53 results about "Actinide" patented technology

The invention relates to the technical field of radionuclide separation, in particular to an Eu (III) / An (III) separation method based on a zirconium phosphonate-based metal organic framework adsorbent, which comprises the following steps: adding the zirconium phosphonate-based metal organic framework adsorbent into an Eu (III) and / or An (III) nuclide-containing aqueous solution with the acid concentration of 0.001-3M to carry out adsorption reaction, and then carrying out solid-liquid separation, wherein the zirconium phosphonate-based metal organic framework adsorbent is a zirconium phosphonate-based metal organic framework material exchanged by metal cations M; the metal cation M is selected from one of alkali metal ions, alkaline earth metal ions or divalent transition metal ions. According to the invention, the interlayer chemical properties of the material are regulated and controlled through ion exchange, so that the adsorbent has ultrahigh selectivity, rapid adsorption kinetics, high adsorption capacity and excellent structural stability on trivalent actinide / simulated nuclides in extremely high-acid, high-salt and intense-radiation environments; the key problems of low separation efficiency and poor stability in a high-acid complex system in the prior art are solved, and the method is suitable for treating high-level liquid waste.

The invention relates to a solvent extraction method for improving a separation factor of trivalent americium and curium, which comprises the following steps: preparing an extraction organic phase solution containing an extraction agent and a synergistic extraction agent, the extraction agent being dithiophosphonic acid with long chain hydrocarbon substituents on two substituted benzene rings on phosphorus atoms; then co-extracting Am (III) / Cm (III) by using an extraction organic phase solution; carrying out selective reverse extraction on Cm (III) by adopting a water-phase solution containing a water-soluble complexing reverse extraction agent; and finally, carrying out primary reverse extraction on the residual Am (III) in the organic phase by using HNO3. According to the method provided by the invention, the separation factor of the trivalent americium and curium is increased from about 4 to more than 10, so that the separation factor of the americium and curium is greatly increased. Compared with the prior art, the extraction agent adopted in the method is higher in stability, the separation factor of americium and curium is high, and separation of trivalent actinide lanthanum and americium and curium can be achieved in one extraction and separation process in a synergistic extraction and complexing reverse extraction mode.

The embodiment of the invention relates to the field of electrolytic production, recovery or refining of metal through a molten liquid electrolysis method, in particular to a cathode assembly beneficial to improving the recovery efficiency of actinide elements and electrolysis equipment used for recovering the actinide elements in chloride salt. The electrolysis apparatus includes an electrolysis vessel forming an electrolysis cell and a cathode mounting interface in communication with the electrolysis cell. The cathode assembly comprises a cathode material, an insulating containing part used for containing the cathode material, a cathode assembly part and a product auxiliary separation part. The cathode assembly part is used for carrying the insulating accommodating part to enter and exit from the electrolytic bath through the cathode mounting interface when being integrally assembled with the insulating accommodating part; and the product auxiliary separation part is used for assisting an electrolytic product formed at the interface of the cathode material melt and the molten salt to leave the interface and settle to the bottom of the insulating accommodating part under the action of self weight. According to the cathode assembly provided by the embodiment of the invention, electrolysis products can be continuously formed, so that the recovery efficiency of actinide elements is improved.

The invention relates to a solid-phase separation material for separating actinide and lanthanum and a preparation method of a main extractant of the solid-phase separation material, the solid-phase separation material is an extraction color layer material adopting an inert support to load a functional reagent, and the main extractant of the functional reagent is phenyl or phenyl derivative substituted disubstituted dithiophosphonic acid. At least one substituent group on a phosphorus atom of the phenyl or phenyl derivative substituted disubstituted dithiophosphonic acid is phenyl or a phenyl derivative. The solid-phase separation material and the main extraction agent thereof provided by the invention have excellent capability of separating trivalent actinide and lanthanide ion groups, and have good chemical stability, radiation resistance and high temperature resistance. Meanwhile, a synergistic extraction reagent is further introduced, so that the adsorption capacity of the trivalent actinide ions is further improved, the pH value of separation operation is reduced, and the difficulty of operation control is further reduced. In addition, the solid-phase separation material for separating the actinide and the lanthanum, provided by the invention, can be stored for a long time and has good reusability.

The invention discloses a separation method of actinide elements. The separation method comprises the following steps: reacting a mixture containing a molten salt phase, a metal phase and a reducing agent; the molten salt phase comprises actinide metal, fission products and molten salt; the actinide metal exists in the form of ions; the metal phase comprises a first metal; the first metal exists in a simple substance form; the reducing agent comprises a first metal and a second metal, and the second metal is lithium; the first metal and the second metal exist in an alloy form; the volume ratio of the molten salt phase to the metal phase is 1: (1-8); the reaction temperature is not higher than 800 DEG C. According to the method, metal lithium in an alloy form is used as a reducing agent for carrying out reduction extraction reaction, and actinide elements and fission products in the molten salt are separated. The separation method can realize efficient separation of actinide elements and fission products in the molten salt, improves the separation coefficient, decontamination coefficient and recovery rate of the actinide elements, and meets the strict requirements of the nuclear industry on separation and purification of the actinide elements.

The present application relates to a kind of preparation methods of pressurized water reactor MOX fuel powder, comprising the following steps: by mixed actinide salt and urea solid preparation mixed actinide eutectic, mixed actinide eutectic is directly obtained by deamination, denitrogenation / decarbonation, reduction, MOX fuel powder;The mixed actinide salt is hexavalent actinide nitrate and tetravalent actinide oxalate;When the mixed actinide salt is uranyl nitrate hexahydrate solid and thorium oxalate hexahydrate solid, the MOX fuel powder is (U x Th 1‑x )O2;When the mixed actinide salt is uranyl nitrate hexahydrate solid and plutonium oxalate hexahydrate solid, the MOX fuel powder is (U x Pu 1‑x )O2.The method provided by the present application has simple raw materials, can greatly simplify the preparation process of MOX fuel powder composed of UO2 and PuO2 (or ThO2), reduce the use of process reagent, and does not produce any process waste liquid, thereby providing a simple, high production efficiency, waste liquid-free, more economical and green MOX fuel powder preparation method.

The application belongs to the technical field of separation and extraction, and particularly relates to a 8C-DAPhen ligand, a preparation method and application thereof. The method is developed from the design of a ligand molecular structure, and a novel cyclic DAPhen ligand is developed through specific structure design. The ligand has a separation factor (SF Eu / Am ) of more than 600 for lanthanide and actinide elements, and exhibits excellent separation performance. The application provides a new perspective for in-depth understanding of lanthanide and actinide separation mechanism, and provides technical support for establishing an advanced spent nuclear fuel reprocessing system and realizing efficient utilization of minor actinide elements.

The application discloses application of an actinide metal catalyst in a carbon dioxide methanation reaction, and the catalyst is prepared by using a carbon source and a metal source to be synthesized through a direct current arc discharge method, and has the advantages of simple process, high preparation efficiency and low cost, and can stably synthesize the actinide metal catalyst with high thermal stability. The prepared actinide metal catalyst has excellent catalytic performance on carbon dioxide hydrogenation conversion, the selectivity of product methane can reach 100%, and the carbon dioxide conversion rate can reach more than 89% under a reaction temperature of 350 DEG C.

Provided in the embodiments of the present application is a preparation method for an MOX fuel powder. An MOX fuel powder comprises a plurality of actinide nuclide oxides. The method comprises the following steps: S1, mixing different actinide nuclide salts with urea to form a mixed actinide eutectic system, the actinide nuclide salts being one of hexavalent actinide nitrates, tetravalent actinide nitrates, hexavalent actinide oxalates and tetravalent actinide oxalates; S2, heating the mixed actinide eutectic system obtained in step S1 to a first preset temperature for performing deamination treatment on the mixed actinide eutectic system, so as to remove the urea therein; and S3, heating the product obtained after the deamination treatment in step S2 to a second preset temperature for performing denitrification and / or decarburization treatment on the product obtained in step S2, so as to obtain a plurality of actinide nuclide oxides. Provided in the embodiments of the present application is the preparation method for an MOX fuel powder, which is simple, does not produce a waste liquid, is more economical and environmentally friendly, and can obtain a completely uniform MOX fuel powder.

The present invention relates to a method for obtaining a bath of molten chloride salts comprising at least one actinide in which the oxidation state and / or degree of oxidation of said at least one actinide are predetermined according to the intended application, comprising the following steps: (a) establishing, for said at least one actinide, its potential-oxoacidity diagram at a given temperature T in a bath of molten chloride salts, (b) from the potential-oxoacidity diagram established in step (a), defining the potential-oxoacidity domain in which said at least one actinide is in the predetermined oxidation state and / or degree of oxidation, (c) selecting at least three different compounds which are different from said at least one actinide setting, at the temperature T, the potential and oxoacidity of the bath of molten chloride salts in the domain defined in step (b), it being possible for one of said at least three compounds to be the chloride ion contained in the bath of molten chloride salts, then (d) introducing, into a bath of molten chloride salts comprising said at least one actinide at the temperature T, said at least three compounds selected in step (c). The present invention relates to the use of various mixtures or compositions in the preparation of a nuclear fuel, during its operation or else during its reprocessing.

The present invention relates to a material comprising actinide oxide-based porous microspheres obtained by an internal gelation process. The invention also relates to the use of this material in the manufacture of nuclear fuels, in particular as a sintering additive in the manufacture of nuclear fuel pellets of the MOX (mixed oxide) type, more particularly as a pore former in place of the pore former conventionally used in said manufacture, i.e. Azodicarbonamide.

The application belongs to the technical field of nuclear fuel cycle and nuclear waste liquid treatment. The application reports a kind of carboxyl-substituted o-phenanthroline water-soluble ligand based on selective masking of part of trivalent actinide under high acidity conditions, and under optimal conditions, the separation coefficient of trivalent europium can reach 120 in 1.5M nitric acid; the separation coefficient of trivalent cerium can reach 4.4. The single crystal structure of the complex is a double metal dimer structure, which guarantees the high acidity tolerance and high selectivity of the ligand. The synthesis of the ligand involved in the application is simple, the extraction efficiency is high, and the extraction mode is novel; the related content will greatly promote the design and performance optimization of water-soluble lanthanum, lanthanum, actinium and actinium separation ligand.

A method of forming an actinide-yttrium hydride includes introducing an yttrium feedstock and an actinide feedstock to a milling / mixing unit. The yttrium and actinide feedstocks are milled and mixed to form a milled mixture. The milled mixture is consolidated in a consolidation unit at a consolidation temperature to form a high-density bulk solid material. The high-density bulk solid material is exposed to hydrogen at a hydriding temperature and a hydriding pressure to form the actinide-yttrium hydride. A composition and system is also disclosed.

The present method entails separating thorium by anion exchange to obtain a solution of a mixture of radium and actinium radionuclides, desorbing the thorium from the anion-exchange resin, separating the actinium and radium by cation exchange, and purifying the actinium of iron salt impurities and organic impurities. The thorium is desorbed using a solution containing a mixture of ammonium acetate and acetic acid. The actinium is purified of iron salts and organic impurities in a column containing two sorbents: a bottom layer comprised of a hydrophilic macroporous resin based on a polymethyl methacrylic acid ether; and a top layer comprised of a sorbent based on di(2-ethylhexyl)phosphoric acid applied to an inert carrier. The total concentration of ammonium acetate and acetic acid used in the desorption of thorium lies in a range of from 0.5 to 4.0 mol / L, and the concentration of hydrogen ions lies in a pH range of from 3.5 to 5.0. For added purification, the column is washed with a solution of hydrochloric acid having a concentration in a range of from 0.015 to 0.025 mol / L. The actinium is eluted using a solution of hydrochloric acid. The method makes it possible to isolate radionuclidically and chemically pure actinium-225 from a mixture containing radium-224, radium-225, thorium-228 and thorium-229 radionuclides.

The embodiment of the invention relates to the field of electrolytic production, recovery or refining of metal through a molten liquid electrolysis method, in particular to a cathode assembly with a stirring function and electrolysis equipment used for recovering actinide elements in chloride salt. The cathode assembly is suitable for the electrolysis equipment, the electrolysis equipment comprises an electrolysis container, and the electrolysis container forms an electrolysis cell and a cathode mounting interface communicated with the electrolysis cell. The cathode assembly comprises a cathode material and an insulating containing part used for containing the cathode material, and further comprises a cathode cover part, a cathode assembly part and a stirring assembly. The cathode cover piece is used for sealing the cathode mounting interface; the cathode assembly part is connected with the cathode cover part and is used for being assembled with the insulating accommodating part or separated from the insulating accommodating part; the stirring assembly is connected with the cathode cover piece and used for stirring the fused salt in the electrolytic bath. According to the cathode assembly provided by the embodiment of the invention, the molten salt in the electrolytic bath can flow to the insulating accommodating part to be in contact with the cathode material melt, so that the recovery efficiency of actinide elements is ensured.

The application provides a DEG-PTD ligand and a preparation method and application thereof, and a method for separating lanthanide elements and actinide elements from waste liquid, and belongs to the technical field of extraction separation. The application starts from the design of a ligand molecular structure, develops a new DEG-PTD ligand by designing the structure of the ligand, and ensures high distribution ratio for Eu, and meanwhile, the introduction of the water-soluble ligand also ensures efficient stripping for Am, 99% of 241Am loaded in the TODGA ligand organic phase can be stripped into the water phase, and the lanthanide-actinide separation factor SF of the DEG-PTD ligand is 1.5*10<4> Eu / Am The separation effect is good above 200. In addition, the ligand has good water solubility and extraction kinetics, and the water solubility is extremely good, and equilibrium can be reached in 5 min. The application provides a new idea for in-depth understanding of lanthanide-actinide separation, and provides important support for sustainable development of nuclear energy, scientific treatment and safe management of existing radioactive waste.

The present invention relates to a material comprising porous microspheres based on actinide oxides, obtained by an internal gelation method. The present invention also relates to the use of this material as a sintering additive in the production of nuclear fuel, particularly in the production of MOX (mixed oxide) type nuclear fuel pellets, and more specifically as a pore-forming agent to replace conventionally used pore-forming agents in said production, namely azodicarbonamides.

The invention relates to a fluorination-electrolysis synergistic process applied to post-treatment of a metal spent fuel, which is characterized in that fluorine gas reacts with the pretreated metal spent fuel under a high-temperature condition, uranium, plutonium and a small amount of fluorinated volatile fission products in the spent fuel are converted into gaseous fluoride to be separated and transferred; and the uranium-plutonium alloy is further prepared by using technologies such as condensation trapping and gaseous fluoride reduction. Meanwhile, fluorinated residues containing a large amount of high-radioactivity and high-heat-release fission elements and minor actinide elements are subjected to electrolytic separation through a villiaumite system, and separation and recovery of the minor actinide elements are achieved. The regeneration and reforming preparation of the metal fuel can be realized on the basis of alloy ingots obtained by a gaseous fluoride reduction process and actinide metal alloys obtained by molten salt electrolysis through injection casting supplementary materials. Compared with the prior art, efficient recovery of uranium and plutonium in the metal spent fuel and deep separation and recovery of minor actinide elements are achieved, the process is simple, and the treatment time is short.

The invention discloses an actinide metal organic framework material and a preparation method and application thereof, and the preparation method comprises the following steps: adding 2, 2 ': 6', 2 ''-terpyridine-4, 4 ', 4 ''-triformic acid and tetravalent actinide salt into a mixed solvent of an organic solvent and water, adding a nitric acid solution into the obtained mixed solution, carrying out solvothermal reaction at 100-140 DEG C, and carrying out vacuum drying to obtain the actinide metal organic framework material. The actinide metal organic framework material is obtained. According to the preparation method, the synthesis process is simple, convenient and efficient, the product yield is high, reactant types are simple, no additional by-product is generated in the reaction process, the prepared actinide metal organic framework material is excellent in structural stability, and an effective solution is provided for efficient capture and long-term safe storage of radioactive iodomethane in nuclear waste.

The invention provides a target design for preparing Ac-225 through irradiation of Ra-226 by an accelerator, and particularly provides an application of a metal niobium film in preparation of medical isotope actinide 225, the niobium metal film is selected as a covering material to cover radioactive radium deposited on a target, and therefore radioactive gas is prevented from polluting a target chamber and the environment. In addition, the annular groove is machined in the target piece base body, so that the upper surface of the target piece is kept flat after the niobium film is added, and installation of the target piece on an accelerator is facilitated.

The quantum bit material of the present disclosure includes a host material of an oxide containing a first element and a second element, or of an oxide containing the second element, a third element, and a fourth element. The host material includes a dopant element. The dopant element is a lanthanide element. The first element is an element of any one of groups 1 to 14. The second element is an element of any one of groups 14 to 17. The third and fourth elements are elements of any one of groups 1, 2, and 4 to 14, lanthanide elements other than the dopant element, or actinoid elements.

This invention discloses a ferrophosphate-based high-entropy glass-ceramic containing a high-entropy monazite phase and its preparation method, used for simultaneously solidifying multiple (minor)actinide nuclides in high-level radioactive waste (HLW). This invention uses five or more simulated (minor) actinide nuclides in equal amounts. Ln ( Ln =La, Ce, Nd, Sm, Eu, Gd, Dy, Ho, Er) simultaneously dissolve into the ceramic crystalline phase, forming a high-entropy monazite phase ( Ln 1 / n )PO4(5≤ n ≤7). The chemical formula of the solidified body is (100- x ()( y Fe2O3-(90- y (P₂O₅-10B₂O₃)-0.5 x P2O5−( x / 2 n ()( Ln 2O3)(mol%), x =0~30mol% y =15~35mol%. The raw materials are mixed according to the formula and then subjected to a one-step melt heat treatment method (melting, nucleation, crystallization) to obtain a solidified iron phosphate-based high-entropy glass-ceramic.

The application discloses an actinide separation method, which comprises the following steps: a mixture of a molten salt phase, a metal phase and a reducing agent is reacted; the molten salt phase comprises an actinide metal, a fission product and a molten salt; the actinide metal exists in an ionic form; the metal phase comprises a first metal; the first metal exists in a simple substance form; the reducing agent comprises the first metal and a second metal, and the second metal is lithium; the first metal and the second metal exist in an alloy form; the volume ratio of the molten salt phase to the metal phase is 1:(1-8); and the reaction temperature is not higher than 800 DEG C. The application separates the actinide from the fission product in the molten salt by using the metallic lithium in the alloy form as the reducing agent to perform a reduction extraction reaction. The separation method can realize efficient separation of the actinide from the fission product in the molten salt, improve the separation coefficient, the decontamination coefficient and the recovery rate of the actinide, and meet the strict requirements of the nuclear industry on the separation and purification of the actinide.

Liquid is irradiated in irradiation device (2) by protons, deuterium nuclei or gamma rays. 226 The radium target, and the resulting product extracted from the irradiated liquid target solution in the first extraction device (6). 225 Origin of A 226 Radium production 225 Actinium. Then, after another irradiation, it was removed. 225 Actinium liquid target solution for further production 225 Actinium. The liquid target solution is preferably circulated through the irradiation device in a closed loop (4) and through the first extraction device (6) in another closed loop (7). The advantage of this method is that the irradiated target solution does not require drying and redissolving to separate radium from the produced actinium, and no further drying and redissolving steps are needed to regenerate the liquid target from the separated radium. Therefore, radium targets can be recovered more efficiently and safely, especially considering the effects of... 226 Radon gas is continuously produced by the decay of radium.

The invention provides a triazole pyridine extractant-hollow tetrazinyl COF composite material as well as a preparation method and application thereof, and relates to the technical field of material preparation. The preparation method comprises the following steps: firstly, taking hydroformylated SiO2 as a hard template, carrying out step-by-step polymerization and alkali etching to obtain a tetraazinyl covalent organic framework (COF) material of an ethylene connecting bond with a hollow structure, and then further introducing an extracting agent molecule with a specific multi-tooth coordination structure into a cavity of the tetraazinyl covalent organic framework (COF) material by utilizing a pressure difference-concentration difference method to obtain the composite material. According to the composite material, the high specific surface area and short mass transfer path of the hollow structure, the strong coordination capability of the tetrazine group and the high selective recognition function of the extraction agent are synergistically integrated, so that the composite material shows high adsorption capacity, fast adsorption kinetics, excellent selectivity and good reusability on thorium ions; the method has important application value in the field of efficient separation and recovery of actinide elements in radioactive wastewater.

The application belongs to the technical field of extraction separation, and particularly relates to an OE-DAPhen ligand, a preparation method and application thereof. The application develops a novel DAPhen ligand through specific structure design from the design of a ligand molecular structure. The separation factor (SF) of the OE-DAPhen ligand for lanthanide and actinide elements is more than 400, and the OE-DAPhen ligand exhibits excellent separation performance. The application provides a new perspective for in-depth understanding of lanthanide and actinide separation mechanism, and provides technical support for establishing an advanced spent nuclear fuel reprocessing system and realizing efficient utilization of minor actinide elements. Eu / Am ).

The invention discloses a preparation method of a calcium defect type pyrochlore ceramic solidified body, which comprises the following steps: adding raw materials CaCO3, CeO2 and TiO2 into absolute ethyl alcohol, carrying out wet grinding uniformly, then carrying out tabletting forming, high-temperature sintering, crushing and grinding, carrying out tabletting again, and continuing high-temperature sintering; and repeatedly grinding, tabletting and sintering for multiple times, taking out the sample after the last sintering, and placing the sample in the air for cooling, thereby obtaining the product. The pressure of 5 to 50 Mpa is adopted for tabletting, and the sintering temperature is 1000 to 1800 DEG C. According to the preparation method, by designing Ca defect components, generation of impure phases is reduced, and a crystalline phase is mainly a pyrochlore phase. In conclusion, the method can be used for treating actinide nuclides and rare earth nuclides in the high-radioactivity waste, and is a novel solidification method.

The disclosure relates to the design and synthesis of selected ligands, dendrimers, polymers and other solid phase substrates for selective chelation of rare earth elements (i.e. lanthanides), and use of those selective ligands for synthesis of resins, polymers and other types of solid supports for separation and recovery of lanthanides from aqueous media. Recovery of critical elements from aqueous media occurs in a simple two-step process: pre-concentration of REE on the adsorbent and recovery by acid elution. The present invention can be used for design of selective ligands immobilized on solid substrates for extraction of various constituents, such as lanthanides, actinides, radionuclides, trace metals, etc., from aqueous media.

The invention relates to an integrated method for trivalent actinides and lanthanides separation and trivalent actinides mutual separation, which comprises: adjusting the pH value of a solution containing trivalent lanthanide and actinide ions to an appropriate value, wherein the actinide ions include americium ions and curium ions; adding an aqueous phase complexing agent, the solution is contacted with the organic phase of an extractant containing alkyl or aryl dithiophosphonic acid and nitrogen containing reagents, and actinides is extracted into the organic phase. The loaded organic phase containing actinide ions was washed by an aqueous phase with a certain pH value. The loaded organic phase is stripped with an aqueous phase of a certain pH value, and the curium ion is stripped into the aqueous phase to achieve the separation of americium and curium. The organic phase is then contacted with dilute nitric acid solution, and the americium ion in the organic phase is stripped into the aqueous phase. The invention can significantly improve the separation factor of americium and curium and reduce the operating pH value. And within the same system, the separation of trivalent actinides and trivalent lanthanides, as well as the mutual separation of trivalent actinides can be realized simultaneously through extraction and stepwise stripping, simplifying the separation process and improving separation efficiency.

The embodiment of the invention relates to the field of electrolytic production, recovery or refining of metal through a molten liquid electrolysis method, in particular to an anode assembly suitable for electrolysis equipment, an anode storage and transfer device and the electrolysis equipment used for recovering actinide elements in chloride salt. The electrolysis equipment comprises an electrolysis container, the electrolysis container comprises a container body forming an electrolytic bath and a container cover body connected with the container body, and the container cover body forms an anode mounting interface communicated with the electrolytic bath. The anode assembly comprises an anode body, an anode cover piece for sealing the anode mounting interface and an anode conductive connecting piece for supplying power to the anode body; wherein the anode conductive connecting piece and the container cover piece are fixed on the container cover body in an insulated manner, and the anode conductive connecting piece is arranged to be detachably and conductively connected with the anode body. According to the anode assembly provided by the embodiment of the invention, the complexity of assembly operation or separation operation of the anode assembly and the electrolysis container can be reduced.