number of hydroxyl groups, the affinity to uranium can be enhanced, the capture capacity to dissociated uranyl ions can be improved, uranium species is grown from an initial uranium monatomic atom to a uranium oxide nanosheet on the surface of the cobalt oxide nanosheet through electric driving fluorine uranium separation, a firm 2O ax -1U-3O eq configuration is formed, and the problem that uranium fluoride complexes are difficult to separate in the prior art is solved.

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.

The invention relates to a carbon-containing uranium and thorium mixed fuel microsphere and a preparation method thereof. The preparation method comprises the following steps: carrying out ball milling on ThO2, UO2 and carbon powder to refine until D50 is less than or equal to 2 microns; thorium dioxide, uranium dioxide and carbon powder are put into a sound wave mixing device according to the mass ratio of 3: 7: 1, deionized water is added till the solid content is 30%-40%, and mixed powder is obtained through dispersion in the argon atmosphere of the sound wave mixing device; adding a binder into the mixed powder, and mixing with deionized water to form a plastic wet material; the plastic wet material forms a continuous strip-shaped material through an extruder, the continuous strip-shaped material is cut into 1.2 mm * 1.2 mm columnar particles through a rotary blade of cutting and rounding equipment, and the columnar particles are rounded into wet-state ball blanks with the diameter of 1.0 + / -0.1 mm through a rounding disc; and in an inert atmosphere, carrying out gradient heating on the wet ball blank to remove the binder and complete a carbon thermal reduction reaction to finally obtain the UC2-ThC2 composite fuel microspheres. The technical problems that uranium and thorium fuel microspheres are non-uniform in mixing, poor in forming quality and low in sintering performance are solved by integrating the technologies of acoustic dispersion, plastic wet material rolling and gradient sintering.

The embodiment of the present application discloses a kind of measuring device and measuring method of fluorination rate of uranium dioxide, measuring device includes pipeline assembly, reaction container, first gas source, second gas source, collection container, vacuum device, switch assembly and heating furnace.Pipeline assembly includes first pipeline unit, second pipeline unit and third pipeline unit.Reaction container includes container body and container end cover, container body is provided with reaction cavity and opening, reaction cavity is used to accommodate uranium dioxide.First gas source is communicated with reaction cavity, and first gas source is used to supply fluorine gas.Second gas source is communicated with reaction cavity, and second gas source is used to supply inert gas.Collection container is communicated with reaction cavity, and collection container is used to collect uranium hexafluoride.Vacuum device is arranged in third pipeline unit, and vacuum device can be used to vacuumize collection container and reaction container.Switch assembly can be used at least selectively to open or close first pipeline unit, second pipeline unit and third pipeline unit.

The embodiment of the present application relates to the technical field of reactor fuel, and particularly relates to a method for determining sintering process parameters of substoichiometric uranium dioxide, which mainly comprises the following steps: determining average grain sizes and porosities of a plurality of different substoichiometric uranium dioxide according to the plurality of different substoichiometric uranium dioxide, and determining average grain sizes and porosities of substoichiometric uranium dioxide corresponding to optimal thermal conductivity according to the optimal thermal conductivity; and determining the sintering process parameters of the substoichiometric uranium dioxide according to the determined average grain sizes and porosities. The method accurately reflects the influence of the substoichiometric dose on the sintering process of the uranium dioxide, and can accurately determine the substoichiometric ratio corresponding to the optimal thermal conductivity and the corresponding sintering process parameters.

The invention relates to the technical field of uranium extraction from seawater and wastewater, in particular to an alternating-current electrochemical system and method for extracting an anionic uranyl complex. According to the method, a high-power and high-efficiency industrial alternating-current power supply is adopted, and a programmable logic controller is integrated to generate and accurately control required asymmetric alternating-current waveforms (voltage, frequency and duty ratio), so that a uranium solution is in full contact with a functionalized electrode; in a forward bias period, uranyl anion complex ions are directionally enriched to an electrode interface; in a negative bias period, uranyl anion complex ions are subjected to in-situ electrochemical reduction into insoluble uranium oxides and are fixed. The process effectively overcomes the inherent defects that a traditional adsorption material is low in capacity, slow in dynamics and poor in selectivity, and a brand new technical scheme with wide application prospects is provided for efficient and high-selectivity extraction of uranium in seawater and wastewater.