159results about "Nuclear power plant control" 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 discloses a reactor instrument control system equipment fault early warning method and related equipment, and relates to the technical field of reactor operation and maintenance, and the technical scheme is characterized in that the method comprises the following steps: obtaining operation parameters of target equipment of a reactor instrument control system in real time, and synchronously collecting historical operation data and historical fault data of the target equipment at the same time; constructing a machine learning analysis model, and training the machine learning analysis model by taking the historical operation data and the historical fault data as training samples; running parameters collected in real time are input into the trained machine learning analysis model, abnormal features representing the change trend and the deviation degree of the real-time parameters are output, a preset equipment abnormal feature library is matched based on the abnormal features, whether the target equipment has faults or not is judged, and a health degree index representing the health state of the target equipment is calculated; and if the fault judgment result is that the potential fault exists or the health degree index is lower than the preset health threshold value, generating a corresponding fault early warning signal and a health degree evaluation result.

The embodiment of the invention provides a pressure control method, electronic equipment, a pressure stabilization control system and a storage medium, and relates to the technical field of nuclear reactors. The method comprises the following steps: acquiring a pressure measurement value and a pressure setting value of the pressure stabilizer; determining a pressure deviation value according to the pressure measurement value and the pressure setting value; if the absolute value of the pressure deviation value is smaller than or equal to a first preset value, the charging and discharging flow of a loop is adjusted according to the pressure deviation value and the pressure measurement value so as to control the pressure of the pressure stabilizer; if the absolute value of the pressure deviation value is larger than the first preset value, the pressure stabilizer is controlled to be inflated or deflated according to the pressure deviation value so as to control the pressure of the pressure stabilizer. According to the embodiment of the invention, the voltage stabilization effect of the voltage stabilization control system can be improved.

The invention discloses a nuclear reactor thermal hydraulic acceleration calculation method based on a modular parallel idea. The method comprises the following specific steps: 1, modeling each subsystem of a nuclear reactor system based on a sorted nuclear reactor subsystem list; 2, establishing a coupling boundary on the basis of modeling the subsystems respectively; 3, carrying out initial state parallel calculation according to a set initial boundary condition; 4, establishing a thermal hydraulic boundary information transmission strategy for the subsystem model; 5, advancing the time step, and using a calculation result in the step 4 as an initial thermal hydraulic parameter for calculation of the next time step; and 6, repeating the steps 4 and 5 until all-time parallel computing is completed. According to the method disclosed by the invention, aiming at the thermal hydraulic calculation process of the nuclear reactor, the thermal hydraulic calculation speed of the nuclear reactor system is improved while the thermal hydraulic parameters are accurately calculated.

The invention relates to the technical field of high-temperature gas cooled reactor power automatic control, and discloses a segmented full-automatic control method and system for a high-temperature gas cooled reactor unit, and the method comprises the steps: monitoring the characteristics of a reactor core and the steam state of a steam generator when a high-temperature gas cooled reactor is started; performing power stage division on the full power range of the unit according to the characteristics of the reactor core and the steam state of the steam generator; according to the method, from starting of the high-temperature gas cooled reactor to full-power operation, the different power stages are analyzed according to the characteristics of the reactor core and the steam state of the steam generator, different control measurements are adopted according to the characteristics of the different power stages, and the power of the reactor is automatically controlled until the power of the reactor is increased to a full-power set value. And full-power stage automatic control is realized, the operation pressure of an operator is reduced, and the risk of misoperation is reduced.

The invention provides a nuclear power plant circulating water adjusting method, system and device and electronic equipment, and relates to the technical field of nuclear power plant control, and the method comprises the steps: obtaining the saturated steam temperature of a condenser of a target nuclear power plant, and periodically collecting the seawater inlet temperature of a water inlet of the target nuclear power plant and the unit output; determining the back pressure of the condenser based on the saturated steam temperature; determining the target circulating water flow of the target timestamp based on the back pressure, the seawater inlet temperature and the unit output; and generating a frequency conversion instruction based on the target circulating water flow, and sending the frequency conversion instruction to the water pump motor. Therefore, by periodically collecting the seawater inlet temperature and the unit output of the water inlet of the target nuclear power plant, determining the backpressure and further determining the target circulating water flow, the backpressure of the condenser can be kept at the optimal level, and therefore the method can adapt to various nuclear energy comprehensive utilization scenes such as power grid peak regulation and nuclear energy heat supply; the power generation efficiency of the nuclear power plant is improved and the power generation cost is reduced.

The present application relates to a data correction method and apparatus for a nuclear reactor, a computer device and a storage medium. The method comprises: acquiring calibration reactor core data of a reactor of a nuclear power plant at a calibration moment, the calibration moment referring to the moment when a detector in the reactor is calibrated; acquiring real-time reactor core data obtained by using the detector to monitor the reactor; performing thermal hydraulic calculation on the reactor on the basis of the real-time reactor core data to obtain an estimated departure from nucleate boiling ratio of the reactor; and on the basis of a reactor core data deviation between the real-time reactor core data and the calibration reactor core data, correcting the estimated departure from nucleate boiling ratio to obtain a corrected departure from nucleate boiling ratio, wherein the corrected departure from nucleate boiling ratio is used for protection and control of the reactor. The method can improve the accuracy of reactor protection and control.

A radiation monitoring system for a nuclear power plant, comprising: a radiation monitoring device (110); a central control platform (120) which is communicatively connected to the radiation monitoring device (110), wherein a data acquisition module (121), an analysis and determination module (122), and a warning and alert module (123) are provided in the central control platform (120); and a smart display terminal (130), which is communicatively connected to the central control platform (120). The data acquisition module (121) acquires monitoring information and parses same to generate parsed information; the analysis and determination module (122) performs statistical analysis processing on the parsed information to generate statistical analysis data, and generates an alert instruction when the statistical analysis data falls within a preset range; the warning and alert module (123) performs alert processing on the basis of the alert instruction and generates warning information; the smart display terminal (130) displays the warning information, and generates scheduled analysis reports on the basis of the statistical analysis data. Real-time monitoring of radiation of nuclear power plants is achieved.

The invention discloses a nuclear power plant distributed control system and control method, and relates to the technical field of nuclear power plant digital control, and the method comprises the steps: a global optimization layer generates a cross-unit general control strategy library based on a transfer learning algorithm according to the decision period of the global optimization layer; when the coordination control layer operates by using the transferable control strategy corresponding to the unit equipment where each two-loop subsystem is located, the corresponding two-loop subsystem control weight is dynamically allocated based on the multi-agent attention network according to the decision period of the coordination control layer; when the execution control layer operates by using the dynamically distributed two-loop subsystem control weight, the first-loop temperature, the first-loop pressure and the coolant flow are obtained according to the decision period of the control layer, a safety constraint matrix is calculated through a safety manifold projection module, and a quadratic programming problem with a relaxation factor is solved; and the corrected control weight of the two-loop subsystem is obtained and converted into a control instruction to be issued to a DCS field control station.

The invention relates to a nuclear power unit power control method and system based on seawater temperature and tide level prediction, and the method comprises the steps: obtaining seawater temperature data of a sea area where a nuclear power unit is located in past first set time, and predicting the seawater temperature after second set time according to the seawater temperature data, so as to obtain a predicted temperature; obtaining tide data of a sea area where the nuclear power unit is located within a third set time, and predicting a tide level after the second set time according to the tide data to obtain a predicted tide level; and obtaining historical operation data of the nuclear power unit, and determining an electric power set value of the nuclear power unit according to the historical operation data, the predicted temperature and the predicted tide level. The power generation power and stability of the nuclear power unit can be improved, and the over-power risk of the unit is reduced.

The invention discloses a nuclear power plant production command center management system, which relates to the technical field of nuclear power and comprises a data storage device, an equipment management platform, a decision management platform and an operation management platform. Wherein the equipment management platform comprises an equipment remote monitoring module and an intelligent early warning module; the decision management platform comprises an intelligent decision center module; the operation management platform comprises an intelligent dispatching module; the equipment remote monitoring module realizes real-time acquisition of equipment operation parameters through an intelligent sensor network, and the intelligent early warning module performs anomaly detection based on a dynamic threshold model, so that the problem of lagging of traditional fixed threshold early warning is solved; the intelligent decision-making center module responds to the early warning information, carries out deep knowledge reasoning by combining historical operation data and a nuclear power plant group reactor knowledge graph, generates a scientific decision-making scheme, and solves the limitation of artificial experience decision-making; and the intelligent dispatching module automatically generates the operation task based on the decision scheme and matches the operation team, so that the operation efficiency is improved.

The invention particularly relates to a nuclear power plant accident handling path feasibility judgment method, computer equipment and a storage medium, and belongs to the technical field of nuclear power plant operation. The method comprises the steps of nuclear power plant state identification; nuclear power plant accident diagnosis and state tracking; evaluating the availability of nuclear power plant accident handling key equipment; evaluating the state of a nuclear power plant accident handling key system; nuclear power plant accident handling path feasibility judgment; and human-computer interface display. The computer equipment and the storage medium are used for implementing the steps of the method. According to the invention, the feasibility of the accident handling path is judged based on nuclear power plant equipment availability and system state evaluation, and when the nuclear power plant has an accident, an operator is helped to find the feasibility of the accident handling path in advance, so that a corresponding coping strategy is made in time, accident consequences are degraded to the maximum extent, and the safety of the nuclear power plant is improved.

The present invention discloses a system and method for starting and stopping a multi-module reactor direct current evaporator nuclear power unit. The system includes a multi-module reactor, a first steam-water separator, a second steam-water separator, a deaerator, a high-pressure heater, a unit drainage tank, a steam turbine and a condenser; the method includes a unit start-up process and a unit stop process. The multi-module reactor nuclear power unit provided by the present invention also starts and stops a reactor one by one during the start-up and stop process. Multiple reactors share two start-stop reactor systems. During the start-up and stop process of the first reactor, the second start-stop reactor system can adjust its parameters to adapt to the parameters of the next reactor to be started and stopped, so the two start-stop reactor systems meet the needs of starting and stopping multiple reactors.

The utility model is suitable for the technical field of nuclear power equipment maintenance, and provides a nuclear power steam generator maintenance device which comprises a conveying belt body, a magnetic adsorption head and an adsorption driving assembly. The magnetic adsorption head comprises a supporting frame and a magnetic assembly. The whole magnetism of the magnetic adsorption head can be controlled by changing the position, away from the working opening, of the magnetic assembly, the magnetic adsorption head can accurately capture and remove filament-shaped foreign matter between pipes, and therefore the narrow space between the evaporator pipes can be cleaned and maintained more conveniently. Meanwhile, the defects of high grabbing difficulty, low grabbing efficiency and the like caused by the fact that existing grabbing tools at home and abroad are limited by the unique and narrow space structure of the evaporator are overcome, the device can work in the narrow space between evaporator pipes, multiple foreign matters can be adsorbed between the pipes through one-time operation, and the working efficiency is improved. And the difficulty of grabbing fragmentary filiform foreign matters between pipes is reduced, the working efficiency of grabbing fine metal foreign matters is improved, and the exposure dose of field operators is also reduced.

The invention relates to a control method and device for a high-temperature gas cooled reactor, and the method comprises the steps: responding to a received power automatic control instruction inputted by a user, and obtaining a power automatic control parameter; and performing rod position control on each control rod in a control rod group of the high-temperature gas cooled reactor based on the automatic power control parameters, and acquiring and displaying at least one parameter of the high-temperature gas cooled reactor in the rod position control process.

The invention relates to a hybrid power generation device. The hybrid power generation device comprises: a nuclear power generator assembly (2) configured to supply primary power (WEP) from a nuclear energy source; an electrolyser (3) operable to produce a mixture (M) containing hydrogen from an inlet water stream (FWI) of gas and / or liquid phase; a hydrogen storage system (5) coupled to the electrolyser (3) to receive hydrogen from the mixture (M); and a hydrogen generator assembly (7) operable to generate secondary electric power (WES) using hydrogen from the hydrogen storage system (5). A power divider (8) coupled to the power distribution network (15) and to the electrolyser (3), configured to controllably distribute the main power (WEP) between the power distribution network (15) and the electrolyser (3).

The invention relates to the field of nuclear power executing mechanisms, in particular to a nuclear power plant intelligent electric executing mechanism and a transformation method thereof.The executing mechanism comprises an executing unit and a control unit, the executing unit comprises a mechanical and transmission mechanism, and the mechanical and transmission mechanism is connected with a torque sensor, a position sensor and a motor; the motor is sequentially connected with the temperature sensor, the temperature sensor is connected with the terminal strip, the torque sensor is sequentially connected with the voltage-to-current module and the terminal strip, and the position sensor is connected with the terminal strip; the control unit comprises a control module, an aviation plug and a current-to-voltage module, the control module is respectively connected with the aviation plug and the current-to-voltage module, the current-to-voltage module is connected with the aviation plug, and the aviation plug is connected with a terminal strip of the execution unit through a cable; the control module transmits a control signal to the execution module through the aviation plug. Through the arrangement of the circuit enhancement module, the communication distance is greatly prolonged, and the signal reliability is also greatly enhanced.

A pressure control method and a pressure stabilizing control system. The method comprises: acquiring a pressure measurement value and a pressure setting value of a pressurizer (101); determining a pressure deviation value on the basis of the pressure measurement value and the pressure setting value (102); if the absolute value of the pressure deviation value is less than or equal to a first preset value, regulating the charging flow and the letdown flow of a circuit on the basis of the pressure deviation value and the pressure measurement value, so as to control the pressure of the pressurizer (103); and if the absolute value of the pressure deviation value is greater than the first preset value, on the basis of the pressure deviation value, controlling the pressurizer to inflate or deflate so as to control the pressure of the pressurizer (104). The method and the system can improve the pressure stabilizing effects of pressure stabilizing control systems.

The invention relates to a nuclear reactor critical heat flux density data evaluation method and device. The method comprises the following steps: determining data points of critical heat flux density data and a type of variables required for evaluating the critical heat flux density data; determining a preset continuous range of one type of variables; generating a first-class variable statistical table according to the data points, the first-class variables and a preset continuous range of the first-class variables; and evaluating the integrity of the critical heat flux density data according to the one-class variable statistical table to obtain an integrity evaluation result. By adopting the method, the reliability of an integrity evaluation result can be improved.

The application relates to a method and device for determining a thermal margin of a nuclear reactor, a computer device and a medium, and relates to the field of nuclear engineering.The method comprises the following steps: determining a first predicted deviation of bubble nucleate boiling ratio of a to-be-detected reactor, wherein the first predicted deviation of bubble nucleate boiling ratio refers to a minimum predicted deviation of bubble nucleate boiling ratio of the to-be-detected reactor; determining a reactor parameter of the to-be-detected reactor at a predicted time corresponding to the first predicted deviation of bubble nucleate boiling ratio, wherein the reactor parameter comprises an axial power distribution of at least one reactor component in the to-be-detected reactor and a total power of the component, and the total power of the component refers to a sum of the axial power and the radial power of the reactor component; and determining the thermal margin of the to-be-detected reactor according to the reactor parameter.The application can improve the accuracy of the thermal margin of the nuclear reactor, and can more accurately reflect the actual situation of the to-be-detected reactor.

The invention belongs to the technical field of nuclear power plant operation control, and particularly relates to a nuclear power plant seawater plate heat exchanger pressure difference and temperature control system and method. An equipment cooling water outlet of the plate heat exchanger is provided with an equipment cooling water thermometer, and a plate heat exchanger bypass pneumatic control valve and a plate heat exchanger differential pressure gauge are arranged between a seawater inlet and a seawater outlet of the plate heat exchanger in parallel. A seawater inlet of the plate heat exchanger is respectively connected with the plate heat exchanger positive flushing inlet pneumatic isolating valve and the plate heat exchanger back flushing outlet pneumatic isolating valve, and a seawater outlet of the plate heat exchanger is respectively connected with the plate heat exchanger positive flushing outlet pneumatic isolating valve and the plate heat exchanger back flushing inlet pneumatic isolating valve. Automatic backwashing of the plate heat exchanger can be achieved, impurities on the seawater side in the plate heat exchanger can be washed away, and the temperature of cooling water of equipment can be controlled to be not lower than the temperature limit value required by design in winter.

The present invention provides a power distribution management method and system for an RPE pump pit, which classifies the single pumps in a single-pump pit according to their usage requirements: when the single pump is a single-pump on-demand pump, it needs to be powered on at any time during overhaul; when the single pump is a single-pump delayed pump, it needs to delay power outages during overhauls, and the power supply to the connected distribution board will not be cut off until the water in the pump pit is drained; when the single pump is a single-pump temporary pump, it may need temporary power supply during overhauls; the above-mentioned single pumps are separately managed for power distribution, so that the single pumps under different classifications can all be powered when needed, so as to be used as needed during overhauls. The present invention also manages power distribution for the dual pumps in the dual-pump pit during overhauls, so that at least one of the dual pumps can be powered, and the dual pumps serve as backup for each other. Therefore, the present invention solves the problem of power distribution management of RPE pump pits during overhauls, realizes effective management of the classification of single and dual pumps, and ensures the maintenance safety of equipment in single-pump pits and dual-pump pits during overhauls.

The present invention relates to a nuclear simulation design analysis method and system based on Monte Carlo calculations, comprising: determining multiple primary factors of the nuclear simulation design analysis scheme based on a preset objective and a basic model of the nuclear simulation design analysis scheme, and constructing multiple auxiliary models of the nuclear simulation design analysis scheme based on the multiple primary factors; performing Monte Carlo calculations on the basic model and each auxiliary model separately, obtaining and fitting all first calculation results to obtain multiple groups of second calculation results; wherein each group of second calculation results corresponds to a set of values ​​of the primary factors of the nuclear simulation design scheme. The present invention obtains a large number of different nuclear simulation design analysis scheme results through multiple Monte Carlo fitting of the basic model and multiple auxiliary models, thereby reducing the computational complexity of the nuclear simulation design analysis while ensuring the stability of the calculation results.

The invention relates to the field of thermal hydraulic characteristic analysis of fuel assemblies, in particular to a method and device for determining a heat transfer correlation formula of a wire winding positioning fuel assembly, and the method comprises the steps: coupling a wire winding assembly model with a light rod bundle assembly model, and constructing a wire winding positioning fuel assembly model; based on the wire winding positioning fuel assembly model, a first average convective heat transfer coefficient under the preset physical working condition is obtained; according to the first average convective heat transfer coefficient and a second average convective heat transfer coefficient of the light rod bundle assembly model under the preset physical working condition, a correction factor is obtained through calculation; and according to the correction factor and a first heat transfer correlation formula corresponding to the light rod bundle assembly model, calculating to obtain a second heat transfer correlation formula corresponding to the wire winding positioning fuel assembly model. According to the method, the thermal hydraulic characteristics of the wire winding positioning fuel assembly are accurately analyzed, and the heat transfer correlation formula, namely the second heat transfer correlation formula, of the wire winding positioning fuel assembly is determined.

This invention discloses a power regulation and coordinated control method for a heat pipe reactor-supercritical carbon dioxide nuclear power plant, comprising: determining that if the percentage deviation between the actual output power and the expected output power of the turbine compressor unit is less than a preset proportional value or proportional range and greater than zero, then using the working fluid flow control process of the turbine compressor unit as the primary regulation mode and the power control process of the heat pipe reactor as the secondary regulation mode; determining that the percentage deviation between the actual output power and the expected output power of the turbine compressor unit is greater than a preset proportional value or proportional range, then using the power control process of the heat pipe reactor as the primary regulation mode and the working fluid flow control process of the turbine compressor unit as the secondary regulation mode. This multi-objective control mode, which integrates primary and secondary controllers in a coordinated manner, helps ensure that the turbine compressor unit has a fast power response to external forces, while improving the energy conversion efficiency of the heat pipe reactor-supercritical carbon dioxide nuclear power plant, thus balancing the energy conversion efficiency and rapid mobility of the nuclear power plant.

This application belongs to the field of digital nuclear power technology and discloses a method and apparatus for tracking the operation of a digital twin system. It can accurately reproduce the operation process and recreate the unit's state at corresponding moments. The method generates an operation sequence based on generated samples and simulates the operation of all equipment in the initialized digital twin system according to the operation sequence. The apparatus includes an inventory acquisition module, an initialization module, a log parsing module, and a simulation module. This application can monitor and simulate the operating status of a nuclear power plant in real time, providing important support for the safe operation and maintenance of nuclear power plants.

Disclosed herein is an adjustable core assembly for a nuclear reactor. The adjustable core can include a plurality of reactivity control cells configured to accommodate reactivity control rods and a plurality of unit cells. The plurality of unit cells define a radial dimension corresponding to an initial output power of the core. Each unit cell of the plurality of unit cells is configured to accommodate a fuel configured to generate energy and a heat pipe configured to transfer thermal energy away from the core. Each unit cell of the plurality of unit cells can be radially coupled to an adjacent unit cell to change its radial dimension, the changed radial dimension corresponding to an adjusted output power of the core, the adjusted output power of the core being different from the initial output power of the core.

The invention provides a multifunctional test bench for researching the performance of a primary loop voltage stabilizer of a reactor and a test method. The test bed is mainly composed of a water supplementing system, a main loop system, a steam and inert gas pressure stabilizing system and a drainage system. The test bench can be used for simulating the cold-state starting process of a primary loop of a traditional reactor, starting characteristics of the loop when a steam pressure stabilizer and an inert gas pressure stabilizer serve as the loop are obtained respectively, and a basis is provided for performance evaluation of different pressure stabilizers.