108 results about "Actinide" patented technology

The invention discloses an NZP-high-entropy (Ln1 / n) PO4 monazite type multiphase ceramic solidified body and a preparation method thereof, the chemical general formula of the NZP-high-entropy (Ln1 / n) PO4 monazite type multiphase ceramic solidified body is (1-x) Sr0. 5Zr2 (PO4) 3-x (Ln1 / n) PO4, x represents molar percentage, the value range is 0.1-0.9, x represents molar percentage, and the value range is 0.1-0.9. The NZP-high-entropy (Ln1 / n) PO4 monazite type multiphase ceramic solidified body is NZP-high-entropy (Ln1 / n) PO4 monazite type multiphase ceramic solidified body. N represents the number of element types of Ln lattice sites in the high-entropy (Ln1 / n) PO4 monazite type, and n is an integer and is larger than or equal to 5 and smaller than or equal to 7; the Ln element types comprise La, Nd, Sm, Eu, Gd, Ho, Er, Sr0. 5Zr2 (PO4) 3 and (Ln1 / n) PO4, and the molar ratio of the Ln element types to the Sr0. 5Zr2 (PO4) 3 to the (Ln1 / n) PO4 can be freely regulated and controlled. The NZP-high-entropy (Ln1 / n) PO4 monazite type multiphase ceramic solidified body adopts an in-situ preparation process, and integration of crystal phase synthesis and sintering densification can be realized. The preparation method comprises the following steps: uniformly mixing the raw materials, pretreating, adding a sintering aid, mixing again, granulating and aging the obtained powder, molding, removing plastic, and sintering to obtain the NZP-high-entropy (Ln1 / n) PO4 monazite type composite ceramic solidified body. The preparation process disclosed by the invention has the advantages of flexible composition design, capability of simultaneously curing various fission products and (sub) actinide series nuclides, short preparation period, low production energy consumption, high product compactness and the like.

Disclosed herein are methods and systems for the use of extraction chromatography resins to concentrate Ac from dilute solutions that may contain Th isotopes and / or other isotopes, such as Ra. As disclosed herein, at least some extraction chromatography resins strongly retain trivalent actinides between 4 and 6 M HNO3 that can be recovered in good yields with small volumes of dilute acids. These small volumes of acid eluents result in significant concentration of the solution of the trivalent / certain tetravalent ions, such that they can be further processed without requiring lengthy evaporation steps that are utilized in standard Ac concentration methods.

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.

A nickel-based brazing filler metal (BFM) consisting essentially of: from about 12 wt% to 16 wt% chromium (Cr); up to about 5 wt% Fe; up to about 2 wt% Mo; about 1 wt% to 5 wt% of tungsten (W); about 3 wt% to 6 wt% of aluminum (Al), about 3 wt% to 6 wt% of tantalum (Ta); hafnium (Hf) from about 0.05 wt% to 1.5 wt%, carbon (C) from about 0.01 wt% to 0.1 wt%; zirconium (Zr) from about 0.005 wt% to 0.05 wt%; from about 0.005 wt% to 0.05 wt% of silicon (Si); a sum of about 0.01 to 0.05 wt% of a rare earth (e.g., Sc, yttrium (Y), actinides, and lanthanides); boron (B) from about 1.7 wt% to 2.6 wt%; the balance being essentially nickel (Ni) and unavoidable impurities. As part of a production and repair process for hot gas channel components, such as, but not limited to, blades, vanes and heat shields in gas turbines, for high strength transient liquid phase bonding (TLB) brazing of gamma'strengthened nickel-based superalloys by capillary action in the range of 1180 DEG C to 1220 DEG C. In particular, it is used for closing the top cover on the blade.

The present invention relates to a method for evaluating the solidification treatment of radioactive waste and a method for evaluating the feasibility of alternative nuclides. The method for evaluating the feasibility of alternative nuclides includes: obtaining at least one actinide nuclide in the solidification treatment of radioactive waste; preparing a single crystal of trivalent coordination salt of the actinide nuclide in a simulated coordination environment for the solidification treatment of radioactive waste; preparing a single crystal of trivalent coordination salt of the lanthanide nuclide in the same simulated coordination environment; comparing the crystal structures of the single crystals of trivalent coordination salts of the actinide nuclide and the lanthanide nuclide; and obtaining the evaluation result of the feasibility evaluation according to the comparison result. By preparing single crystals of trivalent coordination salts of actinide and lanthanide nuclides under the same conditions, comparing the crystal structures of the single crystals, and based on the comparison result, it can be determined whether the lanthanide nuclide can replace the actinide nuclide for use in the research on the evaluation of the solidification treatment of radioactive waste.

The invention relates to a nickel-based superalloy comprising (in% by weight): 6.0% to 12.0% cobalt (Co); from 6.0% to 12.0% of iron (Fe); chromium (Cr) from 14.0% to 18%; up to 2.0% molybdenum (Mo); 2.0% to 5.0% of tungsten (W); from 4.0% to 5.5% of aluminum (Al); from 2.0% to 3.5% tantalum (Ta); up to 0.2% titanium (Ti); from 0.03% to 0.3% of carbon (C); from 0.005% to 0.03% of boron (B); from 0.005% to 0.03% of zirconium (Zr); from 0.005% to 0.5% of silicon (Si); 0.002% to 0.2% of a total content of Y, Sc, lanthanides and actinides; from 0.05% to 3.0% hafnium (Hf); up to 2.0% manganese (Mn); nickel (Ni) and unavoidable impurities.

Compositions, systems, and methods are provided for detecting target particles, such as target actinide particles. The composition includes 2‑(5‑Bromo‑2‑pyridylazo)‑5‑(diethylamino)phenol (Br‑PADAP), 1‑(2‑Pyridylazo)‑2‑naphthol (PAN), and iron in effective amounts such that on a condition that the composition is in contact with the target actinide particle, a colorimetric output is produced.

A fuel design for high burnup in water-cooled reactors is disclosed. A nuclear fuel rod comprising an outer zone of annular fuel pellets composed of actinides and fissile material and an interior comprising a low density, high open porosity ceramic foam of zirconia or magnesia is used. Use of annular pellets lowers fuel temperature, reduces fission gas release and fuel swelling. Foam with low crushing strength is used to accommodate fuel swelling. Fresh fuel reactivity is controlled by burnable absorbers such as gadolinia, incorporated as sintered bodies within the foam or as part of the fuel composition or both. The placement of burnable absorber in the foam reduces the contact area with surrounding media, mitigating diffusional transport and preserving its self-shielding characteristic. A novel two-piece foam holder is proposed to insert the ceramic foam.

The present invention provides an N,N-dihydrocarbyl amide carboxylic acid, preparation method therefor and use thereof. The N,N-dihydrocarbyl amide carboxylic acid can be used as an extractant for enriching rare earth elements from raw materials containing low-concentration rare earth elements, separating and purifying yttrium element from mixed rare earth raw material, and separating elements such as aluminum, iron, radioactive thorium, radioactive uranium and actinide from mixed rare earth raw material, etc. The compound can be synthesized in a simple and cost-efficient way. As an extractant, it has good chemical stability and can withstand strong acid and strong alkali without decomposition.

The present disclosure is generally directed to methods for forming anhydrous rare earth and actinide phosphate minerals. In some embodiments of the present disclosure, the method generally comprises subjecting rare earth or actinide phosphates or salts thereof to microwave radiation.

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.

To provide a method for treating a disease Provided herein are methods for treating a subject in need of treatment, comprising administering to the subject in need of such treatment a therapeutically effective amount of one or more pharmaceutical compositions comprising a 1,2-HOPO chelator and / or a 3,2-HOPO chelator. The subject may be one who has been or may be exposed to one or more known or unknown actinides and / or lanthanides, or mixtures thereof.

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 present invention relates to a process for producing a solid Ziegler-Natta catalyst component in the form of solid particles having a median particle size (D50vol) of 5 to 500 μm and the process comprising steps I. providing a solution of a Group 2 metal dihalide (IUPAC, Nomenclature of Inorganic Chemistry, 2005) by dissolving a solid Group 2 metal dihalide in an alcohol mixture comprising at least a monohydric alcohol (A1) of formula ROM, where R is selected from a hydrocarbyl group of 3 to 16 C atoms and an alcohol (A2) comprising in addition to the hydroxyl group another oxygen containing functional group not being a hydroxyl group, contacting the solution of the Group 2 metal dihalide of step I with a compound in a liquid form of a transition metal of Group 4 to 10, or of a lanthanide or actinide, preferably a transition metal of Group 4 to 6 of Periodic Table (IUPAC, Nomenclature of Inorganic Chemistry, 2005), and III. recovering the solid catalyst component, wherein the amount of Group 2 metal originating from Group 2 metal dihalide constitutes 100% of the whole amount of the Group 2 metal used in the process for producing the solid Ziegler-Natta catalyst component.

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 invention discloses a device and a method for preparing a high-purity tensile sample for part of actinide elements. The device comprises a cathode, a crucible, a gas supply system and an atmosphere control system, the crucible is located in a cavity of the electric arc melting furnace, and the cathode extends into the cavity of the electric arc melting furnace. According to the method, the actinide element purification and tensile sample preparation are performed by adopting an electric arc melting technology, the preparation process is simple, safe and reliable, and the prepared sample is high in purity and stable in performance. The problem that casting cannot be carried out after viscous cooling is solved, and powerful technical support is provided for application of actinide elements and alloys thereof in the fields of nuclear fuel, nuclear reactors and the like.

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.

This invention relates to a dual-column cyclic preparation method. 239 The apparatus and method for Np tracers belong to the field of actinide isotope separation and preparation technology, and solve the problem of... 239 The technical challenges of low purity and unstable recovery rates in Np tracer preparation are addressed by this device, which includes a temperature controller, a radiation shield, a storage tank unit, a chromatography separation unit, a waste liquid collector, a product collector, valves, and a transfer pump. Liquid from the storage tank unit is pumped into the chromatography separation unit, and liquid exiting the chromatography separation unit enters either the waste liquid collector or the product collector. The chromatography separation unit includes a TODGA chromatography adsorption column and LN or P... 204 or P 507 Chromatographic adsorption column. The method includes preparing a feed solution, washing solution, eluent, and chromatographic adsorption column before separation; adsorption, washing, and elution steps for neptunium. This invention is used for preparing... 239 Np tracer.

The present invention relates to a method for forming a thin film, and more particularly, to a method for forming a thin film comprising steps of:i) adsorbing a growth inhibitor for forming a thin film on a surface of a substrate; and ii) adsorbing a metal film precursor, metal oxide film precursor, metal nitride film precursor or silicon nitride film precursor on a surface of a substrate on which the growth inhibitor is adsorbed, wherein the growth inhibitor for forming a thin film is represented by Chemical Formula 1 below, and the metal is at least one selected from a group consisting of tungsten, cobalt, chrome, aluminum, hafnium, vanadium, niobium, germanium, lanthanide, actinoids, gallium, tantalum, zirconium, ruthenium, copper, titanium, is nickel, iridium and molybdenum.AnBmXo  [Chemical Formula 1]wherein A is carbon or silicon, B is hydrogen or a C1-C3 alkyl, X is a halogen, n is an integer of 1 to 15, o is an integer of 1 or more, and m is 0 to 2n+1.According to the present invention, it is possible to suppress side reactions to appropriately lower a thin film growth rate and remove process byproducts in the thin film, thereby preventing corrosion or deterioration and greatly improving step coverage and thickness uniformity of a thin film, even when the thin film is formed on a substrate having a complex structure.

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 invention relates to an organic-phase actinide element standard solution as well as a preparation method and application thereof. The preparation method comprises the following steps: S1, preparing a water-phase actinide element standard stock solution; s2, TBP + TRPO-kerosene is prepared; s3, TBP + TRPO-kerosene is adopted for extracting the aqueous-phase actinide element standard solution stock solution, an organic-phase actinide element standard solution and raffinate are obtained, and the process is completed. Compared with the prior art, a TBP + TRPO-kerosene extraction technology is adopted to extract the aqueous-phase actinide element standard stock solution to obtain the organic-phase actinide element standard solution, and the method is accurate and reliable through theoretical and experimental verification, and can be applied to the field of nuclear fuel closed circulation as a preparation method of the organic-phase actinide element standard solution.

The present application relates to a kind of trivalent actinide and rare earth fission fragment nuclide separation device and separation method, belong to actinide and fission product separation technical field, solve the technical problem that separation exists decontamination low, recovery rate is low, its separation device includes liquid storage tank unit, chromatography separation unit, waste liquid collector, product collection unit, delivery pump.Liquid of liquid storage tank unit enters chromatography separation unit by delivery pump, the liquid that chromatography separation unit flows enters waste liquid collector or product collection unit.Liquid storage tank unit, chromatography separation unit, product collection unit, waste liquid collector are connected by pipeline, on pipeline, set valve.Its separation method includes preparation before separation, the steps of adsorption, washing, leaching of trivalent actinide and rare earth fission fragment nuclide.The present application is used for the separation of trivalent actinide and rare earth fission fragment nuclide.

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.