19results 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.

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.

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.

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.

: 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).

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.