116results 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.

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 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 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.

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.

The invention discloses a nuclear power station master control room control authority switching method and system. The method comprises the following steps: acquiring a preset mode signal of a control display device of a nuclear power station; obtaining respective states of a workstation mode signal, a backup panel mode signal and a remote shutdown station mode signal; outputting one of the workstation mode signal, the backup panel mode signal and the remote shutdown station mode signal in an effective state as an operation mode signal; and judging whether the preset mode signal is consistent with the operation mode signal or not, and if not, generating a locking control quantity on a control display equipment side to shield an execution control and / or prohibit generation and issuing of a control instruction. According to the method, inconsistent operation is directly locked on the nuclear power station control display equipment, unauthorized and misoperation are inhibited, the human safety is improved, the data volume of network transmission signals transmitted to a nuclear power station control cabinet is reduced, the control right switching logic is simplified, and the stability of a main control room is improved.

The invention discloses a design processing method and device of a state machine, a storage medium and electronic equipment, relates to the technical field of nuclear power instrument control systems, and mainly aims to solve the problems of high cost and low reuse rate when the state machine in a nuclear power DCS (Distributed Control System) is designed by adopting existing graphical model design software. Performing modularization customization processing on each state machine element to obtain each corresponding component module; the state machine elements comprise a state machine, a state, migration and bifurcation; tool options corresponding to the state machine, the state, the migration and the bifurcation are set on a toolbar of a software page, and an incidence relation is established between the corresponding tool options and the component module; and when the tool option is selected, calling the corresponding component module based on the incidence relation, and designing a state machine structure on a software page according to the sequence of the engineering structure.

The application discloses a nuclear power device control system and method coupled with a heat pipe reactor and an sCO2 Brayton cycle, including a rotating shaft rotating speed control system and method, which maintains the rotating shaft rotating speed constant by adjusting the working medium flow entering the turbine, so as to ensure the power supply quality of the whole system; a compressor inlet temperature and inlet pressure control system and method, which respectively maintains the compressor inlet temperature and inlet pressure constant by adjusting the seawater flow and the pressure tank pressure, so as to ensure that the temperature and density of the working medium at the compressor inlet are in a supercritical state; and an intermediate heat exchanger outlet temperature control system and method, which adjusts the reactor power by controlling the core reactivity, and then maintains the intermediate heat exchanger outlet temperature constant. The application can provide a reference for the control system design of a high-efficiency compact nuclear power device coupled with a heat pipe reactor and a supercritical carbon dioxide Brayton cycle.

This invention discloses a method and system for the operation and maintenance of a shielded pump in a nuclear reactor experimental setup. By monitoring the stress of devices related to the shielded pump within the experimental setup and utilizing a stress assessment model for the shielded pump, the invention enables a clear, comprehensive, and real-time diagnosis of the pump's operating status. This allows for more rational intervention in preventative partial and full disassembly inspections of the shielded pump, enhancing the targeted nature of maintenance interventions, saving manpower and time costs, and increasing the safety and economy of shielded pump operation and maintenance.

The invention provides a control method and device for a starting separator of a nuclear reactor. The control method comprises the following steps: acquiring target steam pressure of the starting separator, actual steam pressure of the starting separator and predicted pressure output by a steam pressure model; the steam pressure model comprises a non-linear part, and the non-linear part is obtained through online learning according to the deviation between the actual steam pressure and the predicted pressure predicted by the steam pressure model in each control period; the steam pressure model is used for predicting the pressure of the starting separator in the future time period according to the nonlinear part; according to the target steam pressure, the actual steam pressure and the predicted pressure, a predicted control instruction for starting the separator is determined; according to the prediction control instruction and the actual steam pressure, a target control instruction for starting a steam valve of the separator is generated; and adjusting the opening degree of the steam valve according to the target control instruction. According to the embodiment of the invention, the accuracy of starting separator control can be improved.

This invention discloses a megawatt-class small liquid metal-cooled reactor natural circulation simulation test device and method. The test device simulates fuel rods with a total power output in the MW range, realistically simulating the high-power operation of a small natural circulation liquid metal reactor. A vacuum insulation layer covering the heating section of the simulated fuel rods is installed between the rising channel of the hot pool and the descending channel of the cold pool to block lateral heat conduction and eliminate the "thermal short circuit" problem. A three-loop coupled cooling design of liquid metal-heat transfer oil-water is adopted, ensuring high safety. Furthermore, an air lift auxiliary device and a heat tracing device are integrated. The former effectively breaks the fluid stagnation during the initial startup phase through gas-liquid two-phase disturbance to assist in establishing natural circulation, while the latter ensures that the working fluid remains in a molten state during non-operational or low-power conditions, preventing equipment damage due to working fluid solidification. This invention can accurately simulate the real natural circulation characteristics of a small liquid metal-cooled reactor, combining high precision, safety, and operational reliability.

The present invention relates to a system for determining the operation state of a power plant on the basis of an operation mode. The system may comprise: an operation mode determination device for determining the current operation mode of the power plant by using state information regarding each device in the power plant; a power plant operation state assessment device for assessing the current operation state of the power plant on the basis of the current operation mode; and an interface for providing the current operation state of the power plant to an operator.

The present invention belongs to the field of reactor physics technology for heavy water reactor nuclear power plants, and specifically relates to a method for preemptively applying the ROP set value for a heavy water reactor transition core with improved fuel. The method comprises the following steps: Step 1: Calculating the X% increase in the ROP set value for the full core with both the original design fuel and the improved fuel; Step 2: Calculating the D EFPDs required to fully fill the high power region of the core with the improved fuel and the corresponding EFPD at the time point; Step 3: Calculating the conservative increase in the ROP set value by Y% compared to the original design fuel core after the fuel channels in the high power region are fully filled with the improved fuel; Step 4: ROP set value evaluation and calculation of the minimum increase by Y% at full power. The present invention has the beneficial effect of increasing the ROP margin earlier, thereby generating more power generation revenue for the nuclear power plant, and reducing the power reduction during refueling earlier, thereby enhancing the operational safety of the nuclear power plant.

To reduce monitoring cost in a nuclear system.SOLUTION: A nuclear system comprises: a nuclear reactor core; a heat conductor that conducts heat which is generated by nuclear reaction of nuclear fuel in the nuclear reactor core to outside by means of a cooling medium; a reactivity control device that has a plurality of control drums in which a neutron absorption part arranged outside the nuclear reactor core to absorb neutrons is provided; a state detector that detects an operational state of a nuclear reactor; and an analysis unit that reproduces the operation state of the nuclear reactor in a virtual space using a virtual model of the nuclear reactor on the basis of a detection result of the state detector.SELECTED DRAWING: Figure 3