146 results about "Nuclear safety and security" patented technology

The invention provides a nuclear safety shell surface defect automatic detection method and system, and provides a systematic and effective technical scheme to overcome the defects in a nuclear safety shell surface, including cracks, corrosion and seepage. The method comprises: image preprocessing, and then utilizing a contour expansion filtering algorithm to perform denoising enhancement to obtain pretreated binary images; feature extraction comprising utilizing a linear weighting tensor voting algorithm, extracting cracks from discrete noises through detecting high light degree to obtain a result probability graph; and performing object tracking to obtain a crack center line, back-projecting positioning to extract real crack information, measuring crack length and stable width, crack integration and other defect detection and storing information. The invention further provides utilizing an adaptive width template method and an iterative method to respectively binarize images, and obtaining final binary images by performing AND operation on two results. The technical scheme creates an own system, and can meet real engineering demands, and guarantee nuclear shell safety.

The invention relates to a method for measuring the parameter of an omega welding seam defect, in particular to a method for measuring the parameter of an omega welding seam defect, which is applied to a control rod drive mechanism of a nuclear power plant. The method adopts a welding technique estimation test piece of the omega welding seam to manufacture a contrast test block, adopts an automatic detection device containing an ultrasonic detection method and a vortex detection method to measure the parameter of the defect and judges the parameter of the defect integrally. The method can effectively measure the parameter of the omega welding seam defect of the control rod drive mechanism of the nuclear power plant accurately, provides data support for the preventive maintenance of the control rod drive mechanism, provides guarantee for the nuclear safety and is suitable for detecting the parameter of the welding seam defect in nuclear industry system.

The invention discloses an online monitoring apparatus for monitoring influence of nuclear power plant containment fragments upon pressure drop of a fuel assembly and belongs to the technical field of nuclear safety. The online monitoring apparatus comprises a test segment, a full-size fuel assembly, an armored thermocouple, a differential pressure gauge, a pressure meter electric V-shaped ball valve, a flow meter, and a stirring water tank. By simulating the fact that fragments from an accident in a pressurized water reactor nuclear power station penetrate a containment pit strainer and comprehensively utilizing the measuring instruments such as a temperature sensor, a pressure sensor, a differential pressure sensor and a flow meter, the pressure drop of the fuel assembly corresponding to different conditions and fragment quantities is monitored online; by monitoring operation of the apparatus in test, it is possible to study the distribution, attachment and blockage of fragments in the fuel assembly and quantitatively evaluate the influence of fragments in the containment after LOCA (loss of coolant accident) of the pressurized water reactor nuclear power plant, and support is provided to ensure reliable execution of safety functionality of an emergency core cooling system of the pressurized water reactor nuclear power plant.

The invention discloses an IEEE1500 standard based IP nuclear measuring transmission component and a control method thereof and relates to the field of the system chip IP nuclear measuring transmission component and the control method thereof, in order to solve the problems of increasing the functional path delay and extra area expense of the traditional IP nuclear safety control measuring shell unit and the control method thereof in the process of safety translocation. First and second MOS pipe components are in parallel connection, wherein the source electrode and the drain electrode of the first MOS pipe component are respectively connected to the source electrode and the drain electrode of the second MOS pipe component; the grid electrodes of the first and second MOS pipe components are respectively connected to a first control signal GC and a second control signal GC; and the first and second control signals are inverted control signals. The invention has the advantages of short time delay and low extra area expense. The component and the control method of the invention are fit for various occasions requiring the system chip IP nuclear measurement.

The invention discloses a method for completeness evaluation of nuclear power plant debugging test projects. The completeness evaluation is performed in four dimensions including a nuclear safety regulatory guideline dimension, a system and machine unit function decomposition dimension, an accident analysis dimension and other reactor type nuclear power plant transverse comparison dimensions; through the four-dimension test project completeness evaluation, the obtained test projects are integrated; relatively complete test projects can be obtained. The test projects evaluated by the invention conform to the requirements of the Chinese nuclear safety regulatory guideline; the verification on the response event sequence of accidents in the nuclear power plant safety probability analysis is included; the important significance is realized on the overall verification of the nuclear power plant design function and requirements and the guarantee of safe and reliable operation of the machine unit.

The invention relates to the technical field of nuclear reactor thermal hydraulic design and safety analysis, and particularly discloses a CHF relational expression DNBR limit value statistical determination method based on a correction method. The CHF relational expression DNBR limit value statistical determination method specifically comprises the following steps: 1, acquiring CHF experimental data of a fuel assembly; 2, obtaining M / P data of an experiment burning point position; 3, carrying out Bartlett inspection on M / P data at the position of the experimental burning point; 4, carrying out homogeneity test of a data mean value; 5, carrying out normal distribution inspection; 6, determining a DNBR limit value by using an Owen criterion; 7, when the M / P data cannot pass any one of the steps 3 to 5, correcting the degree of freedom by utilizing Sterhware; and 8, substituting the correction degree of freedom obtained in the step 7 into an Owen coefficient expression to solve and obtain an Owen coefficient so as to determine a DNBR limit value. According to the CHF relational expression DNBR limit value statistical determination method, a rigorous, accurate and relatively conservative CHF relational expression DNBR limit value can be obtained, and key parameters can be calculated for CHF relational expression development and CHF experimental data evaluation, and the most concerned design limit value is provided for a nuclear safety department.

The invention relates to a method and a device for detecting and setting a default value of a nuclear power plant digital control system. The method comprises the following steps of: detecting a network fault signal which is received by a non-safe level digital meter control platform, and setting the corresponding default value to be 0; detecting a hard wire signal, and after determining that the hard wire signal is an ineffective signal, setting the default value corresponding to the hard wire signal; and detecting a sensor signal, and after determining that the sensor signal is the ineffective signal, setting the default value corresponding to the sensor signal. The network signal, the hard wire signal and the sensor signal are subjected to classification detection and default value setting on the basis of a safe level digital meter control platform and the non-safe level digital meter control platform of the nuclear power plant digital control system, and the default value is set by combining the nuclear safe particularity of the nuclear-power industry, so that the signal judgment and the parameter setting work of the default value are convenient and optimized.

The invention discloses a method and a device for detecting the concentration of uranium-containing liquid in uranium-containing pipelines. The method comprises the following operating steps: (a) establishing a counting rate calculation model of gamma rays emitted by the uranium-containing liquid in the uranium-containing pipelines; (b) injecting the uranium-containing liquid with different uranium concentrations into the uranium-containing pipelines to obtain counting rates of the gamma rays under the different concentrations; (c) establishing a relation curve between the counting rates and the uranium concentrations by using the uranium concentrations of the uranium liquid as horizontal ordinates and using the counting rates of the gamma rays corresponding to all the uranium concentrations as vertical coordinates; (d) acquiring a standard curve of the uranium concentrations based on the least square fit of the relation curve obtained in the step (c); (e) determining the counting rateof the gamma ray emitted by the uranium-containing liquid to be tested according to the counting rate model obtained in the step (a) and the standard curve of the uranium concentrations obtained in the step (d). According to the technical scheme, the detection method adopts real-time detection, is short in whole detection time and simple to operate, has significantly improved stability and accuracy, and meets the requirements for nuclear safety monitoring in pilot scale and mass production.

The invention discloses a nuclear power plant water loss accident safety injection flow demand analysis method based on a genetic algorithm. The method comprises the steps of considering the importantphysical phenomenon influencing the cladding temperature evolution process in the transient process of a water loss accident; searching a genetic algorithm of an optimal solution based on a simulatednatural evolution process; analyzing the program based on the water loss accident; realizing the process of automatically searching for the optimal safety injection flow demand under the nuclear power plant water loss accident. Compared with the safety injection flow is provided by an existing nuclear power plant safety injection system, the relationship between important parameters in the reactor core and the safety injection flow under a water loss accident is considered, the process of automatically searching the optimal safety injection flow demand is realized, and the water loading amount of key equipment in the safety injection system can be reduced under the condition that the reactor meets the nuclear safety criterion, so that the construction cost of the key equipment in the safety injection system is reduced.

The invention belongs to the technical field of pressurized water reactor nuclear power stations, and particularly relates to a heat-electricity-water combined system suitable for a large pressurized water reactor nuclear power unit and a production process. In the invention, a first nuclear power unit and a second nuclear power unit forms a nuclear steam supply system; a superheater, a steam generator, a deaerator, a secondary feed water heater, a primary feed water heater and a secondary water feeding pump form an industrial steam production system; a desalted water feeding pump, a water feeding tank, a desalted water facility, a seawater desalination facility, a heater, a first regulating valve, a second regulating valve, a seawater taking port, a seawater draining port, a standby fresh water taking port and a production water user form a seawater desalination and desalted water supply system; and a drainage pipeline returns to a pressurized water reactor nuclear power unit condenser through a third regulating valve and a fourth regulating valve to form a steam condensation water system. On the premise of ensuring nuclear safety, a nuclear power unit has the capabilities of producing industrial steam and producing fresh water through a seawater desalination technology.

The invention discloses a nuclear power plant overhaul safety risk management tool development method. The method comprises the following steps of 1, selecting a key safety function; 2, analyzing theoperation state of a power plant; 3, determining risk characterization; 4, establishing a safety function analysis tree; 5, establishing a system fault tree; and 6, creating a risk management guide rule. The method has the beneficial effects that (1) the method is provided for overhaul planners, so that the optimal time period for the system to quit operation is determined from the perspective ofnuclear safety, and the overhaul plan is optimized; (2) the method is provided for overhaul planners, and overhaul stages with high nuclear safety risks are systematically determined from the perspective of safety so as to make appropriate prevention measures and emergency plans; (3) an operator is helped to identify the nuclear safety risk in the overhaul period, and execute prevention measures and emergency plans under the condition of relatively high risk; and (4) the employee is helped to independently examine the overhaul plan from the perspective of nuclear safety.

The invention belongs to the technical field of nuclear power instrument control systems. For solving the technical problems that in the prior art, a security control display device is large in installation space and cannot be directly operated, and a nuclear safety level device is controlled, the invention provides a nuclear security level control display device for a nuclear power station. The nuclear security level control display device comprises a controller, a power supply input circuit, a display, a memory and a frame for integrally connecting the controller, the power supply input circuit, the display and the memory, wherein the security level platform software is stored in the memory; and the controller loads and operates the platform software in the memory, and realizes data display and graphic display of engineering design and direct control of nuclear security level equipment through the platform software. Therefore, the integrated structural design can be realized, and theinstallation space is saved, and the nuclear security level device can be directly operated and controlled.

The invention belongs to the technical field of nuclear safety control and relates to a reactor pit direct retention type reactor core melt capturing device of. The capturing device comprises a reactor pit, an overhauling platform, a reactor pressure container, metal sealing pots, a cooling water tank and a steam exhausting opening, wherein the overhauling platform is placed in the reactor pit andis used for dividing the reactor pit into upper and lower parts; the reactor pressure container is arranged at the upper half part in the reactor pit; the plurality of metal sealing pots are stackedon the lower half part in the reactor pit; after the lower half part in the reactor pit is filled with water, the metal sealing pots can be of a floating state in the reactor pit; the cooling water tank is located outside the reactor pit and forms a circulating loop with the lower half part in the reactor pit through a connection pipeline; the steam exhausting opening is formed in the upper part of a side wall of the lower half part in the reactor pit. The capturing device provided by the invention has a simple principle, is convenient to construct and amount, is easy to maintain and has highcooling efficiency and small negative risks and can cope with serious accident working conditions of the reactor.

The invention discloses a fatigue monitoring and life evaluation system for a nuclear power plant. The system comprises system hardware, a system platform, a calculation program and a system database;the system hardware is composed of a system server, a database server, a backup server and a network switch; the system platform is composed of a man-machine interaction interface and a system management service system, wherein the system management service system is composed of data acquisition, data storage, data processing, parameter display, data retrieval, trend display, report formulation,reference data and system management; and the calculation program is composed of measurement point screening, NCR evaluation, temperature field analytical solution, stress field analytical solution and test verification. The system of the invention is wide in monitoring range and covers all main equipment and main pipelines of the primary loop of a nuclear island. No increase in hardware instrument measuring points is innovatively realized; the temperature state of a concerned position is obtained through a model derivation method; and therefore, the minimum transformation and the highest efficiency of an old power plant during the application of the old power plant are ensured. The influence of a pressurized water reactor coolant environment on metal fatigue is innovatively considered; and the requirements of a nuclear safety bureau on a power plant charging license are met.

The invention discloses a rapid joint analysis method for Pu-239, Sr-90 and Cs-137 in waste liquid. The rapid joint analysis method comprises the following steps of: precipitating plutonium and strontium by controlling a pH value of wastewater, and separating then from cesium; dissolving plutonium and strontium precipitates, meanwhile, adjusting a valence state of Pu to Pu(IV), separating 239Pu and 90Sr through using a TEVA resin and Sr resin double-column series connection method, and measuring 239Pu and 90Sr through using ICP-MS and a liquid scintillation counter respectively; and collectingall liquid in the plutonium and strontium precipitation process, directly enriching 137Cs through using KNiF-PAN resin, naturally airing the resin, and measuring 137Cs by using a gamma spectrometer.The rapid joint analysis method has the advantages of small sample size, simple process, short experiment period, high separation efficiency and the like, and a feasible and efficient technical meanscan be provided for rapid analysis of nuclide in nuclear emergency monitoring and nuclear safety supervision.

The invention belongs to the technical field of nuclear safety instrumentation and control system. In order to solve the technical problems of insufficient visualization and scalability of the known image configuration realization scheme in the nuclear power industry in the prior art, the invention provides a graphics file generation method and system for nuclear safety level application. The method comprises the following steps: S1, identifying each primitive from a graphical picture configured by a user, and identifying the characteristics of each primitive; Each of the primitive characteristics includes a static attribute and / or a dynamic attribute, respectively; S2, defining a graphics file format corresponding to each primitive based on the static attribute and / or the dynamic attribute of each primitive; S3, generating a corresponding total graphic file according to the graphic file format corresponding to each graphic element in the graphic screen configured by the user.

The invention relates to a method for executing safety-related system and equipment periodic test supervision requirements of a nuclear power plant. The method comprises the steps of: dividing the acceptance criterion of each supervision requirement into two types: an availability criterion and a normal operation criterion according to whether the availability requirement is affected or not when the acceptance criterion of supervision requirements is not met; when a periodic test is executed and the availability criterion is not met, executing corresponding measures of operation limit values and conditions; and when the normal operation criterion is not met, performing self-control in the nuclear power plant, taking necessary maintenance measures without the need of executing measures of operation limit values and conditions. The method for classifying the acceptance criterion of the supervision requirements provided by the invention is simple, highly-efficient and convenient to use, avoids the situation that different nuclear power plants execute different periodic test strategies inconsistently, is convenient for nuclear safety supervision and power plants to use, and also avoidsthe operation limitation of a nuclear power unit due to the fact that equipment is improperly judged to be unavailable and enter LCO.

The invention discloses an authority management system for a nuclear power plant safety level DCS, which comprises an electronic key, a central user database, a mirror image database and a maintenancenetwork, and is characterized in that the central user database is arranged at an engineer station; the mirror image database is arranged in each function station; the engineer station is in communication connection with each function station through a maintenance network; an electronic key is used as a port for a user to initiate an authorization request; after the user inserts the electronic key into host equipment of a function station, the host equipment of the function station reads an electronic key identification code of the electronic key and encrypted relevant information of the userand carries out authentication in a mirror image database of the function station; and corresponding operation authority is granted to the user after the authentication is passed, and if the user operation exceeds his / her own authority range, it is determined that the operation is invalid and a prompt is given. Corresponding operation authorities are granted to people with different identities, the unauthorized operation condition of the nuclear safety level DCS is overcome, and system maintenance is facilitated.

The invention relates to the technical field of nuclear safety and nuclear system detection, in particular to a pressurized water reactor nuclear power station primary circuit total gas content measuring device which comprises a shielding protective cover, and a sampling assembly and a sample measuring assembly which are detachably connected with the shielding protective cover. The sampling assembly comprises a sampling bottle; a valve I connected with a bottle opening of the sampling bottle through a pipeline; a valve II connected with the other bottle opening of the sampling bottle through apipeline; a first quick female joint connected with the valve I through a pipeline; a second quick female joint connected with the valve II through a pipeline; and a balance pipeline, wherein one endof the balance pipeline is connected with the pipeline located between the valve I and the first quick female joint, and the other end of the balance pipeline is connected with the pipeline located between the valve II and the second quick female joint. The total gas content measuring device is convenient to disassemble and assemble, convenient to carry, capable of rapidly sampling and accuratelymeasuring the total gas content of the primary loop, and can be effectively used for detecting the gas content of the coolant in the primary loop.

The invention relates to the technical field of nuclear reactor thermal hydraulic design and safety analysis, and particularly discloses a CHF relational expression DNBR limit value statistical determination method based on a grouping method. The method comprises the following steps: 1, acquiring CHF experimental data of a fuel assembly; 2, obtaining M / P data of an experiment burning point position; 3, carrying out a Bartlett test; 4, after passing the Bartlett test, carrying out homogeneity test of a data mean value; 5, carrying out normal distribution inspection by adopting an Epps-Pulley inspection method; 6, determining a DNBR limit value by using an Own criterion; 7, performing W-M-W inspection and K-W unilateral variance analysis respectively; 8, determining a unilateral 95 / 95 limitvalue of free distribution, and obtaining a DNBR limit value; 9, after the Bartlett test is not passed, carrying out a homogeneity test of a data mean value; 10, performing data grouping inspection; 11, determining a DNBR limit value. According to the method, a rigorous, accurate and relatively conservative CHF relational expression DNBR limit value can be obtained, key parameters can be calculated for CHF relational expression development and CHF experimental data evaluation, and the most concerned design limit value is provided for a nuclear safety department.

The invention relates to a nuclear power plant supervision requirement optimization method. The method comprises the steps of obtaining operation technical specifications and supervision items of safety related systems and equipment of a nuclear power plant; performing comparative analysis on the operation technical specifications and the supervision items to obtain a comparative analysis result; determining a supervision category according to a comparative analysis result; and performing argumentation analysis based on the supervision category to obtain a supervision requirement optimization result. By optimizing the nuclear power plant supervision requirements, the problems of inconsistent supervision content and operation technical specification requirements, complicated content, unclear hierarchy and repeated verification of the existing supervision requirements can be effectively solved, safety important supervision is further highlighted, management omissions are found and made up, the safety management level of a power plant is improved. Meanwhile, supervision content is simplified, and autonomy and flexibility of nuclear safety management of the power plant are improved.

The invention belongs to the technical field of nuclear safety evaluation, and relates to a method for evaluating the influence of nuclide content on keff uncertainty by Monte Carlo method. The methodis based on the Monte Carlo method and comprises the following steps of: (1) experimental verification: the experimental data are verified by using a fuel consumption simulation program, and the measured value of the nuclide composition content given in each experiment is compared with the calculated value of the nuclide composition content simulated by the fuel consumption program; (2) determining the distribution of the verification results of each nuclide and adjust the standard deviation of the mean value; (3) performing Monte Carlo sampling calculation; (4) performing statistical analysis of the results of sampling calculation; (5) determining the total uncertainty. The method for evaluating the effect of nuclide content on the keff uncertainty by using the monte carlo method of theinvention can make the calculation of the critical limit value in the nuclear critical analysis more accurate.

The invention belongs to the technical field of biological excretion promotion, and particularly relates to a difunctional heteropolyacid excretion promoting agent which is used for nuclide excretion promotion and ROS removal, and the difunctional heteropolyacid excretion promoting agent comprises phosphotungstic heteropolyacid salt. The novel phosphotungstic heteropolyacid salt molecular cleanup promoting agent is adopted, has excellent biocompatibility, mild biotoxicity and good cleanup promoting effect and ion selectivity on uranyl, and compared with ZnNa3-DTPA, the novel phosphotungstic heteropolyacid salt molecular cleanup promoting agent can remarkably improve the cleanup promoting rate of uranium in kidneys and bones; while uranium is removed, the ROS level rise, caused by radionuclide, in an organism can be effectively reduced, a brand-new and efficient radionuclide cleanup promoting agent is successfully developed, and important practical significance is achieved for safe and efficient development of nuclear energy and nuclear safety emergency.

The invention relates to a buffer base suitable for ocean nuclear power platform equipment. The buffer base comprises a buffer module, the buffer module comprises a base panel and a supporting structure thereof, a fixed rigid base and a buffer mechanism, the base panel and the supporting structure thereof comprise a base panel, a main supporting column, a plurality of T-shaped sections and a bottom circular pressing plate, and the T-shaped sections are evenly and symmetrically arranged around the main supporting column in the circumferential direction; the top faces of the main supporting column, the T-shaped profile web and the T-shaped profile panel are all connected with the base panel. The bottom circular pressing plate is provided with a through hole, the bottom face of the T-shaped profile is connected with the bottom circular pressing plate, the main supporting column penetrates through the bottom circular pressing plate through the through hole, and the bottom face of the mainsupporting column is connected with the fixed rigid base through the buffer mechanism. According to the buffer base, the influence of inertia force generated by ship movement on nuclear power platformnuclear safety equipment can be reduced, energy brought by impact is absorbed, and the stable operation environment of the nuclear safety equipment is guaranteed.

The invention relates to the field of safety monitoring of major civil and structural engineering and nuclear safety, in particular to a method for evaluating performance of a containment based on a white light interference sensing technology. The method comprises the following steps that sensing optical fiber is laid on the containment; a true strain value and radial displacement value of a cylinder are acquired; a theoretical strain value and radial displacement value of the cylinder are acquired; performance evaluation is carried out on the containment, and it is judged whether or not the containment meets the requirement of the overall strength. The method for evaluating the performance of the containment based on the white light interference sensing technology has the advantages thatthe overall performance evaluation can still be carried out under the situation of the failure of a pre-buried vibration wire sensor of a concrete structure of the containment, multi-area laying, longdistance, high precision, large data volume and visualizability are achieved, the accidental error of a local position can be eliminated, the upgrading and reconstruction needs of an existing pre-buried strain monitoring sensor before supplementing and failure, and guarantee is provided for the operation of long service life of a nuclear power plant.

The invention discloses a diversity protection system and method for an ocean nuclear power platform. The diversity protection system comprises a diversity driving cabinet and a diversity driving disc, wherein the diversity driving cabinet comprises a first signal receiving module and a judgment module which are connected with each other, and the diversity driving disc comprises a second signal receiving module electrically connected with the judgment module. The invention relates to the technical field of nuclear safety protection of nuclear reactors. With the diversity protection system andmethod in the invention, when anticipated transient and design basis accident occur in a reactor and a reactor protection system is out of operation, the reactor and a steam turbine of the ocean nuclear power platform are shut down emergently, two paths of signals independently collected from the reactor and a reactor protection system are respectively adopted for comprehensive judgment, so unnecessary shutdown of the reactor and the steam turbine caused by signal errors are prevented, and the safety, economical efficiency and reliability of the diversity protection system are realized.

The invention relates to the technical field of nuclear safety and nuclear emergency, particularly to a nuclear power plant accident online diagnosis and state tracking prediction method. The existingnuclear power station diagnosis, evaluation and prediction system has the defects that the system seriously depends on a sample library, a diagnosis result is not accurate enough, and accident progress prediction is not enough. The method comprises the following steps: 1, identifying the state of a power plant; 2, identifying accident type / sequence; 3, carrying out arbitrary initialization; 4, carrying out a thermal hydraulic analysis program; and 5, carrying out comparing treatment. The method is comprehensive in accident diagnosis information, quick in response and accurate in judgment, canpredict three intervention schemes at the same time, and is high in efficiency.

The invention relates to the technical field of containment overall leakage rate tests, and particularly relates to a pressure relief system and method for an integral leak rate test of an AP1000 containment. The pressure relief system for the integral leak rate test of the AP1000 containment comprises a pressure relief valve I, a pressure relief valve II, an isolation valve, a tee joint I, a butterfly valve, a tee joint II, a VFS exhaust air filter unit and a temporary pressure relief hose, wherein the pressure relief valve I and the pressure relief valve II are arranged in parallel; the teejoint I is arranged on a discharge pipeline at the downstream of the pressure relief valve I and the pressure relief valve II; the butterfly valve is arranged on the discharge pipeline between the teejoint I and a chimney; the tee joint II is arranged on the exhaust pipeline between the downstream of the isolation valve and the chimney; the VFS exhaust air filter unit is arranged on the exhaust pipeline between the tee joint II and the chimney; one end of the temporary pressure relief hose is connected with the tee joint I, and the other end of the temporary pressure relief hose is connectedwith the tee joint II. According to the pressure relief system for the integral leak rate test of the AP1000 containment, the VUS and the VFS are combined to change a pressure relief flow channel forthe integral leak rate test of the unit overhaul containment, the purpose of purifying pressure relief air is achieved, and thus the nuclear safety guide rule is met.

The invention discloses an application of a hydrogel bead material which is obtained by copolymerizing acrylic acid and acrylonitrile, performing amidoximation and then performing ionic crosslinking, and the hydrogel bead material is used for adsorbing uranium in a high-fluorine system. The material disclosed by the invention is an adsorption material which is low in cost, good in chemical stability, easy to separate and capable of adsorbing uranium in the high-fluorine system, has an excellent adsorption effect, can successfully realize separation of uranium and fluorine, reduces the radioactive content of fluorine tailings, successfully realizes clean management and control and recycling of resources; The preparation method of the material is simple and convenient to operate, the production process is easy to control, and large-scale application can be achieved; after uranium adsorption, the material is automatically separated from the system, centrifugation and other operation are not needed, and operability is high; the material is particularly suitable for treating uranium-containing wastewater in a fluorine-containing environment in nuclear industry enterprises in China; and uranium can be successfully and selectively separated from the fluorine-containing wastewater through the material, and important technical support is provided for solving the current nuclear potential safety hazard problem.

The invention relates to the technical field of nuclear safety and nuclear system detection, in particular to a method for measuring total gas content of a pressurized water reactor nuclear power primary circuit, which comprises the following steps: S01, sampling a primary circuit coolant through a sampling assembly; s02, measuring gas content measurement data in the sample coolant through a sample measurement assembly and a sampling assembly; and S03, calculating the total gas content of the sample coolant according to the gas content measurement data and a gas content measurement formula. The method for measuring the total gas content of the pressurized water reactor nuclear power primary loop is simple and convenient to operate, and can be used in cooperation with the measuring device to quickly and accurately measure the gas content in the primary loop, so that the degassing effect of a primary loop coolant is conveniently judged, and safe and efficient operation of a main pump isensured.