26results about "Uranium dioxide" 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.

A reactor (1) for hydrolysis of uranium hexafluoride comprises a tubular injector (9) comprising first (11), second (13) and third (15) concentric fluid circulation ducts intended to be connected respectively with a source of UF6, a source of inert gas and a source of water vapor. The tubular injector (9) is obtained by additive manufacturing.

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 discloses a filtering and recycling method and system for concentrated slurry of uranyl carbonate and ammonium diuranate in crystallization mother liquor, and relates to the technical field of uranium purification, the recycling method comprises the following steps: S1, adopting at least two solid-liquid separation devices which are arranged in parallel to carry out alternate circulating filtering treatment on tower bottoms of crystallization mother liquor containing uranyl carbonate and / or ammonium diuranate, filtering to obtain filtered clear liquid and an ammonia-containing filter cake; s2, carrying out air stripping treatment on the ammonia-containing filter cake obtained in S1 to separate ammonia-containing gas and obtain a solid material with low ammonia content; s3, carrying out absorption treatment on the ammonia-containing gas separated in S2 to obtain an ammonium-containing product; and S4, carrying out pyrolysis treatment on the solid material with low ammonia content obtained in S2 to obtain a uranium-containing product. Through cooperation of double-machine alternate circulating filtration, air stripping separation, absorption recovery and pyrolysis conversion, the technical problems of ammonia nitrogen pollution transfer, dependence on manual operation, long flow and high water content of a filter cake are solved.

Disclosed is a micro nuclear reactor that includes a core filled with nuclear fuel and moderator which are formed of particles, a heat pipe inserted in the core and transferring outwards heat generated by a nuclear reaction, and a power converter receiving heat from a condenser of the heat pipe and converting thermal energy into electrical energy.

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 method for preparing, in the internal volume of at least one channel, a monolithic support on which uranyl cations are immobilised. The method comprises: (a) activating the inner surface of the channel(s); (b) introducing, into the internal volume of the channel(s), a polymerisation solution comprising: a monomer comprising a phosphate group, at least one crosslinking agent, several solvents, and a radical polymerisation initiator; (c) polymerising the polymerisation solution; (d) rinsing the monolithic support obtained in step (c); and (e) contacting the monolithic support previously rinsed, with a solution comprising uranyl cations. A method for capturing proteins that selectively bind uranium by means of a monolithic support prepared by the above-mentioned method, as well as to a method for recovering proteins that selectively bind uranium with the capture method.

The application discloses a method for preparing UO2 from U3O8, which adopts activated carbon to reduce U3O8, shortens the intermolecular distance between the activated carbon and U3O8 by tabletting, and prevents the entrainment of gas to the powder in the reaction process; the porous structure of the activated carbon greatly increases the contact area with U3O8, improves the reduction efficiency, and the activated carbon is removed in the final product UO2, so that the purity of UO2 is ensured; the method is simple, short in period, and high in safety, and is suitable for large-scale reduction of U3O8 to prepare UO2.

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

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.

The invention provides a method for separating and recovering useful components in rare earth production radioactive wastewater, and relates to the technical field of chemical separation. The method comprises the following steps: adding oxalic acid, controlling the pH value, filtering, dropwise adding sulfuric acid into the filtrate, stirring and precipitating at the set pH value, filtering to obtain filtrate I and calcium sulfate dihydrate whiskers, and heating and dehydrating to obtain anhydrous calcium sulfate whiskers; adding a NaOH solution into the filtrate I, controlling the pH value, and filtering to obtain thorium hydroxide and filtrate II; and adding sulfuric acid into the filtrate II to form acidified waste liquid, and adding resin or aerogel to adsorb and separate uranium. Through the procedures of fractional precipitation, hydrolysis, roasting, acidification, adsorption, elution, denitration reduction and the like, useful components Ca < 2 + >, U6 < + > and Th < 4 + > in the wastewater are effectively separated, and high-value products of calcium sulfate whiskers, thorium oxide and uranium oxide are obtained.

A conversion process for converting uranium hexafluoride into uranium dioxide includes the steps of hydrolysis of UF6 to uranium oxyfluoride (UO2F2) in a hydrolysis reactor (4) by reaction between gaseous UF6 and dry water vapour injected into the reactor (4), pyrohydrolysis of UO2F2 to UO2 in a pyrohydrolysis furnace (6) by reacting UO2F2 with dry steam and gaseous hydrogen (H2) injected into the furnace (6), extracting excess gas in the reactor (4) via a collecting device (50) comprising several filters (52), periodically cleaning the filters (52) by injecting a neutral gas into the filters (52) from the outside to the inside of the reactor (4) to remove powder stuck on the filters (52), and measuring the relative pressure in the reactor (4). The conversion method further includes carrying out point cleaning of the filters (52) when the relative pressure in the reactor (4) exceeds a predetermined point cleaning threshold.

: A method of manufacturing particles of UO2 for a porous matrix of a target for use in the manufacture of 99Mo, comprising: infiltrating a solution of uranyl nitrate into a polymer template (62); either (i) introducing an alkali chemical to the uranyl nitrate infiltrated polymer template (64), causing precipitation of uranium oxide / hydroxide, and converting the uranium oxide / hydroxide to U3O8 (68) and concurrently removing the polymer template (70), by heating the infiltrated polymer template; or (ii) converting the uranyl nitrate to U3O8 (66) and concurrently removing the polymer template (74), by heating the infiltrated polymer template; and reducing the U3O8 to UO2 via heating in a reducing atmosphere (72).

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 shutdown or start-up phase of the conversion installation.

Disclosed is a dry aftertreatment method for spent fuel employing plasma. The method comprises the following steps: S1, performing treatment on spent fuel which has undergone cladding removal to acqui

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.

Disclosed herein are methods and systems for the recovery of uranium from used nuclear fuel containing nuclear metal using nonaqueous chemistry without electrometallurgy techniques, such as pyroprocessing. The methods and systems use chloride-based volatility (CBV) as the basis for uranium recovery from the used nuclear fuel containing uranium metal. The solid used nuclear fuel containing uranium metal is maintained under UCl4 sublimating conditions in a chlorinating environment for a period of time sufficient to convert at least some uranium metal into UCl4.

The invention discloses a Ce / Zr bimetal doped UiO-67 material for nitric oxide detection and a method and application thereof, and belongs to the technical field of functional materials and electrochemical sensing. The preparation method disclosed by the invention comprises the following steps: dissolving 4, 4 '-biphenyldicarboxylic acid in a polar solvent, performing ultrasonic dispersion, adding a mixed solution of cerium salt and zirconium salt, and stirring to obtain a precursor solution; adding a carboxylic acid conditioning agent into the precursor solution, and then carrying out solvothermal reaction to obtain a reaction product; and carrying out post-treatment on the reaction product to obtain the Ce / Zr bimetal doped UiO-67 material for nitric oxide detection. The method solves the problems of insufficient activity and poor stability of the traditional MOF material in electrochemical sensing.

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 a method for preparing, in the internal space of at least one channel, a monolithic support on which uranyl cations are immobilised. According to the invention, the method comprises: (a) activating the internal surface of the channel(s); (b) introducing, into the internal space of the channel(s), a polymerisation solution comprising: - a monomer comprising a phosphate group, - at least one cross-linking agent, - a plurality of solvents, and - a radical polymerisation initiator; (c) polymerising the polymerisation solution; (d) rinsing the monolithic support produced in step (c); and (e) contacting the rinsed monolithic support with a solution comprising uranyl cations. The invention also relates to a method for capturing proteins that selectively bind uranium by means of a monolithic support prepared using the aforementioned method as well as a method for recovering proteins that selectively bind uranium in accordance with the capturing method.

The present invention relates to a process for the production of a metal or an alloy comprising the metal from a metal precursor. The process comprises reacting the metal precursor with a reducing agent in a self-sustaining exothermic reaction and separating the metal or the alloy comprising the metal from a slag produced during the reaction. The invention also relates to the use of an intermetallic compound defined by the chemical formula Ca8(Al 3‑x M x ) as a reducing agent in the reduction of a metal precursor to produce a metal or a metal alloy.

The application provides a selective dissolution and separation method of lanthanide elements in spent fuel. The method comprises the following steps: adding a mixture of lanthanide oxide and uranium dioxide into an aqueous solution of hydroxyethylidene tripepiperazine ammonium hydrochloride or an aqueous solution of butyl hydroxyethylidene tripepiperazine ammonium hydrochloride, and heating and stirring to selectively dissolve the lanthanide oxide. Then, centrifugal separation is performed, the dissolved lanthanide oxide is in the solution, and the undissolved actinide compound uranium dioxide solid is recovered, so that the lanthanide elements and the actinide elements in the spent fuel are efficiently separated. The application is based on the good dissolution behavior of the hydroxyethylidene tripepiperazine ammonium hydrochloride compound with radiation resistance to the lanthanide elements, and applies the selective dissolution and separation technology to the spent fuel reprocessing. Compared with the traditional solvent extraction method, the method avoids the use of organic solvents and a large amount of waste liquid, the used alkaloid hydroxyethylidene tripepiperazine ammonium hydrochloride compound is non-toxic and harmless, green and environment-friendly, low in cost, strong in radiation resistance, simple in process flow, and easy to be applied in industrialization.

Methods and systems (300) for separation of thorium from uranium and their decay products are provided. The method comprises combining a nuclear fuel feedstock (102) comprising thorium and uranium with a first acid (104, 204) to form a first solution. The first solution is contacted 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 uranium or thorium from the second solution to form a precipitate (110, 210). The precipitate (110, 210) is separated from the second solution (106, 206).

A micro nuclear reactor includes a core filled with nuclear fuel particles mixed with moderator particles. A heat pipe extends into the core and transfers heat generated by the core. A power converter receives heat from the heat pipe and converts thermal energy into electrical energy. A valve is configured to open and close an outlet located at a lower portion of the core. When the valve is open the nuclear fuel and the moderator are discharged by gravity from the core through the outlet.