2504 results about "Nuclear plant" patented technology

The invention discloses an intelligent management system for nuclear power plant personnel situation prediction and risk assessment, and relates to the field of intelligent safety management systems, and the system comprises a data collection unit, a multi-dimensional situation awareness unit, a risk prediction and assessment unit, an intelligent decision intervention unit and a visual interaction unit. And multi-source data acquisition, real-time situation construction, dynamic risk prediction and evaluation, intelligent early warning intervention and information visualization are realized. According to the invention, real-time monitoring, dynamic risk prediction and intelligent management of the safety state of the operating personnel can be realized, and the defects of real-time monitoring, dynamic prediction and intelligent management of the operating personnel in a high-risk area in the prior art are overcome, so that the safety management level is improved, the life safety is guaranteed, and the accident occurrence probability is reduced.

The invention relates to the technical field of nuclear power plant equipment inspection and data transmission management, and provides a nuclear power plant equipment inspection data transmission management method and system. According to an existing nuclear power plant equipment inspection system, reliable transmission and intelligent processing of data are difficult to achieve in complex high-radiation, high-humidity and strong-electromagnetic-interference environments, the data are prone to interference in the inspection process, and the fault detection precision is insufficient. The device state information is dynamically collected in real time through the multi-modal data collection technology, the sampling frequency is adjusted according to the environment change through the fuzzy logic control algorithm, and the timeliness and accuracy of data are improved. After data preprocessing, high-dimensional fusion and feature extraction of multi-modal data are realized through a Bayesian network and a deep neural network (DNN). The system realizes intelligent inspection and data transmission management of nuclear power plant equipment in a high-interference environment, has high reliability and real-time adaptability, and can be widely applied to an intelligent safety inspection system of a nuclear power plant.

The invention discloses a nuclear power plant radiation environment monitoring method and system, and the method comprises the steps: collecting the real-time monitoring data of a nuclear power plant radiation environment, and obtaining an environment monitoring real-time data set; performing dynamic threshold comparison on the environment monitoring real-time data set to generate an abnormal identifier or a normal feature vector; under the condition that the abnormal identifier is not generated, inputting the normal feature vector into a pre-trained environment trend prediction model, and outputting a radiation environment change trend curve and a potential abnormal probability in a future preset time period; and based on the change trend curve and the potential abnormal probability, generating graded early warning information, and finally outputting a monitoring early warning result including an early warning grade, prediction time efficiency and a coping strategy. According to the embodiment of the invention, the method can achieve the intelligent monitoring of the whole process of the radiation environment from the abnormal recognition to the prediction intervention, and improves the early warning accuracy, timeliness and decision support capability.

The invention relates to the technical field of nuclear power internal disaster analysis, and discloses a nuclear power plant internal flooding experiment verification method, device, equipment and medium, and the method comprises the steps: carrying out the modeling analysis of a target plant prototype, and determining a plurality of model parameter values of an experiment scaling model; constructing an experiment scaling model according to the plurality of model parameter values, and determining experiment parameters of an internal water logging experiment; performing an internal water logging experiment on the experimental scale model by using the experimental parameters to obtain an internal water logging experiment result of the experimental scale model; and obtaining an internal water logging simulation result of the numerical model, comparing the internal water logging simulation result with an internal water logging experiment result, and verifying the accuracy of the internal water logging simulation result according to a comparison result. According to the method, the experiment scale model of the nuclear power plant prototype is constructed, the flooding spreading conditions under different working conditions are simulated and compared with the flooding simulation result, the accuracy of flooding simulation calculation is verified, and more accurate guidance is provided for follow-up nuclear power project internal flooding analysis.

The invention particularly relates to a method for predicting the service life of a boric acid solution corrosion concrete member in a nuclear power plant. The method comprises the following steps: determining boric acid solution accelerated corrosion concrete member test control parameters; based on the boric acid solution accelerated corrosion concrete member test control parameters, boric acid solution accelerated corrosion concrete member tests are carried out in groups; constructing a boric acid solution corrosion concrete member life prediction model based on boric acid solution accelerated corrosion concrete member test data; and predicting the boric acid solution corrosion concrete member service life based on the boric acid solution corrosion concrete member service life prediction model. According to the method, a model for predicting the service life of the concrete member corroded by the boric acid solution is established by considering the water-binder ratio of the concrete, the temperature, the concentration of the boric acid solution and the corrosion time on the basis of research test results of physical properties and macroscopic mechanical properties of the concrete member corroded by the boric acid solution or field actual measurement data; and the service life of the boric acid solution corroded concrete member is predicted.

The invention particularly relates to a discharge machining simulation control system for a nuclear power plant in a high-radiation boron-containing water environment, which belongs to the technical field of special maintenance in the high-radiation environment of the nuclear power plant and comprises a man-machine interaction system, a discharge machining control system and subsystems. Each subsystem comprises a pulse power supply system, a servo power system, a washing system, a purification system, a grounding system and a tool system; the man-machine interaction system is used for man-machine interaction and displaying the discharge machining implementation data fed back by the subsystems; the electric discharge machining control system is used for generating an electric discharge machining instruction according to the man-machine interaction information and controlling the subsystems to cooperate to perform electric discharge machining work; and each subsystem is used for implementing electric discharge machining work according to the electric discharge machining instruction and feeding back electric discharge machining implementation data. According to the invention, simulation control of discharge processing in the high-radiation boron-containing water environment of the nuclear power plant is realized.

The invention discloses a nuclear power plant fault diagnosis method and device, and the method comprises the steps: collecting nuclear power plant data, training and constructing an edge end model group and a cloud model group according to the nuclear power plant data, and the nuclear power plant data comprises operation regulations, real-time operation data and historical operation data; deploying the edge end model group to an edge computing node near equipment in a plant, and deploying the cloud end model group to a cloud platform; inputting the real-time operation data of the nuclear power plant into the edge end model group, detecting a fault type, reconstructing sensor parameters if the fault type is sensor abnormality, and uploading an abnormality report to the cloud platform if the fault type is system abnormality, and reasoning and tracing the abnormal report through the cloud model group, and pushing fault information and a decision scheme.

The invention provides monitoring equipment and system for monitoring a reactor core of a nuclear power plant. The monitoring device comprises a data preprocessing module, a model updating module, a monitoring module, a shared data pool and a monitoring interface module. On one hand, on-line monitoring of a reactor core is realized by tracking actual operation data of a reactor, and the accuracy of a monitoring result is improved; and on the other hand, the reactor core three-dimensional power distribution reconstruction is carried out in combination with the actual measurement current, so that the monitoring result (such as the second reactor core physical parameter) obtained based on the reconstructed reactor core three-dimensional power distribution is more matched with the actual operation condition of the reactor core, and the accuracy is higher.

The invention provides a containment pressure suppression test system. The containment pressure suppression test system comprises a containment simulation body, a steam conveying assembly, a pressure suppression pool and a detection assembly. The containment simulation body is provided with a first cavity; the steam conveying assembly communicates with the first cavity and can convey steam into the first cavity. The suppression pool is communicated with the first cavity and can store an ice-water mixture; the detection assembly is arranged in the suppression pool; when first pressure in the first cavity reaches a first threshold value, steam in the first cavity can enter the suppression pool and make contact with ice-water mixtures with different ice content rates, and the detection assembly can obtain temperature parameters and pressure fluctuation parameters in the suppression pool; and according to the temperature parameter and the pressure fluctuation parameter, the operation state of the suppression pool under different ice contents is determined. The influence of the performance, reliability and stability of the suppression pool under different water contents can be evaluated and analyzed, and a guarantee is provided for safe operation of a nuclear power plant, a floating nuclear power plant and a nuclear business ship in a low-temperature environment.

The invention discloses a management method and system for guaranteeing the purchase quality of spare parts of a nuclear power plant, and relates to the technical field of equipment reliability and maintenance, and the method comprises the steps: collecting and coding the information of nuclear power spare parts; before the spare parts are purchased, the qualification of the supplier is verified; a demand list is automatically generated in the purchasing stage according to the coding information of the spare parts; matching suppliers meeting conditions in a supplier database according to the content of the demand order, and generating a purchase order; the unique code of the safe and important spare part is collected, and an RFID label or a two-dimensional code label is pasted on the spare part; spare part codes and supplier information are checked by scanning RFID or two-dimensional code labels, it is ensured that the spare part information is consistent with a purchase order, and then the spare part information is put in storage. According to the invention, spare part purchasing quality process control is realized, technical clarification is greatly reduced, one-code multi-object problems are reduced, purchasing efficiency is improved, and counterfeiting is prevented, so that quality of safe and important spare parts is ensured, nuclear power plant equipment reliability is improved, and safe and stable operation of a unit is ensured.

The invention belongs to the technical field of reactor pool water turbidity control before nuclear power plant refueling overhaul loading / unloading, and particularly relates to a reactor pool turbidity control method before nuclear power plant overhaul loading / unloading. Front-end control is carried out on work related to reactor pool water quality from a unit refueling overhaul preparation stage, a descending stage, a low-low water level stage and an ascending stage; in the preparation stage, a water filling source refueling water tank before reactor pool unloading is purified before unit refueling overhaul, it is expected that purification is conducted till the turbidity of the refueling water tank is smaller than 0.1 NTU, and it is guaranteed that the water filling source of the reactor pool is clear before unloading; in the descending stage, the unit overhaul first loop 80 DEG C platform purification is completed, and discharging is completed; in the low-water-level stage, a primary loop coolant is discharged to a refueling water tank after unloading is completed, and a primary loop system is filled with water after low-water-level maintenance is completed; and the ascending stage is from primary loop water filling to charging completion. Therefore, the purpose of controlling the influence of turbid water in the reactor pool before loading / unloading on normal loading / unloading work is achieved.

A long-period fuel management method based on accident fault-tolerant fuel comprises the following steps that a replacement fuel assembly with a first enrichment degree, a second enrichment degree and a third enrichment degree is provided, and 5% lt is provided; a first enrichment degree lt; a second enrichment degree lt; according to the third enrichment degree, accident fault-tolerant fuel is adopted for the fuel replacement assembly; taking out n groups of fuel assemblies in the reactor, putting the n groups of fuel assemblies into a spent fuel pool, selecting a group of fuel assemblies with the highest backup reactivity from the spent fuel pool, arranging the selected group of fuel assemblies in the center of the reactor, and taking (n-1) groups of replacement fuel assemblies and the remaining fuel assemblies of the reactor as reloading groups; calculating an installation sequence of the refueling group according to a reactor design limit value and a safety certification condition, and carrying out installation; and the materials are repeatedly changed every 24 months. According to the method, the fuel circulation period can be prolonged to 24 months, the fuel utilization rate is effectively increased, and the economical efficiency of the nuclear power plant is improved.

A long-term fuel management method based on accident-tolerant fuel, comprising the following steps: providing replacement fuel assemblies with a first enrichment degree, a second enrichment degree, and a third enrichment degree, wherein 5% < the first enrichment degree < the second enrichment degree < the third enrichment degree, and the replacement fuel assemblies adopt accident-tolerant fuel; taking out 69 groups of fuel assemblies in the reactor and putting them into the spent fuel pool, selecting a group of fuel assemblies with the highest reserve reactivity from the spent fuel pool and setting it at the center of the reactor, and taking 68 groups of replacement fuel assemblies and the remaining fuel assemblies in the reactor as a refueling group; calculating the installation sequence of the refueling group according to the reactor design limit and safety certification conditions and installing it; repeating refueling every 24 months. This method can extend the fuel cycle period to 24 months, effectively improve the fuel utilization rate, and improve the economy of nuclear power plants.

The invention discloses nuclear power plant wall surface cleaning intelligent equipment, and relates to the technical field of cleaning equipment.The nuclear power plant wall surface cleaning intelligent equipment comprises a negative pressure adsorption structure and a cleaning structure, the cleaning structure is arranged on the negative pressure adsorption structure, the negative pressure adsorption structure is provided with walking wheels, and the negative pressure adsorption structure can achieve negative pressure adsorption on the surface to be cleaned; the cleaning structure is used for cleaning a surface to be cleaned. The intelligent equipment for cleaning the wall surface of the nuclear power plant can realize intelligent path planning and autonomous obstacle avoidance. The intelligent equipment for cleaning the wall surface of the nuclear power plant can realize intelligent cleaning of the wall surface.

The invention discloses a valve with an ultra-erosion-resistant function for a nuclear power plant, and relates to the technical field of valves. Through cooperation of the driving motor, the threaded rod, the threaded cylinder and the valve plate, lifting of the valve plate is accurately controlled conveniently, and the function of accurately adjusting the flow of the valve is achieved; the sealing performance of the valve element and the valve port is monitored through the analysis module, sealing hidden dangers can be found in time, medium leakage caused by the sealing problem is avoided, the sealing reliability of the valve is effectively improved, and the defect that the sealing problem is prone to occurring due to rigid abutting connection of the valve element and the valve port of a traditional valve is overcome. By comprehensively considering the pressure of the valve core and the valve port, the number of impurity particles, the particle size, the hardness, the medium flow rate, the impact angle and other factors, the calculation of the abrasion loss is more accurate, and the abrasion condition of the nuclear power plant valve in actual operation can be more accurately reflected.

The invention belongs to the technical field of three-dimensional model modeling, and particularly relates to a method for constructing a foreign body model of a nuclear power plant foreign body digital museum. Comprising the following steps: step 1, collecting close-range image data of a target foreign body, generating point cloud data by using feature matching and a neural radiation field, calculating a normal vector of each point, and segmenting the point cloud data through a region growing algorithm to obtain a point cloud model of the target foreign body; 2, generating a surface grid model of the foreign matter by adopting a Poisson reconstruction algorithm, and finishing and optimizing the grid model by analyzing the solid structure of the foreign matter to obtain a refined foreign matter fine model; and 3, generating high-fidelity texture mapping by adopting a method based on texture reconstruction and rendering in combination with foreign matter surface features and close-range image data, and rendering illumination and reflection effects for the model to realize visual optimization. The method has the beneficial effects that the high-fidelity three-dimensional model of the foreign matter can be efficiently generated, and the damaged area can be effectively repaired.

The invention discloses a three-dimensional modeling and intelligent operation and maintenance system and method for a nuclear power plant, and the system comprises a design drawing marking module which is used for carrying out the digital processing of a design drawing of the nuclear power plant; the data acquisition module is used for acquiring multi-dimensional data from a plurality of business systems of the nuclear power plant; the data processing module is used for carrying out data cleaning, conversion and standardization processing on the annotation file and the collected multi-dimensional data; the engineering integrated database is used for storing structured geometric information and multi-dimensional engineering data; the three-dimensional modeling module is used for generating a three-dimensional digital model based on the structured data; the real-time data updating module is used for dynamically updating the equipment state, the fault information and the position relation in the three-dimensional digital model; the three-dimensional visual display module is used for displaying the equipment layout, the pipeline trend and the building structure of the nuclear power plant; and the intelligent operation and maintenance module is used for equipment fault prediction. According to the invention, intelligentization and high efficiency of full-life-cycle management of the nuclear power plant are promoted.

The invention discloses a nuclear power plant reactor core optimization power control method based on imitation learning. A reactor core optimization power control model comprising a plurality of intelligent agents and an average layer is constructed based on the imitation learning method; and averaging the output results of the trained agents as a final control result. According to the invention, nuclear power plant reactor core optimization power control is developed based on an imitation learning method, an intelligent agent is trained in a supervised manner to learn a nuclear power plant power control system and decision logic of an operator, and serious consequences caused by direct interaction between an immature intelligent agent and a power plant are avoided while real data of the power plant are effectively utilized. A plurality of agents are integrated by adopting an integrated learning method, so that the model can be suitable for multiple working conditions of peak regulation, normal start-stop and emergency shut-down of a plurality of nuclear power plants. Each agent is optimized based on the Bayesian method in the training process, the optimal hyper-parameter is searched, and the network training efficiency and the fault diagnosis precision are improved.

The invention provides an off-site full-range radiation risk evaluation method and system under a severe accident of a nuclear power plant, and the method comprises the steps: S1, building a radioactive accident risk evaluation basic model architecture, adapting to a model demand, and building a corresponding basic database; s2, constructing radioactive smoke plume characteristics after the accident; s3, calculating accident site space gridding and smoke plume transmission discretization; s4, sampling meteorological conditions, and simulating smoke plume transmission; s5, calculating the radiation dose in the early stage of the grid domain, and carrying out risk assessment; s6, performing risk assessment on the health effect in the early stage of the grid domain; s7, calculating the medium-and-long-term stage radiation dose in the grid domain, and performing dose risk assessment; and S8, performing health effect risk assessment and accident consequence assessment in the medium-and-long-term stage. According to the method, the whole transportation process of the radioactive smoke plume under a serious accident is simulated, loss effects such as dry and wet settlement and radioactive decay of the smoke plume are considered, and the activity of the radionuclide deposited on the ground and retained in the air in the off-site space range is determined.

The utility model relates to a nuclear power plant emergency air compressor system and an air compressor test system thereof, the nuclear power plant emergency air compressor system comprises a plurality of air compressors and a drying unit used for drying compressed air output by the air compressors, and the air compressor test system comprises a plurality of pressure detection units corresponding to the air compressors one by one; the pipeline switching unit is connected with the air inlet and the air outlet of the drying unit and the plurality of pressure detection units; and the pressure relief unit is connected with the plurality of air compressors. According to the utility model, the air compressors do not influence each other, so that the functions of the air compressors can be tested under the condition that all the air compressors are not switched off, and the air supply stability, safety and reliability of the nuclear power plant are improved.

The invention relates to the technical field of nuclear power plant safety, in particular to a nuclear power plant fault diagnosis method based on data balancing, which establishes a data set balancing technology for deep learning through SA-TimeGAN, can effectively solve the problem of data imbalance caused by large normal data volume and small fault data volume in a nuclear power plant, and improves the fault diagnosis accuracy of the nuclear power plant. According to the method, firstly, small sample fault data are preprocessed, then, SA-TimeGAN is used for conducting adversarial training on the multiple types of small sample fault data, generated fault samples are added into an original training data set to achieve balance of sample data, and finally, the balanced data set is input into a CNN-BiLSTM diagnosis model for training, so that fault diagnosis is achieved. According to the method, the recognition capability of the model on minority classes of fault samples is effectively enhanced, and real data distribution is approached through adversarial training, so that high-quality synthetic data is generated.

The apparatus for reverse cutting of a structure, according to the present invention, comprises a pair of reverse pulley devices, the pair of pulley devices comprising: main tubes inserted and mounted so as to penetrate two holes; a pair of first support pulleys mounted on the inner sides of the front ends of the main tubes to rotatably support a wire; a pair of second support pulleys mounted on the inner sides of the rear ends of the main tubes to rotatably support the wire; openings which are cut in portions where the two main tubes face one another and allow the wire to move; and cap covers which are coupled to the front ends of the main tubes to seal same.

The present invention relates to an adaptive assessment method and apparatus for nuclear power plant component reliability, a storage medium, and an electronic device. The method comprises the following steps: constructing a Kriging initial model; gradually updating the Kriging initial model on the basis of first-layer samples of ICE to obtain a Kriging surrogate model; gradually updating the Kriging surrogate model on the basis of last-layer samples of the ICE; after the updating of the Kriging surrogate model is completed, obtaining a current Kriging model; and on the basis of the current Kriging model, assessing the reliability of a component to be assessed. In the present invention, first-layer adaptive Kriging gradually updates and explores a failure domain on the basis of the first-layer samples of the ICE until a set stopping criterion is satisfied, and on the basis of a currently constructed Kriging model, the last-layer samples of the ICE are also updated accordingly. Thus, it is ensured that an estimated failure probability converges unbiasedly to a real failure probability, and unnecessary updating of Kriging is avoided, thereby greatly improving calculation efficiency and saving calculation costs.

The present invention relates to a method for dynamically controlling corrosion in a steam- water circuit of a nuclear power plant, the method comprising: injecting the compounds ethanolamine and optionally ammonia, the system comprising: injection units, located at a dosing point and configured for injecting the compounds ethanolamine and ammonia in the fluid circulating in the circuit; and measurement units, configured for measuring a value of the parameters pH and at least one of the parameters selected from redox potential and corrosion potential in the circuit; the system being characterized in that it further comprises a control unit, the control unit comprising: a first acquisition module, connected to the measurement unit and configured for acquiring the measured values of the parameters and configured for acquiring at least one set-point value for said parameters; a command module, connected to the injection units and configured for commanding the injection of the compounds ethanolamine and ammonia by the injection unit as a function of the acquired at least one set-point value and measured values of the parameters, characterized in that ethanolamine is used as reducing agent. The present invention further relates to a system for dynamically controlling corrosion in a steam-water circuit of a nuclear power plant.

The invention provides a method for measuring silver content in boron-containing water, which comprises the following steps: preparing a blank standard solution and a standard working solution by using a first-stage diluent and a second-stage diluent, and respectively pretreating the blank standard solution, the standard working solution and a to-be-measured sample solution by using a third-stage diluent; the first-stage diluent comprises a nitric acid solution, the second-stage diluent comprises a boron solution, and the third-stage diluent comprises a nitric acid solution; and measuring the silver content of the to-be-measured sample solution by adopting inductively coupled plasma emission spectrometry. According to the determination method provided by the invention, the silver content in the boron-containing water of the nuclear power plant can be accurately and rapidly detected through pretreatment of the test solution, the problem of boron interference in the boron-containing water is solved, the detection accuracy of the silver content is improved, and support is provided for chemical control of nuclear power loop water.

The invention relates to a nuclear power plant solenoid valve life prediction method and terminal equipment. The method comprises the following steps: based on a set sequence model and an attention mechanism, establishing a multi-layer perceptron model which can perform sequence data processing and perform training by pertinently utilizing set feature data; obtaining a training set and a verification set, training the multi-layer perceptron model by using the training set, determining hyper-parameters of the multi-layer perceptron model by using the verification set, and obtaining the trained multi-layer perceptron model; performing model distillation processing on the trained multi-layer sensor model to obtain a prediction model capable of predicting the service life of the electromagnetic valve; acquiring characteristic data of the tested electromagnetic valve; and inputting the feature data into the prediction model to obtain a prediction result output by the prediction model. The service life of the electromagnetic valve is predicted, and the prediction model has the advantages of being simple in structure and high in prediction accuracy.

The invention discloses a nuclear power plant primary loop zinc injection control method, and the method comprises the steps: carrying out sampling detection, determining that the detected primary loop water chemical parameters meet the thermal state function test zinc injection conditions, and preparing a zinc ion solution; monitoring the zinc ion concentration of the primary loop, filling the zinc ion solution and adjusting the adding rate; monitoring is carried out, and zinc injection is stopped when abnormal conditions occur; after passivation timing is finished, zinc injection is stopped before boronizing; the unit commercial operation zinc injection control method comprises the steps that sampling detection is carried out to determine that hydrochemical parameters meet unit commercial operation zinc injection conditions; monitoring the zinc ion concentration of the primary loop, and filling a zinc ion solution; and zinc injection is continued after abnormal conditions are processed, and zinc injection is stopped when the conditions are met. According to the invention, primary side stress corrosion cracking can be relieved, and the external radiation field of the reactor core and the collective dose during unit overhaul can be reduced.

The invention relates to a nuclear power plant accident management and control method, device and system. The method comprises the steps of obtaining operation data of nuclear power equipment in a preset time period, selecting target data changing along with time from the operation data, generating a target sequence of the target data by using the target data, identifying whether an accident occurs based on the target sequence and the operation data, and obtaining an identification result, generating a plurality of accident mitigation strategies by using the operation data and the identification result, performing index evaluation on the plurality of accident mitigation strategies to obtain index evaluation values, selecting the accident mitigation strategy with the highest index evaluation value from the plurality of accident mitigation strategies to obtain a target accident mitigation strategy, and inputting the target accident mitigation strategy into the nuclear power accident simulation model for simulation to obtain a simulation result, and executing the target accident mitigation strategy under the condition that the simulation result meets a preset condition. According to the scheme, the accuracy and efficiency of execution of the accident mitigation strategy can be improved.

The present invention relates to an apparatus and method for measuring acoustic excitation noise in a steam pipeline of a nuclear power plant. Multiple layers of sound absorption materials are arranged at the end in a shielding cavity close to a bottom side, and a flexible attaching mechanism is arranged at an opening of the shielding cavity; during measurement, the flexible attaching mechanism is attached to the outer wall of a steam pipeline of an nuclear power plant; the multiple layers of sound absorption materials isolate noise in an environment outside the shielding cavity; by means of a through hole, a microphone is placed in the shielding cavity for measurement, implementing non-destructive indirect measurement of noise in the steam pipeline of the nuclear power plant from the outside of the steam pipeline, and realizing the advantages of interference prevention and easy installation; the noise isolation of the shielding cavity and the multiple layers of sound absorption materials and the tight attachment of the flexible attaching mechanism on the pipe wall eliminate the impact of field environments on measurement; and the direct attachment of the shielding cavity to the pipe wall solves the problem that measurement cannot be carried out due to a narrow space size, realizing effective noise tests in high-temperature and high-pressure environments.

The invention relates to the technical field of nuclear power plant equipment safety, in particular to a bearing fault diagnosis method and device, electronic equipment and a storage medium. According to the bearing fault diagnosis method provided by the embodiment of the invention, target bearing vibration data is acquired; performing multi-scale feature extraction on the target bearing vibration data through the bearing fault diagnosis model to obtain multi-scale bearing vibration features; carrying out bearing state detection on the target bearing based on the multi-scale bearing vibration characteristics to obtain a bearing vibration state; performing feature filtering processing on the multi-scale bearing vibration features to obtain filtered bearing vibration features; and carrying out bearing fault diagnosis on the target bearing according to the multi-scale bearing vibration characteristics, the bearing vibration state and the filtered bearing vibration characteristics to obtain a bearing fault category. Therefore, the accuracy of bearing fault diagnosis can be improved.