251 results about "Nuclear industry" patented technology

The invention relates to the technical field of data processing, and discloses a thinking instruction data generation method and system, electronic equipment and a storage medium, and the method comprises the steps: integrating a multi-source heterogeneous nuclear industry data corpus, and forming a standardized data source; thinking chain instruction data are constructed through an instruction generation framework driven by a large language model, multi-level security constraint processing is performed on the thinking chain instruction data, and contents which do not conform to nuclear industry security specifications are identified and eliminated; executing multi-dimensional quality evaluation on the thinking chain instruction data passing the security constraint, and removing the thinking chain instruction data which does not reach a quality threshold value; and performing real-time updating and dynamic optimization on the nuclear industry data corpus and the thinking chain instruction data based on newly collected nuclear industry field data. The method can improve the flexibility, accuracy and timeliness of instruction generation.

The invention relates to a crown ether covalent organic framework material as well as a preparation method and application thereof. The preparation method comprises the following steps: dissolving a crown ether monomer and an amino monomer in an o-dichlorobenzene solvent, adding a catalyst acetic acid aqueous solution, and uniformly dispersing to form a reaction solution; the method comprises the following steps: freezing a reaction solution through a liquid nitrogen bath, vacuumizing, filling nitrogen, carrying out a heating reaction, and carrying out post-treatment operations such as washing and drying on a reaction product, so as to obtain the crown ether covalent organic framework material BADT-DB18C6-COF. The crown ether covalent organic framework material is formed by condensation of a crown ether aldehyde monomer and an amino monomer, the crown ether monomer is 4, 4 ', 4' ', 4 '''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde, and the crown ether monomer is 4, 4 ', 4' ', 4'''-(6, 7, 9, 10, 17, 18, 20, 21-octahydrodibenzo [B, K] [1, 4, 7, 10, 13, 16] hexaoxacyclo octadecene-2, 3, 13, 14-tetrayl) tetrabenzaldehyde. The amino monomer is 4 ', 5'-bis (4-aminophenyl)-[1, 1 ': 2', 1 ''-terphenyl]-4, 4''-diamine. The crown ether covalent organic framework material BADT-DB18C6-COF prepared by the invention is mainly used as an adsorbent, is used for adsorbing and recovering uranium, thorium and americium ions from a solution, shows good adsorption capacity on radionuclides U-238, Th-232 and Am-243, and is particularly suitable for the field of spent fuel post-treatment in the nuclear industry.

The invention discloses a multi-component composite oxide dispersion strengthening molybdenum alloy and a preparation method thereof, and relates to the technical field of refractory metal powder metallurgy preparation. The molybdenum alloy comprises a molybdenum matrix and a Y-X-O composite oxide nano precipitated phase dispersed and distributed in the molybdenum matrix. The molybdenum alloy comprises the following components in percentage by mass: 1%-2% of Y2O3, 0.3%-1.2% of an active element X, 96.5%-98.7% of molybdenum and inevitable impurities, wherein X is selected from one of Ti, Zr and Hf; the Y-X-O composite oxide nano precipitated phase and the molybdenum matrix form a coherent or semi-coherent interface, and the interface mismatch degree is less than or equal to 5%. The molybdenum alloy has excellent high-temperature strength and creep resistance, and can be applied to the high-temperature fields of nuclear industry, aerospace and the like.

The invention discloses a temperature control method for low-temperature high-NOx-concentration flue gas denitration in the nuclear industry, and relates to the technical field of flue gas denitration control. According to the method, flue gas temperature time sequence distribution, NOx concentration gradient and environment radiation intensity data are integrated through a space-time alignment algorithm, a synchronous data matrix representing a system coupling state is generated, then a temperature and efficiency response curved surface is established, a high-efficiency temperature area and an inactivation risk area are dynamically divided through fuzzy logic control, and an optimal temperature regulation and control threshold value is recognized. The drift effect of quantitative radiation on the surface temperature of the catalyst is analyzed through regression, a quantitative relation model of radiation dose and temperature zone deviation is established, and real-time correction of anti-radiation compensation parameters is achieved. And finally, optimizing the multi-objective cost function by adopting a non-dominated sorting genetic algorithm, and obtaining dynamic balance between the maximum NOx conversion rate and the minimum temperature fluctuation. And multi-parameter coupling control of a temperature field and a radiation field is realized, and the denitration efficiency stability under a low-temperature condition is remarkably improved.

Provided in the present invention are a preparation method for a crown ether covalent organic framework material, and the use thereof in iodine adsorption. In the present invention, an 18-crown-6-ether group is introduced into the structure of a covalent organic framework (COF) material for iodine adsorption work. The covalent organic framework material can form a one-dimensional pore channel structure in order, and the 18-crown-6-ether can be selectively complexed with iodine molecules, thereby achieving the purpose of efficient iodine adsorption. The material can perform specific iodine adsorption from high-temperature iodine steam, and the iodine steam adsorption capacity of the COF material is as high as 5.56 g / g. The COF material has a high adsorption capacity, is easy to prepare and simple to operate, has a high adsorption efficiency, and is particularly suitable for adsorbing and separating radioactive iodine in the nuclear industry.

The invention discloses an underwater operation nuclear industry robot which comprises an amphibious moving system, and the amphibious moving system comprises a chassis, a land moving module, an underwater moving module and an adsorption fixing module; the two sets of operation systems are installed on the front side of the top face of the chassis, each operation system comprises a multi-degree-of-freedom mechanical arm and a grabbing mechanical arm, the multi-degree-of-freedom mechanical arms are installed on the top face of the chassis, and the grabbing mechanical arms are installed at the ends of the multi-degree-of-freedom mechanical arms; the recovery and storage system is mounted on the top surface of the chassis; the visual recognition system is installed on the rear side of the top face of the chassis, and the visual recognition system comprises a supporting module, a lighting module and a recognition module; and the amphibious moving system, the operation system, the recovery and storage system and the visual identification system are all connected with the terminal controller. The underwater robot can flexibly enter an underwater operation area from the land, and is suitable for operation in various complex environments.

The invention discloses a cable channel generation method and device and a cable channel display method and system, and relates to the technical field of nuclear industry. The cable channel generation method comprises the following steps: acquiring a target model and a target cable information list; associating the N groups of target cable data information with a target model; determining P groups of target cable characteristic information in one-to-one correspondence with the P groups of target cable data information for the P groups of target cable data information contained in the target cable path channel; and generating a target cable channel according to the P groups of target cable data information and the P groups of target cable characteristic information. According to the embodiment of the invention, the method can achieve the real-time updating of the cable channel according to the upstream input change, can reduce the redundant work in the design, can reduce the human input, and can improve the design quality of a construction drawing through the development of the subsequent bridge design based on the cable channel, thereby reducing the no-load condition of a site bridge.

The invention belongs to the technical field of nuclear power, and particularly relates to a nuclear industry uranium mine mining equipment inspection management system. According to the nuclear industry uranium mine mining equipment inspection management system provided by the invention, the inspection point configuration model taking the equipment assets as the core is constructed, and deep binding of the inspection task and the equipment ledger is realized. On the basis, the system can automatically generate a matched inspection plan according to key parameters such as the operation cycle and the inspection frequency of the equipment, and accurate compilation and standardized configuration of tasks are achieved. The generation of the inspection plan does not depend on artificial experience any more, artificial omissions and configuration errors are effectively reduced, the systematicness and normalization of task coverage are improved, timely inspection and high-frequency management and control of high-risk equipment are guaranteed, and the rigid requirement of the nuclear industry for high-reliability operation and maintenance management is met.

The invention discloses a multifunctional variable-diameter nuclear waste conveying pipeline cleaning robot, and relates to the technical field of nuclear industry pipeline maintenance equipment.The multifunctional variable-diameter nuclear waste conveying pipeline cleaning robot comprises a walking module, a variable-diameter module, a cleaning module and a detection module, and all the modules are integrated on a machine body frame; a collaborative linkage design of a series elastic driver (a main driving end) and a bionic joint leg (a driven wheel end) is adopted as the core, the main driving end adapts to nuclear pipeline advancing impact through elastic buffering, and the driven wheel end adapts to a complex inner wall through flexible posture adjustment. The variable-diameter module is linked with a connecting rod to adjust the radius of the wheel set to realize pipe diameter self-adaption; the cleaning module integrates a rotating brush and a cavitation jet device to remove radioactive scales; the detection module carries a sensor and a camera to monitor the pipeline state. The robot can complete cleaning and monitoring of nuclear pipelines of multiple specifications, and stable and smooth operation of the robot in pipelines of different pipe diameters is guaranteed through linkage of the series elastic drivers and the bionic joint legs.

The invention discloses a unit area heat release rate determination method and a related method and related equipment thereof, and relates to the technical field of nuclear industry. The unit area heat release rate determination method is used for a target solvent and comprises the steps that the initial unit area heat release rate of the target solvent in an initial state, the initial combustion duration of the target solvent in the initial state and the target combustion duration of the target solvent in a target state are obtained, and the target solvent is in an inverted cone shape in the combustion process; and determining a target unit area heat release rate of the target solvent in the target state according to the initial unit area heat release rate, the initial combustion duration and the target combustion duration. According to the embodiment of the invention, the method fully considers the influence of the initial combustion duration and the target combustion duration on the heat release rate of the target unit area, so that the accuracy of determining the heat release rate of the target unit area can be improved.

The sealing swing hydraulic joint comprises a joint sleeve, a fork frame, a rotary hydraulic cylinder, a sealing assembly and a wire passing hole, the joint sleeve is connected with a driving arm pipe, a cavity is formed in the joint sleeve, and the sealing assembly can seal and block an opening of the cavity in a movable and rotary mode in the axial direction of the cavity so that the sealing assembly and the joint sleeve can form a sealing cavity wrapping the rotary hydraulic cylinder. The rotary hydraulic cylinder is contained in the sealing cavity, a cylinder body and a driving end are fixedly connected with the joint sleeve and the sealing assembly respectively, the fork frame is connected with the driven arm pipe and fixedly connected with the sealing assembly, the sealing assembly serves as the hinged position of the fork frame and the joint sleeve, and under driving of the rotary hydraulic cylinder, the driven arm pipe is driven to swing relative to the driving arm pipe. The wire passing hole is an internal continuous hole formed in the joint sleeve, the sealing assembly and the fork frame so that the pipe cable can penetrate from the driving arm pipe to the driven arm pipe in a sealed mode. The joint is reasonable and ingenious in structural layout, excellent in sealing performance and easy to decontaminate and maintain. The invention further provides operation equipment for the nuclear industry.

The invention discloses a nuclear power plant fuel operation accident source item result determination method and device and a nuclear power plant fuel operation accident alarm method and system, and relates to the technical field of nuclear industry. The method for determining the fuel operation accident source item result of the nuclear power plant comprises the following steps: acquiring a target decontamination factor corresponding to a target nuclide in a first time period and the total concentration of aerosol iodine retained in a spent pool of the nuclear power plant in a second time period; determining a first release amount corresponding to the target nuclide in the first time period according to the target decontamination factor; according to the total concentration of iodine in the aerosol, determining the share of iodine volatilized in the form of elements in the spent fuel pool in the second time period; determining a second release amount of the element iodine in a second time period according to the iodine share; and generating a fuel operation accident source item result of the nuclear power plant according to the first release amount and the second release amount. According to the embodiment of the invention, the method can improve the accuracy of determining the fuel operation accident source item result of the nuclear power plant.

Methods for processing metal components from the nuclear industry are described. In some embodiments, a method includes processing a component having a surface contaminated with radionuclides by dissolving at least a portion of the surface with an acid solution to produce a waste solution including the radionuclides, and separating the radionuclides from the waste solution. In some embodiments, a method includes processing a component including a target radionuclide by immersing at least a portion of the component in an electrolyte bath, applying an electric current to cause electrodissolution, and electroplating the target radionuclide from the solution onto a cathode. In some embodiments, a method includes processing a metal composite component by dissolving it to produce a solution including dissolved metal and particulate material, and separating the particulate material.

The application provides a spring bidirectional composite metal rubber damper and a working method thereof. The spring bidirectional composite metal rubber damper comprises a vertically arranged outer sleeve, one end of the outer sleeve is closed, the other end of the outer sleeve is internally provided with an inner sleeve, one end of the inner sleeve is closed, the other end of the inner sleeve is movably connected with a core shaft, an annular clamping shoulder is arranged in the middle of the inner sleeve, the annular clamping shoulder divides the inner sleeve into an upper damping cavity and a lower damping cavity, the shaft section of the core shaft located in the upper damping cavity is externally provided with an upper damping pad, the shaft section of the core shaft located in the lower damping cavity is externally provided with a lower damping pad, a lower connecting mechanism is arranged on the closed end of the outer sleeve, the end of the core shaft extends out of the open end of the inner sleeve, and an upper connecting structure is arranged on the extending end. The application is convenient for damping of aerospace, aviation, nuclear industry and military instruments, can solve the poor rigidity adjustability of the current single metal rubber damper, that is, it is difficult to realize the coordination between low rigidity and high durability, meanwhile, the structure has high lateral rigidity and compact structure.

The invention provides a recovery treatment method of fluoride slag in uranium hexafluoride preparation, belongs to the technical field of nuclear industry waste slag treatment, and solves the technical problems that a product obtained by an existing fluoride slag treatment method is high in metal impurity content and low in uranium recovery rate, secondary fluoride slag is easily generated, and the product cannot be returned to a uranium purification and conversion production line for reuse. The method comprises the following steps: at the temperature of 400-600 DEG C, contacting the fluorinated slag with excessive water vapor to carry out high-temperature hydrolysis reaction, so that UF4 and uranium intermediate fluoride in the fluorinated slag generate uranium oxides U3O8 and UO3 with the fluorine ion content of less than 1mg / g, and filtering to obtain uranium oxides; mixing uranium oxide with a nitric acid solution, dissolving and reacting to obtain a uranyl nitrate solution; the pH of the uranyl nitrate solution is adjusted to 3-4, hydrogen peroxide is added for a precipitation reaction, solid-liquid separation is conducted, and uranium peroxide solid precipitates are obtained. The method is high in uranium recovery rate, low in product impurity content and capable of returning to a uranium purification and conversion production line for reuse.

The invention relates to the technical field of anisotropic rubber, and particularly discloses anisotropic radiation-resistant rubber and a preparation method thereof, aiming at the problems that rubber used in nuclear industry such as shock absorption and isolation rubber components is always easy to crack, discolor, harden and the like due to the influence of a nuclear radiation environment at present, and particularly discloses the anisotropic radiation-resistant rubber and the preparation method thereof. Comprising 10%-30% of a rubber matrix, 65%-68% of irradiation-resistant particles, 1%-3.5% of an active agent, 1.2%-3.7% of an anti-aging agent, 3%-5% of a reinforcing agent, 0.8%-1.2% of a softening agent, 0%-0.2% of an accelerant and 0.5%-1.2% of a vulcanizing agent. Each radiation-resistant particle comprises a core layer, a middle layer and a shell layer, the core layer comprises lead tungstate and bismuth oxide, the middle layer comprises a biphenyl-nano SiO2 bonding body, and the shell layer comprises diphenyl acetylene grafted graphene oxide. By introducing the radiation-resistant particles with multiple structures and functional characteristics, the material performance stability of the anisotropic rubber in a radiation environment can be remarkably improved, so that the use requirements of the fields and scenes such as nuclear industry devices and the like are met.

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 invention discloses a preparation method of a composite aerogel for adsorbing cesium and strontium in wastewater. The composite aerogel prepared by the invention shows multiple technical advantages: due to a unique hierarchical pore structure, the composite aerogel has ultralow density and high porosity; rich surface functional groups of MXene and the ion screening effect of PB generate a synergistic effect, and selective removal of radionuclides such as Cs and Sr in wastewater is realized. Compared with traditional ion exchange resin and zeolite adsorbents, the adsorbent has the advantages of being large in adsorption capacity, short in adsorption time, high in adsorption efficiency, good in circulation effect and the like, and is particularly suitable for the fields of nuclear industry wastewater treatment, radioactive pollution emergency repair and the like.

The invention belongs to the technical field of nuclear industry equipment maintenance, and particularly relates to a nuclear power reactor bottom simulation experiment device which comprises a pressure vessel wall, a plurality of replaceable flux tube assemblies, a plurality of fixed flux tubes and at least one supporting leg. At least one supporting leg is installed at the edge of the bottom end of the pressure vessel wall and used for supporting the pressure vessel wall. The replaceable flux pipe assemblies and the fixed flux pipes penetrate through the wall of the pressure container and are detachably installed on the wall of the pressure container. The top end of each replaceable flux tube assembly is provided with at least one quick release connector. The quick-release connector is used for being externally connected with machining equipment. The simulation experiment device for the reactor bottom of the nuclear power reactor is arranged for an experiment instead of being operated on a reactor site, so that the technical problems that the neutron flux measurement precision is seriously influenced and potential safety hazards of a unit are generated due to the fact that weld joints of existing neutron flux tubes are prone to deformation and cracking are solved.

This application discloses a core power prediction method and apparatus, a core monitoring method and system, relating to the field of nuclear industry technology. The core power prediction method includes: acquiring target power detector data for a target core; the target power detector data includes values ​​from multiple power detectors; processing the target power detector data using a target power prediction model to obtain a predicted power value for the target core; wherein the target power prediction model employs a data grouping processing method, trained using historical power detector training data and their corresponding first historical power values. According to embodiments of this application, this core prediction method can monitor core power using a small number of detectors while ensuring the accuracy of core power prediction.

The invention discloses a preparation method and application of a titanium-based hydrogen-resistant coating structure on the surface of a metal hydride, the coating structure is composed of a metal hydride matrix and a titanium-based hydrogen-resistant layer, and the titanium-based hydrogen-resistant layer covers the surface of the metal hydride matrix; the chemical composition of the titanium-based hydrogen-resistant layer is titanium nitride. The preparation method comprises the following steps: (1) sequentially carrying out polishing treatment, chemical cleaning and drying treatment on a metal hydride base material to obtain a metal hydride matrix; (2) putting the metal hydride base material into a magnetron sputtering coating chamber, vacuumizing, and introducing argon to perform etching cleaning; and (3) after etching cleaning is finished, magnetron sputtering is carried out with a titanium target as a sputtering target material and nitrogen as reaction gas, and the titanium-based hydrogen-resistant coating structure on the surface of the metal hydride is obtained. When the metal hydride is used as a nuclear material in the high-temperature working condition of the nuclear industry or hydrogen storage equipment, the problem of performance degradation of the metal hydride caused by hydrogen diffusion can be solved.

The invention discloses preparation of a graphene-doped modified reverse osmosis membrane and a method for preparing deuterium-depleted water by using the graphene-doped modified reverse osmosis membrane, and particularly relates to the technical field of preparation of deuterium-depleted water. An aromatic polyamide separation layer containing functionalized graphene oxide is constructed on the surface of a polysulfone supporting layer through an interfacial polymerization method to form a conjugated large pi bond network; a 20-30MPa extra-high voltage driving process is combined, so that the deuterium and hydrogen isotope separation efficiency is remarkably improved. The single-stage separation coefficient of the reverse osmosis membrane reaches 1.8-2.0, the deuterium content of produced water is smaller than or equal to 50 ppm, and the reverse osmosis membrane has high pollution resistance, low energy consumption and long service life and is suitable for preparing high-purity deuterium-depleted water in the fields of medical treatment, semiconductors and nuclear industry.

The invention relates to the technical field of nuclear industrial robot control, and discloses an anti-radiation variable-stiffness mechanical arm which comprises a plurality of torsion modules, and every two adjacent torsion modules are connected through a rotating module, so that the torsion modules can spin around the longitudinal axes of the torsion modules and rotate around the horizontal axes on the rotating modules. A top cover is installed at the top of the uppermost torsion module, a bottom plate is installed at the bottom of the lowermost torsion module, the top cover and the torsion modules, the torsion modules and the torsion modules and the bottom plate are connected through a plurality of SMA artificial muscles, and wires are led out from the two ends of the SMA artificial muscles and connected into independent drive circuits. According to the anti-radiation variable-rigidity mechanical arm, control over the rigidity and deformation of the mechanical arm can be achieved, the size and weight of the mechanical arm are greatly reduced, and the power density is improved.

The invention provides an isotope quantitative analysis method and system based on laser-induced breakdown spectroscopy, and relates to the technical field of laser spectrums.The method comprises the steps that a liquid analysis environment is constructed according to the form of a sample to be detected; focusing a pulse laser beam in the liquid analysis environment to ablate a sample to be detected so as to generate plasma; collecting an emission spectrum signal generated in the plasma cooling process; and extracting an isotope characteristic spectral line of a target element from the emission spectrum signal, and carrying out isotope quantitative analysis based on spectral line information of the isotope characteristic spectral line. According to the device and the method, spectral line broadening and self-absorption can be effectively inhibited in mechanism, the resolution ratio of the isotope spectral line is remarkably improved, the device is simple, the cost is low, and a reliable and economical solution is provided for realizing high-performance isotope in-situ and rapid quantitative analysis in the fields of nuclear industry and the like by a laser-induced breakdown spectroscopy technology.

The application relates to the technical field of alloy materials, and discloses a high-temperature-resistant ternary nickel-based titanium-niobium alloy with a service temperature of 2500 DEG C and a preparation method thereof, which takes nickel, titanium and niobium as core components, and the atomic percentage composition is as follows: 44-46% of nickel, 43-45% of titanium and 9-11% of niobium; and 0.1-0.5% of one or more trace elements of oxygen, aluminum and chromium can be further added. Through optimization of component design, the alloy is thermodynamically stable under an ultrahigh-temperature environment of 2500-2900 DEG C. The preparation method adopts vacuum consumable arc melting or electron beam melting (vacuum degree >= 10 ‑3 Pa), melting, and solves problems such as component segregation and poor machining performance. The alloy is suitable for aerospace and nuclear industry ultrahigh-temperature parts, and the process is stable and easy to mass-produce.

The invention discloses a reflecting layer equivalent model determination method and device and a nuclear design calculation method and system, and relates to the technical field of nuclear industry. The reflecting layer equivalent model determination method comprises the following steps: acquiring an initial reflecting layer model of a reactor and a reactor core neutronics characteristic theoretical value; according to the theoretical value of the neutronics characteristic of the reactor core, adjusting initial information of the initial reflecting layer model so as to simplify the initial reflecting layer model and obtain a target reflecting layer equivalent model; according to the embodiment of the invention, the method can simplify the equivalent model of the reflecting layer under the condition that the theoretical value of the neutronics property of the reactor core meets the requirement, and further can guarantee that the calculation accuracy and calculation efficiency of a deterministic reactor core program meet the nuclear design requirement.

The invention relates to the technical field of material distribution equipment, and provides a powder equally-dividing tool and equally-dividing method.The powder equally-dividing tool comprises a base, a material dividing box, a material storage container, a subpackaging container and a rotor; a feeding hole is formed in the upper end of the material distribution box, and a discharging hole is formed in the lower end of the material distribution box; the storage container is detachably connected to the upper end of the distribution box and communicated with the feed port, and the split charging container is detachably connected to the lower end of the distribution box and communicated with the discharge port; the rotor is rotationally connected into the material distribution box, a material storage cavity is formed in the rotor, the volume of the material storage cavity is consistent with that of the split charging container, and the rotor can rotate to switch the opening of the material storage cavity to face the feeding port or the discharging port. Powder flying dust leakage is avoided through a closed structure, the equal division precision is guaranteed through volume matching of the material storage cavity and the split charging container, remote mechanical arm operation is adapted, the problems that in the prior art, sealing protection is insufficient, and the equal division precision is poor are solved, and the high-precision operation requirement of the nuclear industry and other fields is met.

The invention belongs to the technical field of personnel efficiency prediction and selection, and particularly relates to a behavior efficiency prediction method based on psychological characteristics. According to the method, corresponding psychological indexes and behavior efficiency evaluation indexes are established for nuclear power master control room operators, correlation analysis is adopted to explore the relationship between psychological quality and behavior efficiency, and a behavior efficiency prediction model based on psychological quality is established based on a multiple regression equation. The technical problem of how to improve the working efficiency of a nuclear power plant and further prevent human errors by improving the adaptability of psychological conditions and post requirements of nuclear industry personnel is solved.

The invention discloses a device and a method for preparing a tungsten-copper particle gradient composite material through vibration induction, and belongs to the crossing field of powder metallurgy and material forming machining. Aiming at the problems of powder agglomeration, non-uniform component distribution, difficulty in accurate control of a gradient structure and poor interface continuity in the preparation of a traditional tungsten-copper gradient composite material, a powder mixing-vibration mold filling-hot pressing sintering integrated process is adopted; tungsten copper powder distribution is optimized through a vibration external field capable of being precisely regulated and controlled. High-frequency small-amplitude vibration breaks particle aggregation so as to improve the filling uniformity, and low-frequency large-amplitude vibration guides tungsten copper two phases to form a preset gradient. The preparation device is used for completing powder homogenization and gradient regulation and control, and the tungsten-copper gradient composite material is prepared in cooperation with follow-up spark plasma sintering. According to the method, the vibration external field is only applied in the tungsten-copper particle mold filling stage, the process complexity does not need to be increased, the microstructure uniformity of the tungsten-copper composite material can be remarkably improved, and the gradient structure transition continuity is enhanced. The prepared composite material has the characteristics of high temperature resistance and low expansion of tungsten and high heat conductivity and high electric conductivity of copper, and can be widely applied to the high-end manufacturing fields of electronic power, aerospace, nuclear industry and the like.

The invention relates to the field of fluorine chemical processes, in particular to a molybdenum hexafluoride sampling device and method. The device comprises a molybdenum hexafluoride steel cylinder, an inert gas steel cylinder, a buffer bottle, a first-stage absorption bottle, a second-stage absorption bottle and a third-stage absorption bottle, the absorption liquid is filled into the three-stage absorption bottle and is weighed, the pipeline is vacuumized and replaced after the three-stage absorption bottle is connected, the absorption bottle is purged, and the buffer bottle is vacuumized. The method comprises the following steps: heating a molybdenum hexafluoride steel cylinder, filling the molybdenum hexafluoride steel cylinder into a buffer bottle, filling inert gas, introducing the inert gas into an absorption bottle for absorption, purging a pipeline, weighing the absorption bottle, and detecting metal impurities by using ICP-OES (Inductively Coupled Plasma-Optical Emission Spectrometer). According to the method, the problem that metal impurities in high-purity molybdenum hexafluoride are difficult to safely and accurately sample is solved, and a reliable and practical solution is provided for purity quality control of the key material molybdenum hexafluoride in the fields of semiconductors, nuclear industry and the like.