479results about "Computer implemented nuclear control" patented technology

A control method and system for controlling a nuclear power plant includes, in the absence of detection of an imbalance between a primary power signal (S1) and a secondary power signal (S2), the implementation of a setpoint-following mode. The nuclear power plant is controlled as a function of an operational power setpoint (COP), and in the event of detection of an imbalance, automatically implementing a power-limiting mode, including the calculation of a target equilibrium power (PEC) less than or equal to the primary power (P1) and less than or equal to the secondary power (P2), and controlling the nuclear power plant (2) as a function of the target equilibrium power (PEC).

The invention discloses a self-adaptive control method and system for a condenser for a nuclear power station, relates to the field of condenser control, and aims to predict a future load change trend in a prospective manner by analyzing historical data of a unit power instruction and steam flow, and calculate feed-forward circulating water flow based on a prediction result. On the basis, a feedback control loop based on actually measured pressure is reserved. More importantly, a self-adaptive bias correction mechanism is introduced, and continuous learning and correction are carried out on output of feedback control, so that steady-state errors of the system are eliminated, and uncertainty of the model is compensated. Finally, feed-forward, feedback and self-adaptive correction components are intelligently fused to form a final control instruction which can quickly respond to load changes and can accurately maintain pressure stability, and the problems of severe pressure fluctuation and energy waste under the dynamic working condition are effectively solved.

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.

The invention relates to a control rod drop function availability automatic test method and a DCS system, and the method comprises the steps: receiving a test start instruction, entering a test mode, and in the test mode, obtaining the test sequence of each test rod group, and the measurement rod position and the counting rod position of a control rod; sending a bar moving instruction to the current test bar group; after the control rod action of the current test rod group is completed, the counting rod position is consistent with the target rod position, and the change of the measuring rod position of the control rod is detected; judging whether the change of the measurement rod position of the control rod accords with a preset threshold value; and if yes, judging that the control rod drop function availability test of the current test rod group is qualified. According to the invention, the availability test of the rod falling function of the control rod is changed from the traditional manual operation to the automatic test, so that the loss caused by human errors is effectively avoided; time consumed by the whole test is shortened, manpower is saved, and test execution efficiency is improved; and automatic checking is carried out, and a test result is given, so that human errors caused by long manual checking time are avoided.

An exemplary system and method provide nuclear reactors with optimized detector placement. The exemplary system and method include a nuclear reactor model, Markov decision process, and reward function, where reinforcement learning can be used to iteratively generate placements of detectors to candidate positions within the nuclear reactor.

The invention discloses a small modular prismatic high-temperature gas cooled reactor core optimization design method and system based on data driving. The method comprises the following steps: constructing a training data set; a data-driven agent model is trained to replace and execute high-cost simulation calculation; and exploring a design space and generating a group of Pareto optimal candidate solutions by using a multi-objective evolution algorithm and taking the proxy model as a fitness function evaluator. The key feature of the method is a closed loop of iterative verification and model updating: verifying the candidate solution through high-fidelity simulation, if the prediction precision does not meet the convergence criterion, expanding the verified new data point to a training data set, and retraining the agent model. According to the method, the calculation cost can be remarkably reduced, a complex design space can be efficiently explored, and a physically reliable reactor design scheme with better performance can be obtained.

A method for controlling a pressurized water reactor, computer program product and control system, the pressurized water reactor includes a reactor core and a primary cooling circuit. The primary cooling circuit includes a primary cooling medium, which includes: acquiring a plurality of measurable reactor process variables and obtaining a plurality of non-measurable reactor process variables. The method further includes calculating future axial offsets at the end of a predetermined prediction time interval for a plurality of different possible boration / dilution actions based on the plurality of measurable reactor process variables and the plurality of non-measurable reactor process variables, the axial offset being a normalized difference between power of an upper half of the reactor core and a lower half of the reactor core. The calculation of the future axial offset for each of the plurality of different possible boration / dilution actions is performed in parallel.

The invention relates to the technical field of nuclear power station event early warning, aims to solve the problems of how to detect whether the current operation state is normal or not and how to predict the occurrence time of an abnormal event in the operation process of a unit, and provides an event early warning method and device based on nuclear power control logic. The method comprises the following steps: acquiring real-time data, calculating the real-time data, judging whether an abnormal event occurs or not, and when judging that the abnormal event occurs, predicting the change trend of the real-time data by adopting a key working condition parameter trend prediction algorithm based on a self-adaptive linear model so as to predict the occurrence time of the abnormal event. And outputting the occurrence time of the abnormal event and giving an alarm. According to the invention, abnormal events such as shutdown, shutdown and load shedding in the operation process can be monitored, analyzed and pre-warned, whether the current operation state is normal or not can be detected, the occurrence time of the abnormal events can be predicted, and thus auxiliary support is provided for ensuring stable and safe operation of a power plant.

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 invention discloses a power coordination control method of a high-temperature gas cooled reactor nuclear steam supply system, which relates to the technical field of nuclear power, and comprises the following steps of: coupling multiple modules in the high-temperature gas cooled reactor nuclear steam supply system through a common conventional island; the key state parameters of the to-be-adjusted module are used as main control variables, and the other modules are used as auxiliary control variables, so that the mutual influence relationship among the modules can be comprehensively considered; reasoning the control quantities based on a preset fuzzy rule to determine a fuzzy control strategy, so that adverse effects among modules can be counteracted, overall parameters are kept in a design value range, and safe and stable operation of the system is guaranteed; flexible adjustment can be performed according to key state parameters of different modules by using a fuzzy control strategy, the power of the reactor is quickly and accurately controlled, the dynamic response capability of the system is improved, and the requirement that the power of the reactor is quickly adjusted along with the load requirement of the steam turbine generator unit is met.

Disclosed is a digital twin platform provision system for detecting and responding to a fire in an installation region of electrical equipment in a nuclear power plant, which includes: a digital twin build unit for building an initial digital twin by digitally modeling target objects including equipment and subsidiary materials related to fire outbreak and damage spread; an attribute information application unit for applying fire reactivity information received from an outside and serves as attribute information about fire outbreak and spread factors to the initial digital twin, thereby building a digital twin; and a simulation information provision for unit generating simulation information, which is information fire outbreak about prediction, fire spread simulation upon fire break, and fire response and evacuation scenarios upon the fire break, by using the digital twin according to a user input and outputting the simulation information to a user terminal.

The invention relates to the technical field of nuclear power plant reactor nuclear measurement, in particular to a nuclear instrumentation system source range alarm threshold value self-adaptive adjustment method and system, and aims to realize alarm threshold value automatic adjustment of high neutron fluence rate alarm in source range shutdown by adopting the self-adaptive adjustment method suitable for the nuclear instrumentation system source range alarm threshold value. The whole process does not need human intervention, only the uplink and downlink modes need to be manually adjusted, the initial value of the setting value of the source range counting rate change rate is manually set, regular manual adjustment is needed, human intervention is not needed in the whole process during the whole loading and reactor power increasing or reactor power decreasing and unloading period, the labor cost can be greatly saved, and the working efficiency is improved. The human error risk is reduced; the method can greatly improve the adjustment efficiency of the alarm threshold value during the material loading (replacement) and power increase of the nuclear power plant, reduces the labor cost and human error risk in the adjustment process of the alarm threshold value, and releases the critical path during the material loading (replacement) and overhaul of the reactor.

The invention provides a pressurized water reactor core axial power offset phenomenon prediction method and electronic equipment. The prediction method comprises the following steps: firstly, constructing a prediction model of the axial power offset phenomenon of the reactor core; wherein the prediction model is used for representing a model that the axial power offset phenomenon of the reactor core changes along with oxidation corrosion sediment CRUD and fuel burn-up on the surface of a cladding, and the prediction model comprises a neutron physical module, a thermal hydraulic module and a CRUD material module which have a coupling relationship; and then, carrying out coupling calculation on the prediction model within a preset duration, and generating prediction information of the axial power offset phenomenon of the reactor core. According to the prediction mode, the reactor core axial power offset phenomenon changing along with fuel burnup and CRUD deposition in the operation process of the pressurized water reactor is described, and therefore the prediction precision of the reactor core axial power offset phenomenon is improved.

Provided is a three-dimensional thermal-hydraulic analysis method and system for a reactor core. The method includes: analyzing a type of a reactor core, dividing an outer-layer mesh and a computational mesh, and establishing a conservative mapping relationship and a set of transport equations; decomposing a coolant viscosity-induced frictional effect, and representing a turbulent mixing-induced exchange of a physical quantity through a source term; establishing a three-dimensional set of governing equations including mass, momentum, and energy conservation equations to describe a flow and heat transfer phenomenon within coolant channels, and forming a fully assembled matrix system based on the set of transport equations; setting a boundary condition and an initial condition for a physical field of the reactor core, and setting an initial field; and iteratively solving the fully assembled matrix system, and obtaining a thermal-hydraulic parameter.

The invention provides a nuclear power plant safety system based on causal reasoning, and relates to the technical field of nuclear power plant safety monitoring and fault diagnosis. The nuclear power plant safety system comprises a multi-source sensing layer and a causal engine layer, the multi-source sensing layer is used for collecting operation parameters of a nuclear power plant, and the causal engine layer comprises a dynamic causal network construction module and an online reasoning module. The dynamic causal network construction module is used for constructing a dynamic causal network according to the operation parameters, a continuous causal edge generated based on an equipment physical equation and a discrete event edge generated based on a probabilistic safety evaluation event tree; and the online reasoning module is used for determining a key fault propagation path according to the dynamic causal network so as to locate a fault source. According to the method, the equipment physical equation is embedded into the dynamic causal network, so that the reasoning process is ensured to accord with thermal hydraulic and safety physical rules, the problem of high false alarm rate caused by correlation misjudgment in deep learning is avoided, and the safety of the nuclear power plant is improved.

Disclosed herein are methods, systems, media, and techniques for training machine learning models to generate nuclear fusion reactor configurations. Further disclosed herein are systems, methods, computer-readable media, and techniques for configuring a stellarator, including: (a) obtaining one or more parameters of a target stellarator; (b) generating a plurality of target stellarator approximations based at least in part on the one or more parameters of the target stellarator, wherein at least a subset of the plurality target stellarator approximations: (i) have a toroidal profile, and (ii) comprise a plurality of concentric toroids; and (c) presenting at a graphical user interface a representation of the plurality of target stellarator approximations.

A method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel. The method comprises: step S1, obtaining related data of vortexes, a temporal distribution characteristic, a spatial distribution characteristic and a power spectral density of turbulence in an inlet area of each fuel assembly, and a turbulence energy spectrum; step S2, obtaining a water gap change situation of each fuel assembly, and on the basis of the water gap change situation of each fuel assembly, determining whether a three-dimensional flow field model of each fuel assembly needs to be corrected, until a water gap of each fuel assembly does not change; step S3, obtaining a reactor core nuclear power fluctuation amplitude value between fuel assemblies; and step S4, determining whether the reactor core nuclear power fluctuation amplitude value falls within a target range, and if the reactor core nuclear power fluctuation amplitude value does not fall within the target range, changing a three-dimensional structure inside a reactor pressure vessel until the reactor core nuclear power fluctuation amplitude value falls within the target range. The method for constructing and controlling a turbulence form of a fluid inside a reactor pressure vessel can eliminate or ameliorate the nuclear power fluctuation of a reactor core, reduce the wear of grids of the reactor core, and ensure the safety of reactor operation.

The invention discloses a xenon oscillation automatic suppression auxiliary early warning prompting method based on a 3AO method, and belongs to the technical field of nuclear power reactor core monitoring. Firstly, various high-precision sensors are arranged at key parts of a nuclear reactor, occurrence signs of xenon oscillation are captured in time, and the system can accurately identify early signals of xenon oscillation by analyzing data of the sensors, so that a basis is provided for formulation of a subsequent control strategy. The core of the control strategy is fusion of a 3AO method and a deep Q network algorithm, the algorithm comprehensively considers the real-time state of the reactor core, and a control decision is optimized by learning historical data, so that the optimal control rod position and boron concentration adjustment scheme is calculated. And then, the system automatically prompts an operator to adjust the control rod and the boron concentration according to a control instruction generated by an intelligent algorithm so as to respond to the change requirement of the nuclear reactor in real time, and an efficient closed-loop control system is formed by feeding back an adjustment effect in real time, so that the accuracy and the reaction speed of a control strategy can be continuously improved.

Disclosed is an intelligent control system for thermal energy recovery of circulating cooling water in a nuclear power station. The system comprises: a thermal energy recovery apparatus used for transferring thermal energy of circulating cooling water to other media by means of a heat exchanger; and a cooling water pump used for conveying the circulating cooling water between the thermal energy recovery apparatus and a reactor. The intelligent control system for thermal energy recovery of circulating cooling water in a nuclear power station further comprises a controller used for adjusting working parameters of the thermal energy recovery apparatus. Thus, the water consumption and discharge amount of the nuclear power station can be effectively reduced, greenhouse gas emissions are reduced, and the reliability and service life of the nuclear power station are improved.

The present invention relates to a method for producing a nuclear design core model accounting for load following, the method comprising the steps of: receiving load-following requirements for a corresponding cycle, wherein the load-following requirements include the amount of power reduction, the number of load-following operations, and at least one of the time or duration of operation at reduced power; determining the amount of insertion of a control rod on the basis of the load-following requirements; determining the duration of insertion of the control rod on the basis of the amount of insertion of the control rod and the maximum insertion limit thereof; determining the time of insertion of the control rod by simulating core combustion by reflecting the duration of insertion of the control rod in a nuclear design code at each of a plurality of time points of the corresponding cycle; and producing a nuclear design core model accounting for load following on the basis of the determined amount of insertion of the control rod, the determined duration of insertion of the control rod, and the determined time of insertion of the control rod.

The invention relates to the field of nuclear power control, in particular to a steady-state working condition reactor, machine and heat coordination control system and method under a multi-module high-temperature gas cooled reactor cogeneration working condition. The coordinated control system comprises a heat supply island control module, a steam turbine control module, a unit power set value module, a power distribution module, an NSSS module control system module, a main steam pressure control module and a water supply pressure control module. The coordination control method comprises coordination control under the condition that the power grid assessment limit is not considered and under the condition that the power grid assessment limit is considered. Coupling of control among the reactor, the steam turbine and heat supply is avoided, on the premise that nuclear safety and stable operation of the NSSS module are guaranteed, the heat supply requirement is met preferentially, residual heat power is used for power generation, and safe, stable and reliable operation of the whole heat supply system is guaranteed.

The application relates to a reactor core power distribution measurement uncertainty analysis method, device, equipment, storage medium and product. The method comprises the following steps: constructing a simulation operation state based on an actual operation state of a reactor core, determining state parameters of the reactor core in the simulation operation state, and obtaining group constants of the reactor core; performing calculation based on a material calculation subprogram and the state parameters to obtain material compositions of each fuel rod in the reactor core; performing calculation based on a power calculation subprogram and the material compositions of each fuel rod to obtain a real power distribution of the reactor core and obtain detector reaction intensities of the reactor core; determining power parameters of the reactor core in the simulation operation state based on the state parameters and the group constants, and performing flux map simulation based on the power parameters and the detector reaction intensities to obtain a simulation power distribution of the reactor core; and determining a measurement uncertainty of the reactor core power distribution based on a difference between the real power distribution and the simulation power distribution. The method can improve the accuracy and acquisition speed of the reactor core power distribution measurement uncertainty.

Techniques for detecting, forecasting, and reducing or avoiding disruptions in fusion devices are described. Data corresponding to operation of a fusion device may be received. Based on the data and one or more physics-based models, an occurrence of an event and a criticality level of the event may be determined. Based on criticality level of the event, whether a disruption of plasm will occur may be determined. An indication identifying the event and whether the disruption of plasma in the fusion device will occur may be output to a computing device associated with a user. In some embodiments, based, at least in part, on the criticality level of the first event, a control system associated with the fusion device may be instructed to perform one or more actions.

The invention discloses a nuclear power station DCS data acquisition system, method, device and equipment and a medium, and is applied to the technical field of nuclear power, and the method comprises the steps: configuring data point configuration data, such as data point names, data types and data point acquisition frequencies, of data points corresponding to a target data acquisition task through employing a data point configuration file; the user can conveniently fill and edit the data point configuration data, the learning curve and complexity of the user are reduced, the risk of data point configuration errors is reduced, the flexibility and convenience of data point configuration are improved, and the personalized configuration requirement of a specific application scene is met. Moreover, data exchange is carried out between the nuclear power station DCS data acquisition system and the nuclear power station DCS system by adopting a publishing and subscribing mode, so that unnecessary data polling can be eliminated, the real-time performance of data updating is improved, and the waste of network bandwidth and processing resources is reduced. In addition, data publishing and subscribing are carried out in a multi-thread mode, the performance can be optimized, and the response speed can be increased.

The invention provides a multipurpose small reactor control method, device, equipment and medium, and the method comprises the following steps: in an automatic temperature control mode, if the deviation between the actually measured temperature of a reactor outlet and the set temperature meets a preset automatic temperature control condition, determining a first control instruction of each adjusting rod of an adjusting rod group; a first control instruction is sent to an adjusting rod set executing mechanism so as to control an adjusting rod to move till the deviation between the actually measured temperature and the set temperature is within the temperature adjusting dead zone range; in the power automatic control mode, if the deviation between the set power and the actually measured power of the reactor meets a preset power automatic control condition, determining a second control instruction of the circulating pump of the intermediate loop; and a second control instruction is sent to the circulating pump executing mechanism so as to control the rotating speed of the circulating pump until the deviation between the set power and the actually measured power is within the power adjusting dead zone range. According to the invention, the temperature and the power are respectively controlled, and the control performance is optimized.

A method and system for calibrating nuclear instrumentation of a reactor is disclosed. The system comprises a reactor configured to operate at a low power level and an external heater configured to heat the reactor to its operating temperature and maintain the reactor in a critical state. The reactor maintains its temperature at a steady state by repeating steps of reducing the heater's power and increasing the reactor's power by changing one or more reactivity settings, wherein the reduced heater's power is compensated by using the increased power generated by the reactor itself. The nuclear instrumentation is configured to measure the reactor's output power and compare it with the calculated reactor's power. Based on the discrepancy between the calculated power and the measured power exceeds a threshold; and the nuclear instrumentation is appropriately calibrated by adjusting one or more settings.

The invention relates to the technical field of compressed air system intelligent control, and discloses an intelligent control and air supply distribution method and system for a nuclear facility compressed air system pipe network. Carrying out numerical simulation on the distributed arrangement and operation conditions of different air compression stations and the flow and pressure characteristics of the compressed air pipe network under the intermittent load of different process equipment; on the basis of data generated by numerical simulation, the pipe network tail end pressure of the compressed air pipe network system is adjusted through an intelligent fuzzy self-tuning PID control method, and the operation state of an air compressor unit of the compressed air pipe network system is optimized on the basis of a flow matching strategy; pipe network operation parameters of the compressed air pipe network system are collected in real time, and the control strategy is fed back and adjusted. The whole compressed air system can operate efficiently and stably, the requirements of all links of nuclear facilities for compressed air are guaranteed, meanwhile, the operation intelligence and reliability of the system are improved, and adverse effects caused by unstable pressure and the like are reduced.

This invention relates to the monitoring and diagnosing of nuclear power plants for its thermal performance using the NCV Method. Its applicability comprises any nuclear reactor such as used for research producing a useful output. Its greatest applicability lies with conventional Pressurized Water Reactor and Boiling Water Reactor nuclear plants generating an electric power. Its teachings of treating fission as an inertial process, a phenomena which is self-contained following incident neutron capture, allows the determination of an absolute neutron flux. This process is best treated by Second Law principles producing a total fission exergy. This invention also applies to the design of fusion thermal systems regards the determination of its Second Law viability and absolute plasma flux.

The invention discloses a safety display and control integrated system of a nuclear power station, which relates to the field of safety level control of the nuclear power station and comprises an FPGA module, an MCU module and a CPLD module. The FPGA module is connected with the MCU module; the FPGA module and the MCU module are both connected with the CPLD module; the FPGA module is used for collecting and transmitting ring network data; the MCU module is used for display control and communication processing; the CPLD module is used for diagnosing the running state of the FPGA module and the running state of the MCU module; the MCU module comprises a display core and a communication core; the display core is used for running a GUI graphic function; the communication core is used for operating communication, control and recording functions. The safety display unit which is safe, stable and reliable in operation is provided for a nuclear power station safety level control system.