367 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.

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.

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 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 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 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 provides an in-situ leaching uranium mining pipe cable lifting, releasing and bundling device and method, and belongs to the technical field of in-situ leaching uranium mining well site mining. A lifting pipe and a cable double-winch mechanism are arranged, the lifting pipe and a cable can be quickly wound / released, the cable is prevented from being wound and knotted, the cable does not need to be coiled on the ground, and the cable is prevented from being contaminated by soil or dirt; the pollution to the leaching liquid in the well caused by lifting and placing the cable for multiple times is avoided; when the lifting pipe and the cable need to be put down, the lifting pipe is led out of the first winch mechanism, the cable is led out of the second winch mechanism, and the lifting pipe and the cable penetrate out of the same end of the bundling device. The lifting pipe and the cable which are unwound continuously are bundled into a whole through the bundling device, the automation level of well site operation is improved, the labor cost is saved, the labor intensity is greatly reduced, the submersible pump lowering efficiency is improved, the contact time of an operator and a pipe cable assembly is greatly shortened, and the safety of submersible pump lowering operation is improved.

The application discloses an automatic air supplement method and device for an in-situ leaching uranium gas station and belongs to the technical field of in-situ leaching uranium. The automatic air supplement method comprises the following steps: according to the pressure of a gaseous gas sample storage tank and the injection pressure, automatically controlling the air supplement process of the gaseous gas sample storage tank, so that the pressure of the gaseous gas sample storage tank is greater than the injection pressure; and according to the injection pressure and the liquid injection pressure, automatically controlling the injection process, so that the injection pressure is greater than the liquid injection pressure. Through real-time detection of the pressure of each part of the in-situ leaching uranium gas station, the embodiment of the application realizes automatic and accurate regulation and control of the carbon dioxide system of the in-situ leaching uranium gas station, improves the efficiency and saves manpower.

This invention provides an optimized method and system for laser cleaning of product containers in uranium enrichment plants. It proposes a method for optimizing laser cleaning parameters. In the actual laser cleaning process, real-time data is input into the model to obtain the prediction results, i.e., the optimal laser cleaning parameter configuration, so as to achieve the best cleaning effect. A double wedge mirror spiral scanning method is used to achieve full coverage scanning of the inner wall of the container. The cleaning speed can reach up to 9000 mm / s, which improves the speed and efficiency of cleaning.

The invention relates to a method for electroplating a nickel-zinc alloy on the surface of uranium metal fuel. The method comprises the following steps: removing an oxide layer on the surface of the uranium metal fuel; the uranium metal fuel with the oxide layer removed is placed in a ferric trichloride solution to be etched and activated; and the etched and activated uranium metal fuel serves as a cathode and is placed in an electrolyte containing zinc ions and nickel ions for electroplating, and the nickel-zinc alloy coating is obtained on the surface of the uranium metal fuel. The nickel-zinc alloy coating obtained on the surface of the uranium metal fuel through the method has the advantages of being bright in appearance, low in toxicity, small in hydrogen embrittlement, good in toughness, good in weldability and the like, the surface protection performance and corrosion resistance of the uranium metal fuel are remarkably improved, then the service life of the uranium metal fuel is prolonged, and the safety of the uranium metal fuel is improved; the uranium metal fuel is suitable for being stored and used in severe industrial atmosphere and harsh marine environment, and the method has the advantages of simplified preparation process, high production efficiency and low production cost, so that further development of the uranium metal fuel in the fields of national defense and nuclear energy is promoted.

The invention discloses a magnesium-aluminum hydrotalcite / carbon nanotube composite material and a preparation method and application thereof, and the preparation method of the magnesium-aluminum hydrotalcite / carbon nanotube composite material comprises the following steps: S1, carrying out acid pretreatment on a carbon nanotube; and S2, preparing the magnesium-aluminum hydrotalcite / carbon nanotube composite material. Compared with the prior art, the preparation method of the magnesium-aluminum hydrotalcite / carbon nanotube composite material has the advantages that Mg (NO3) 2.6 H2O, Al (NO3) 3.9 H2O, CO (NH2) 2 and pretreated carbon nanotubes are adopted, and magnesium-aluminum hydrotalcite grows on the carbon nanotubes through a hydrothermal homogeneous precipitation method; through the synergistic effect of the carbon nanotubes and the magnesium-aluminum hydrotalcite, aggregation among nanoparticles is effectively prevented, the growth size of the particles is limited, and the magnesium-aluminum hydrotalcite / carbon nanotube composite material is good in crystallinity, small in particle size, high in specific surface area and high in adsorption performance. In a radioactive uranium element adsorption experiment, the adsorption efficiency is as high as 95.6%, and the adsorption capacity is as high as 17.32 mg / g.

The invention belongs to the nuclear industry and can be used for the production of fuel pellets from uranium-molybdenum metal powder enriched with uranium-235 to 7%, which can be used as fuel elements for nuclear reactors. The sintering of the pellets is carried out in an inert argon atmosphere at a temperature in the range of 1100°C to 1155°C, as the initial powder, a particle size of 160 μm is used, with a molybdenum content of 9.0 to 10.5 wt.%. The powder is preheated at a temperature of 500°C for 10-20 hours (in an argon medium), and then cold-pressed into pellets in a mold at a pressure of up to 950 MPa. According to another method, when preparing uranium-molybdenum pellets using a binder (plasticizer), the pellets are heated in argon from 300°C to 450°C for a duration of 2-4 hours before sintering, in order to remove the binder. The invention allows to increase the uranium content in the fuel, to reduce the heat accumulated in the core of the nuclear reactor, and to reduce the energy release in the event of a violation of the normal operating conditions of the nuclear reactor, which will improve its safety and emergency stability.

The application discloses a kind of nuclear material plate uranium uniformity detection devices, it includes the feeding roller way conveying device (1) sequentially arranged from left to right, support (2) and discharging roller way conveying device (3), support (2) is respectively provided with detector assembly (4) and calibration assembly (5) on front and rear ends, detector assembly (4) is set to the front side of feeding roller way conveying device (1), detector assembly (4) includes linear module (6), detector (7) and gamma ray radiation source (8), linear module (6) is set to the top surface of support (2), the slider (22) of linear module (6) is fixed with connecting frame (9), the rear end of connecting frame (9) is provided with L plate (10), and the horizontal plate of L plate (10) is provided with gamma ray radiation source (8).The beneficial effects of the present application are: compact structure, realize the automatic flow detection of the uranium surface density value of nuclear material plate, improve the detection efficiency, simple operation.