449 results about "Uranium" patented technology

The invention discloses a composite electrode material for removing uranium through photoelectrocatalysis and a preparation method and application of the composite electrode material, and belongs to the technical field of catalytic materials and wastewater treatment.The preparation method of the composite electrode material comprises the following steps that a ZIF-8 material, copper salt and ferric salt are mixed, and a first metal composite material is prepared; reacting the first metal composite material with a KOH solution to prepare a second metal composite material; carrying out carbonization treatment on the second metal composite material, and then carrying out acid treatment by using acid to obtain a bimetallic atom catalytic material; the preparation method comprises the following steps: dispersing a bimetallic atom catalytic material and a g-C3N4 material in a mixed solution containing phytic acid and a Nafion solution to obtain catalyst ink; and coating the catalyst ink on a substrate, and carrying out illumination drying to obtain the composite electrode material. According to the composite electrode material, under the synergistic effect of the bimetallic single atomic cluster catalyst and g-C3N4, respective advantages are fully exerted, and efficient deposition removal of uranium is achieved.

The embodiment of the invention relates to the field of exploration or detection by applying nuclear radiation, in particular to a method for determining a uranium metallogenic area of a to-be-surveyed area, which comprises the following steps of: determining natural gamma logging data and energy spectrum logging uranium content data of the to-be-surveyed area; determining the positions of a plurality of drill holes of the to-be-surveyed area; obtaining a rock core sample of one of the determined drill holes; determining uranium contents at different depths according to the rock core sample; determining the relationship between the uranium content and the energy spectrum logging uranium content data of the drill hole and the relationship between the energy spectrum logging uranium content data and the natural gamma logging data so as to determine the relationship between the uranium content and the natural gamma logging data; and determining a natural gamma logging threshold value of the uranium metallogenic area so as to determine the uranium metallogenic area of the area to be surveyed. According to the method provided by the embodiment of the invention, the natural gamma logging threshold value of the uranium metallogenic area can be accurately determined, so that the uranium metallogenic area of the area to be surveyed can be accurately determined, and the exploration cost is greatly reduced.

A uranium-based phosphor selected from (i) phosphors having formula I or II:where the phosphor having formula I or II is doped with an activator ion including Pr3+, Sm3+, or mixtures thereof, where 0≤a≤1, 0≤b≤1, 0.75≤x≤1.25, 0.75≤y≤1.25, 0.75≤z≤1.25, 2.5≤p≤3.5, 1.75≤q≤2.25, and 3.5≤r≤4.5; (ii) a phosphor having formula I or II, where the phosphor having formula I or II is doped with an activator ion selected from Eu3+, Pr3+, Sm3+, and mixtures thereof; and a counter ion comprising one or more alkali metal ions; and (iii) phosphors having formula IIIwhere the phosphors having formula III are doped with an activator ion selected from: Eu3+, Pr3+, Sm3+, and mixtures thereof, where A is Li, Na, K, Rb, Cs, or a combination thereof.

The invention discloses a preparation method of a porous few-layer carbon nitride / amidoxime-based seawater uranium extraction composite material with a multilevel structure, and relates to the field of seawater uranium extraction materials. The composite material is a porous few-layer carbon nitride material with high photocatalytic performance obtained by repeatedly calcining melamine at a high temperature; polyacrylonitrile is subjected to amidoxime functionalization, and a polyamidoxime solution is obtained; uniformly dispersing porous few-layer carbon nitride in the polyamidoxime solution, and adding a sacrificial pore-forming agent polyvinylpyrrolidone at the same time to obtain a spinning solution; and finally, preparing the porous few-layer carbon nitride / amidoxime-based seawater uranium extraction composite material with high adsorption performance by utilizing a one-step electrostatic spinning technology. The obtained composite material has excellent uranium adsorption performance and adsorption rate and can be recycled, the preparation method is simple in process and free of pollution to the environment, and the prepared composite material has excellent uranium adsorption performance and adsorption rate, can be recycled and has wide application prospects in the field of uranium extraction from seawater.

Disclosed herein are methods and systems for the recovery of uranium from used fuel salts using nonaqueous chemistry without electrometallurgy techniques. The used fuel salts are maintained in a molten state in a chlorination chamber, and are then chlorinated for a period of time sufficient to convert at least some UCl3 into a different uranium chloride, thereby generating a uranium-containing gas phase. The uranium-containing gas phase is separated from the resulting residue, and is condensed for collection.

According to the COF nanofiber membrane, the preparation method thereof and the application of the COF nanofiber membrane to uranium enrichment, adsorptive COF powder is synthesized through a solvothermal method, the maximum uranium enrichment amount of the adsorptive COF powder can reach 520 mg / g under the condition that the pH of the adsorptive COF powder is equal to 5, and the adsorptive COF powder is processed into the COF nanofiber membrane through an electrostatic spinning method. The COF nanofiber membrane prepared by the method can retain the properties of COF powder, not only has adsorption capacity to U (VI), but also has photocatalytic reaction capacity, and can convert soluble U (VI) into insoluble uranium and release additional adsorption sites, so that the enrichment capacity of a COF material to uranium is improved, and the enrichment efficiency of the COF material to uranium in water reaches 97% or above. In addition, the COF nanofiber membrane is also beneficial to use and recovery, and is more conveniently applied to the fields of uranium-containing wastewater treatment, uranium extraction from seawater and the like.

The invention discloses a nuclear power plant reactor core fuel assembly uranium contamination fission product calculation method and system, and the method comprises the following steps: monitoring the characteristic nuclide of a loop coolant of a reactor core fuel assembly in real time, and judging whether the surface of the reactor core fuel assembly has uranium contamination; if the uranium contamination exists, calculating the uranium contamination amount of the uranium contamination distributed in the primary loop coolant on the surface of the reactor core fuel assembly; and calculating uranium contamination fission product distribution of the reactor core fuel assembly according to the uranium contamination amount. According to the method, the uranium contamination fission product distribution in the primary loop coolant is collected, so that the risk of erroneous judgment of the damage of the reactor core fuel assembly caused by uranium contamination can be remarkably reduced.

The invention relates to the technical field of cable logging, and particularly discloses a mineral component and content testing device and method suitable for a uranium-bearing reservoir, and the device comprises a cable logging system, a logging depth calibration system and a time synchronization system, the logging tool assembly comprises a spectral logging tool, a natural gamma-ray spectrum logging tool, an epithermal neutron logging tool and a neutron capture gamma-ray spectrum logging tool. The method comprises the steps of combining a logging tool assembly and a cable logging system, configuring a logging depth calibration system, configuring a time synchronization system, carrying out PTP time synchronization on the cable logging system and the logging depth calibration system, collecting original logging time and original logging data, carrying out time resampling and depth correction on the original logging data, and carrying out time synchronization on the logging depth calibration system and the cable logging system. And resampling the logging data according to the logging depth and time to divide a stratum structure. According to the method, acquisition and data fusion of the mineral components, the radioactive element content and the porosity characteristics of the stratum around the well can be realized, and the exploration precision, efficiency and evaluation accuracy are improved.

The embodiment of the invention relates to the technical field of electrolytic production, recovery or refining of metal through a molten liquid electrolytic method, in particular to molten salt electrolytic refining equipment suitable for a hot chamber and a method for recovering uranium metal in spent fuel through molten salt electrolytic refining. The cathode and cathode transfer device is used for transferring the cathode into the electrolytic heating device or transferring the cathode away from the electrolytic heating device; the anode and anode transfer device is used for transferring the anode into the electrolytic heating device or transferring the anode away from the electrolytic heating device; the electrolytic product scraping and collecting device is used for carrying out scraping treatment on the cathode and collecting scraped objects; the electrolytic heating device moving assembly is used for enabling the electrolytic heating device to be moved to the hoisting station; and the cathode conductive device is connected with the cathode so as to electrify the cathode. According to the device provided by the embodiment of the invention, the electrolytic reaction can be continuously carried out, and the treatment requirement of the spent fuel is met.

The invention belongs to the technical field of inorganic powder, and discloses low-radioactivity spherical aluminum oxide as well as a preparation method and application thereof. The preparation method comprises the following steps: adopting ultrapure water at 20-60 DEG C, and adding acid to adjust the pH value to be less than or equal to 2.5 to obtain acid liquor; taking aluminum hydroxide / boehmite / industrial alumina as raw materials, adding the raw materials into the acid liquor, and stirring and dispersing to obtain mixed slurry A; adding alkaline anion exchange resin into the mixed slurry A, carrying out hydrothermal stirring treatment for 8-12 hours, and removing the alkaline anion exchange resin to obtain mixed slurry B; carrying out suction filtration and drying on the mixed slurry B, and carrying out heat preservation treatment on the dried material at 1300-1400 DEG C for 6-9 hours; and S4, crushing and dispersing the material subjected to heat preservation treatment in the step S4, and melting and spheroidizing at 1100-1300 DEG C to obtain a low-radioactivity spherical aluminum oxide crude product. The prepared low-radioactivity spherical alumina powder is compact in structure, high in thermal conductivity, low in uranium content and suitable for EMC production and application.

Disclosed in the present invention is a fuel rod for a water-cooled reactor. The fuel rod comprises a cladding tube, an upper end plug and a lower end plug that match at two opposite ends of the cladding tube, and fuel pellets stacked in the cladding tube, wherein the fuel pellets comprise at least one first pellet; the first pellet is formed by at least two pellet units connected in an enclosed manner in a circumferential direction; and a through hole penetrating two opposite ends of the first pellet is formed in the first pellet, and is filled with a filler containing burnable poison nuclides. In the fuel rod for a water-cooled reactor of the present invention, by means of providing a hole in a pellet and filling the hole with the filler containing the burnable poison nuclides, a gas generated by the pellet during the operation of the reactor can be absorbed, thereby reducing the internal pressure of the cladding tube, reducing the risk of fuel rod failure, preventing the leakage of radioactive substances, increasing the uranium utilization rate, and providing a relatively large design margin.

The invention discloses a production process of spherical alumina powder with high purity and low uranium content, which comprises the following steps: by taking angular alumina with the uranium content of 50-100ppb as a raw material, sequentially carrying out flame melting spheroidization, gradient pickling, hydrothermal strengthening uranium removal, wet grading coarse grain removal, low-temperature drying and airflow depolymerization to finally obtain the finished product spherical alumina powder with the uranium content of less than 30ppb. Through a multi-stage uranium removal system including incoming material management and control, gradient pickling, hydrothermal coupling and particle size optimization, the traditional single pickling efficiency bottleneck is broken through, and deep dissolution of lattice uranium and effective removal of coarse-grain uranium enrichment nuclei are achieved; the morphology and the specific surface are locked through whole-process low-temperature hydrothermal and low-shear drying, so that the powder is ensured to be used as a filler in resin sealing materials of high-integration integrated circuits, large-scale integrated circuits and super-large-scale integrated circuits, and the powder has the comprehensive performance of high filling, low thermal resistance and zero alpha-ray soft errors; and the limit requirements of node packaging materials of 3nm and below are met.

The invention relates to a construction method of an organic phase uranium working curve, which comprises the following steps: completely extracting uranium in a water phase uranium stock solution by adopting a 30vol% TBP-kerosene extraction technology, transferring the uranium into an organic phase, and determining the value of the uranium in the organic phase by combining an automatic potentiometric titrator and a trace uranium analyzer analysis technology. Compared with the prior art, the method provided by the invention is accurate and reliable and extremely good in uranium separation effect through theoretical and experimental verification, and can be used for indirect valuing of uranium in an organic phase, so that the recovery rate of uranium is effectively improved, and the method has important application value for closed circulation of nuclear fuel and reuse of uranium resources.

The present invention belongs to the technical field of in-situ uranium leaching, and specifically relates to a method for increasing the uranium concentration of in-situ uranium leaching solution. The method comprises the following steps: determining the series connection sequence of mining areas; connecting the liquid extraction main pipe of the previous mining area with the liquid injection main pipe of the next mining area according to the series connection sequence of mining areas determined in step one; and connecting the leaching agent pipeline with the liquid injection main pipes of each mining area respectively. According to the series connection sequence of mining areas, the valves of the liquid extraction main pipe and the liquid extraction main pipe of the last mining area are opened, and the submersible pumps of the liquid extraction holes of each mining area are started in sequence from back to front to maintain the balance of extraction and injection of the mining area, and then the amount of leaching agent added is calculated according to the leaching agent concentration demand of the corresponding mining area and the residual leaching agent concentration in the leaching solution of the previous mining area. The method superimposes the uranium leached from multiple mining areas, and at the same time, the uranium concentration of the leaching solution is increased, and the residual leaching agent in the leaching solution is reused, thereby effectively improving the production capacity of the in-situ leaching mine and reducing the material consumption during the in-situ leaching of uranium.

The invention discloses a powder double-layer suction pipe matched with an upper-opening storage barrel and a method, and relates to the technical field of uranium conversion related equipment. The device is used in cooperation with a storage barrel with an upper opening. The device comprises a double-layer pipe and a material gathering plate, the double-layer pipe comprises an outer pipe and an inner pipe coaxially arranged in the center of an inner hole of the outer pipe, an annular column area is formed between the inner pipe and the outer pipe, and a plurality of feeding holes are formed in the lower end of the inner pipe; the plurality of material gathering plates are annularly and uniformly distributed and fixedly connected to the outer circular surface of the lower end of the outer pipe; a feeding hole is formed in the excircle surface of the lower end of the outer pipe. A powder suction method matched with a storage barrel with an upper opening is used for being matched with a negative pressure system to suck powder materials in the storage barrel with the upper opening, and by being matched with an original eccentric rotation suction method, powder material residues are avoided, and the suction efficiency is improved. The invention provides a powder material suction structure and a powder material suction method which are adaptive to an upper-opening storage barrel based on a technical route of conveying powder materials through a gas path, and the powder material suction structure and the powder material suction method are particularly suitable for efficient and residue-free transfer of UO2 powder in the nuclear fuel preparation process.

The embodiment of the invention relates to the technical field of electrolytic production, recovery or refining of metal through a molten liquid electrolytic method, in particular to an electrolytic product scraping and collecting device, a method for recovering uranium metal in spent fuel through molten salt electrolytic refining and a method for designing a cutter for scraping precipitates. The device comprises a containing assembly used for containing a cathode and arranged to enable the cathode to rotate, and an electrolysis product is formed on the cathode; the scraping assembly can be close to or far away from the cathode located in the containing assembly and is used for scraping electrolytic products on the cathode in the rotating process; the electrolytic product collecting piece is arranged to collect scraped electrolytic products in the scraping process of the scraping assembly and transfer the collected electrolytic products to the next procedure after scraping is completed. According to the embodiment of the invention, the material scraping assembly is used for scraping the electrolytic products on the rotating cathode, then the electrolytic product collecting piece is used for collecting the scraped electrolytic products and transferring the scraped electrolytic products, the recycling process does not need participation of operators, the efficiency is relatively high, and the safety is also relatively high.

The invention discloses a preparation method and application of an S-type heterostructure material with a plasma resonance effect, and the preparation method comprises the following steps: grinding a precursor material, calcining, collecting solid powder, carrying out reflux purification in deionized water, and collecting white powder; dispersing the white powder in a hydrazine hydrate solution, carrying out solvothermal reaction to obtain a faint yellow solution, repeatedly dissolving and precipitating by respectively using acidic and alkaline reagents, washing and drying to obtain a faint yellow solid product; uniformly grinding the faint yellow solid product, calcining the faint yellow solid product in an atmosphere furnace, and cooling to obtain orange solid powder; and adding the orange solid powder into an organic solvent of a precursor containing a tungsten source, and centrifuging, washing and drying after the solvothermal reaction is finished to obtain the S-type heterostructure material with the plasma resonance effect. The S-type heterostructure material with the plasma resonance effect has relatively high efficiency for simulating separation and removal of uranium species in post-treatment wastewater and shows relatively good anti-radiation stability.

The invention belongs to the technical field of production of isotopes by a sodium-cooled fast reactor, and aims to solve the problems that the moderation effect is limited, the reactor core disturbance is relatively large and fast neutron energy spectrum is not beneficial to the production of the ultra-uranium isotopes during the production of the ultra-uranium isotopes. The invention provides a method for producing a super-uranium isotope by using a sodium-cooled fast reactor and a sodium-cooled fast reactor core. According to the method, moderator assemblies and super-uranium isotope production assemblies are arranged on a first circle of a radial conversion area in a flower arrangement mode, the super-uranium isotope production assemblies are arranged on a second circle of the radial conversion area, and the side, facing the center of the reactor core, of the first circle is coated with a thermal neutron absorption material; the reactor core is provided with a radial conversion area, moderator assemblies and super-uranium isotope production assemblies are arranged on a first circle of the radial conversion area, the super-uranium isotope production assemblies are arranged on a second circle of the radial conversion area, and thermal neutron absorbing materials are arranged on the side, facing the center of the reactor core, of the first circle. The super-uranium isotope is concentrated in the radial conversion region for production, the moderation effect is good, the disturbance to the reactor core is small, and the productivity is high.

The invention relates to the technical field of remote sensing prospecting, and discloses a uranium-copper multi-metal remote sensing prospecting method which comprises the following steps: S1, data acquisition and preprocessing; s2, alteration information is extracted; s3, multi-source information comprehensive analysis and prospecting prediction; s4, alteration abnormity verification is carried out; and S5, delineating the prospecting target area. According to the method, vegetation spectrum interference is stripped by carrying out principal component analysis on visible light-short wave infrared of a remote sensing image, weak iron staining and hydroxyl alteration signals are extracted, a geological and remote sensing coupling technology is carried out, alteration anomaly is superposed with a geological map, geochemical data and a construction interpretation result, a comprehensive remote sensing anomaly map is obtained, and the comprehensive remote sensing anomaly map is obtained. According to the method, the multi-parameter grading model GGS is established, and the target region is graded based on the multi-parameter grading model GGS, so that the probability of missed judgment of the target region is reduced.

The invention belongs to the technical field of heat pipe reactors, and discloses a hexagonal annular heat pipe cooling reactor, which comprises a pipe wall of a heat pipe, a silk screen wick, a steam cavity structure of the heat pipe, uranium-molybdenum alloy fuel, an outer pipe wall of the heat pipe, a wick attached to the inner wall surface of the heat pipe, and uranium-molybdenum alloy. The device is formed by coupling a hexagonal annular heat pipe with a hexagonal fuel element. Firstly, in the aspect of the structure, a framework structure composed of graphite is not needed to fix the heat pipe and the fuel element, an assembly composed of the heat pipe and the fuel element can be directly adopted for expansion to form the fuel element, and gap thermal resistance brought by the framework is eliminated. In addition, in terms of heat transfer performance, the heat transfer performance of the hexagonal annular heat pipe is superior to that of a traditional cylindrical heat pipe. The same heat transfer area is adopted, and under the condition of the same heat transfer power, the temperature of the outer surface of the pipe wall of the hexagonal annular heat pipe is lower than that of a cylindrical heat pipe. Therefore, the material stress requirement in the heat pipe pile and the thermoelectric conversion form of the condensation section can be richer.

The invention discloses a low-grade uranium-gold polymetallic ore rare earth element leaching method and device, and belongs to the technical field of rare earth element extraction. The low-grade uranium-gold polymetallic ore rare earth element leaching method comprises the steps that S1, low-grade uranium-gold polymetallic ore containing rare earth elements is pretreated; s2, preparing an acidic solution, and putting the acidic solution into a leachate container; s3, preparing a hydrogen peroxide solution, and putting the hydrogen peroxide solution into a leachate container; s4, the acid solution and the hydrogen peroxide solution are injected into the reaction container from bottom to top at the speed of 0.5 ml / min. The method has the advantages that the rare earth element leaching efficiency is improved, the environmental pollution risk is reduced, the equipment corrosion is reduced, and the subsequent purification process is simplified.

Uranium conversion material multi-round continuous feeding system and feeding method, which relate to the technical field of uranium conversion related equipment. The uranium conversion material multi-round continuous feeding system includes a plant, a storage tank, a transfer tank, a material merging assembly, and a material feeding assembly. The material merging assembly includes a transfer tank rotating device, a storage tank translation trolley, a storage tank outer cover disassembly and assembly device, a storage tank inner cover opening and closing device, a transfer tank upper cover disassembly and assembly device, and a lower pipeline docking device. The material feeding assembly includes a transfer tank translation trolley, an upper pipeline docking device, and a transfer tank inner cover opening and closing device. The uranium conversion material multi-round continuous feeding method is applied to the uranium conversion material multi-round continuous feeding system, and realizes the automatic feeding of UO2 powder material before the start of the uranium conversion process in the nuclear fuel preparation process. On the basis of the traditional feeding process, the present invention optimizes the feeding process, realizes the multi-round continuous feeding operation of UO2 powder material, and meets the needs of large-scale, fast-paced, and high-efficiency uranium conversion production.

The invention discloses a composite photocatalyst for removing uranium in nuclear waste liquid and a preparation method thereof, and the preparation method comprises the following steps: dropwise adding a solution B dissolved with Na2S. 9H2O into a solution A dissolved with CdCl2. 5H2O to obtain a mixed solution; heating the mixed solution for reaction; after cooling to room temperature, centrifuging and washing a product; freeze-drying the washed product overnight to obtain cadmium sulfide; dispersing cadmium sulfide to obtain a suspension; agNO3 is added into the suspension liquid; heating the suspension to react; and washing for several times, and freeze-drying to obtain the silver-loaded cadmium sulfide composite photocatalyst with the cadmium vacancy. The prepared composite photocatalyst is used for adsorption and photo-reduction treatment of radioactive uranium in nuclear waste liquid, and efficient photocatalytic removal of uranium in uranium waste water within the pH value range of 4-8 is achieved. On the premise of being green and environment-friendly, only light and a photocatalyst are needed, pollution to the environment is avoided, operation is easy and convenient, the reaction condition is mild and not harsh, the reaction catalyst can be recycled, and cost is saved.

The invention discloses a uranium adsorption method. The uranium adsorption method comprises the following steps: (1) mixing paraffin containing 6-10 carbon atoms with a span emulsifier to obtain a continuous phase; the preparation method comprises the following steps: mixing 1-vinyl imidazole, N, N-methylene bisacrylamide and azodiisobutyronitrile to obtain a dispersed phase; wherein on the basis of 45 mL of 1-vinyl imidazole, the use amount of the N, N-methylene bisacrylamide is 3 to 7 g, and the use amount of the span emulsifier is 10 to 30 mL; adding the dispersed phase into the continuous phase, heating to T1, and then heating to T2 to obtain the uranium adsorption material, wherein T1 ranges from 85 DEG C to 95 DEG C, the time from the initial temperature to T1 ranges from 1.5 h to 2.5 h, and T2 ranges from 95 DEG C to 110 DEG C; (2) placing the uranium adsorption material in a uranium solution for adsorption; the pH value of the uranium solution is 4.5-6. By adopting the uranium adsorption method, the uranium adsorption capacity of the uranium adsorption material can be improved.

The invention relates to the technical field of extraction separation, and provides an extraction separation device which comprises a body, the body is provided with a sample disc, a reagent bottle and an extraction column disc, the extraction column disc is provided with a plurality of separation columns, and the separation columns are used for selective adsorption and separation of uranium matrixes and impurity elements; the first mechanical arm is arranged on the body; the liquid adding assembly is arranged on the first mechanical arm, and the liquid adding assembly is used for taking liquid from the reagent bottle and executing liquid adding operation; the pipetting assembly is arranged on the first mechanical arm, and the pipetting assembly is used for sampling from the sample disc and executing sample loading operation; the second mechanical arm is arranged on the body; the pressurizing device is arranged on the second mechanical arm and is used for executing sealing pressurizing operation on the separation column; the receiving assembly is arranged on the body, can move relative to the body and is used for collecting the sample solution and the waste liquid. A plurality of problems during manual separation operation are solved, and the whole process of sample loading, leaching, elution, receiving, constant volume, waste liquid collection and the like is automatically completed.

The present application relates to the technical field of water treatment and environmental governance, and particularly relates to a nickel-strontium titanate electrode for removing uranium, a preparation method and application thereof, and the preparation method comprises the following steps: mixing a titanium source, an alcohol amine carrier and an alkaline solution to obtain a precursor solution; adding a nickel source and a strontium source into the precursor solution, performing ultrasonic treatment and hydrothermal reaction to obtain a nickel-strontium titanate material; and coating the nickel-strontium titanate material on the surface of a carbon fiber base material to obtain the nickel-strontium titanate electrode for removing uranium. Through the combination process of nickel doping, hydrothermal reaction and carbon cloth base coating, the prepared electrode forms a porous and uniformly dispersed composite structure, and meanwhile, the electrode can have optimal selectivity and removal efficiency for uranium in the range of pH 3-9. The nickel-strontium titanate electrode for removing uranium prepared by the present application is suitable for large-scale water treatment applications, and has a good application prospect in the fields of nuclear waste water and drinking water purification.

The application discloses a uranium-based metal organic framework material and a preparation method and application thereof, and the chemical formula of the uranium-based metal organic framework material is [(UO2 2+ )(H2L 2‑ )·H2O] n , [(UO2 2+ )3(HL 3‑ )2(H2O)3·3H2O] n or [(UO2 2+ )(H2L 2‑ )·2H2O] n , wherein n is an integer greater than 0, and L is an aromatic tetracarboxylic acid. A series of uranium-based MOF materials are synthesized by using a solvent-thermal method, the X-ray stability of the MOF structure is significantly improved by using the strategy of preparing luminescent uranium-based MOF materials by using flexible carboxylic acid aromatic ligands, the prepared uranium-based MOF material has the advantages of diversified structure, stable property, excellent performance and wide application range, and the uranium-based MOF material is applied to the imaging field for the first time.

The present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

The invention discloses an equivalent amplification design method of an electrolytic bath for uranium electrolytic refining in a dry post-treatment process, and belongs to the field of spent fuel post-treatment. According to the method, the influence of a stirring device on uranium ion distribution in a molten salt electrolytic bath and the uranium electrolytic refining rate and yield is obtained through a simulation result, the optimal rotating speed of the stirring device is determined, and then the stirring Reynolds number and the end face speed are further determined as amplification criteria of the electrolytic bath; after the amplification design is carried out, the change condition of each amplified key parameter is estimated by virtue of simulation software, and meanwhile, the influence on uranium ion distribution in the molten salt electrolytic bath and the uranium electrolytic refining rate and yield under different amplification criteria is compared and analyzed, so that an optimal amplification criterion is obtained; and finally completing equivalent amplification design of the electrolytic bath for uranium electrolytic refining in the dry post-treatment process, thereby realizing amplification design of the molten salt electrolytic bath. By adopting the method disclosed by the invention, the design and amplification of the electrolytic cell can realize precision and high efficiency, and the time cost and the capital investment are saved.

The invention belongs to the technical field of uranium concentration, and particularly relates to a uranium concentration factory product abundance real-time monitoring method, device and system. The method comprises the following steps: introducing UF6 gas to be measured into a measuring container, and obtaining the total count rate of gamma rays in the measuring container by using a NaI detector; acquiring abundance of the UF6 gas according to a background obtained in advance, the total count rate, a scale coefficient and the pressure and temperature in the measuring container; wherein the background is determined by the following steps of: measuring counting rates of characteristic gamma rays in the measuring container under different pressures by using a NaI detector to obtain a plurality of counting rates; the lower the pressure is, the longer the measurement time of the NaI detector is; and fitting the counting rate of the corresponding gamma ray when the pressure is zero according to the plurality of obtained counting rates and the corresponding pressure to obtain the background. According to the invention, the dependence of product abundance monitoring on operators is eliminated, the abundance measurement precision is ensured, and the operation and maintenance cost is reduced.