288results 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 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.

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.

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.

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 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.

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 discloses an in-core component with highly uniform core outlet flow rate and its design method, relating to the field of nuclear reactor design technology. It solves the problem of poor uniformity in coolant flow distribution at the core outlet in existing reactors, balances flow resistance at various points at the core outlet, and reduces lateral flow between different channels. The specific solution is as follows: it includes a core top plate and a support column assembly mounted above the core top plate. The core top plate is provided with several water flow holes and mounting holes. The water flow holes consist of circular and square water flow holes. The circular water flow holes are divided into first and second circular water flow holes with different functions. The area ratio of the square water flow holes to the circular water flow holes is 1.1:1 to 1.3:1.

Disclosed herein is a stellarator comprising two sets of coils, namely a set of encircling coils which encircle the plasma axis, and a set of shaping coils which do not encircle any other coil or the plasma. In some embodiments, the encircling coils include a structural element to maintain their shape under magnetic forces. In some embodiments, the shaping coils are mounted onto one or more structural elements which, together with the shaping coils, constitute a field shaping unit. Also disclosed is a controller which may modify the electrical current flowing in one or more subsets of the coils in order to achieve target plasma parameters. Also disclosed is a method of designing a set of shaping coils by discretizing a surface dipole or current potential distribution.

The present invention provides a method for testing the leak rate of a reactor containment building, comprising: a partitioning step of separating the reactor containment building into an upper region and a lower region using an isolation device; a measurement step of measuring the leak rate of the containment building after pressurizing the inside of the containment building to a maximum pressure; and a determination step of determining whether the leak rate measured in the measurement step meets the leak rate acceptance criteria for the containment building.

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 unit for generating simulation information, which is information about fire outbreak 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 present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

The invention discloses a penetration probability-based characteristic line neutron transport calculation accelerated scanning method, which comprises the following steps of: firstly, establishing characteristic line information, and dividing each characteristic line into a plurality of characteristic line segments; establishing a penetration probability coefficient from the incident boundary angle flux to the emergent boundary angle flux directly along each characteristic line; scanning the whole characteristic line in all directions, calculating an emergent boundary angle flux from the incident boundary angle flux by using the penetration probability coefficient, obtaining an incident boundary angle flux of the boundary point in the other direction from the emergent boundary angle flux, calculating the emergent boundary angle flux from the incident boundary angle flux by using the corresponding penetration probability coefficient again, and repeating the process to obtain the emergent boundary angle flux of the boundary point in the other direction. The incident boundary angular flux in all directions of the boundary point is converged; and for each feature line in all directions, only one time of feature line segment-by-segment calculation is carried out, flux information on the fine grid is obtained, and scanning is ended. According to the method, multiple times of feature line segment-by-segment calculation before convergence are avoided, the calculation amount is reduced, and the calculation efficiency of the feature line method is improved.

METHOD FOR DETERMINING A MINIMUM CRITICALITY ACCIDENT SOURCE TERM The invention relates to a method for determining a minimum criticality accident source term, comprising the following steps: obtaining a number of leaky neutrons or photons DN for a minimum criticality accident by determining a minimum number of fissions Vmin, a minimum average number of fission neutrons Nmin, and a minimum leakage rate of neutrons or photons Dmin from respective devices in an installation; obtaining, based on the number of leaky neutrons or photons DN for the minimum criticality accident and an energy spectrum of the minimum criticality accident source term, a leakage source term corresponding to the minimum criticality accident. [Fig. 1]

A nuclear reactor system for use with an electric power grid is disclosed, the nuclear reactor system including a nuclear reactor, an energy storage system coupled to the nuclear reactor, and control circuitry coupled to the nuclear reactor and the energy storage system, wherein the control circuitry is configured to monitor power demand on the electric power grid, monitor power output generated from the nuclear reactor, detect changes in power demand, cause the energy storage system to temporarily compensate for the changes in power demand, and adjust the power output based on the changes in power demand.

The invention belongs to the technical field of nuclear power, and particularly relates to an intelligent parameter monitoring method and device for a nuclear power plant. According to the intelligent parameter monitoring method for the nuclear power plant provided by the invention, parameter monitoring under two different working conditions of a steady state and a transient state is realized, different monitoring methods are introduced for different working conditions, and the precision and reliability of parameter monitoring are improved; parameter abnormity can be found in advance, operators can have sufficient intervention and processing time, and safe operation of the nuclear power plant is facilitated; the whole-process dynamic monitoring and abnormity identification of the parameters are realized, the monitoring pressure of operators is reduced, and the working efficiency is improved; the influence of human factors on the monitoring result is reduced, the inherent limitation of people is eliminated, and the parameter monitoring is more comprehensive and timely.

An ultra-real-time calculation method for axial power deviation, an electronic device and a storage medium. The method comprising: setting a boundary condition; determining interface data on the basis of the boundary condition; initializing a simulation reactor core physical model based on the interface data and in combination with real-time data; using the interface data in the simulation reactor core physical model to performing ultra-real-time calculation, and outputting and displaying the result of the ultra-real-time calculation. Using the described method, a trend for a next half-hour may be predicted, and a preview and verification platform that provides an axial power adjustment response scheme in the event of an accident for nuclear power plant operators is able to predict dynamic characteristics of the reactivity of a power plant and the impact on neutron flux distribution during accident conditions relatively accurately, simulating in advance the course of development of various accident conditions, and predicting various risks for a later period, thus allowing the operators more sufficient time to reduce or even eliminate risk.

A method for transitioning a nuclear reactor during initial cycle startup to a power generating state is disclosed. The method includes placing the reactor in a zero power state, omitting a low power physical test (LPPT) for the current cycle of the reactor based on a predetermined set of criteria, and transitioning the reactor to the power generating state without conducting the LPPT based on a cross-check. The cross-checking includes predicting a first set of values ​​for elements of the LPPT using a first design code, developing empirical equations for the elements of the LPPT using data from past cycles of the reactor, predicting a second set of values ​​for the elements of the LPPT using the empirical equations, and cross-checking the first values ​​against the second values.

The present application relates to a method and apparatus for quantifying indication signals of a nuclear instrument system, and a device, and a storage medium. The method comprises: simulating the power distribution of a reactor core under various operating conditions, so as to obtain power distribution parameters of the reactor core under the various operating conditions; and according to the power distribution parameters of the reactor core under the different operating conditions and according to corresponding benchmark power distribution parameters, quantifying indication signals of the reactor core. The use of the present method allows for quantifying the uncertainty of indication signals of the reactor core by means of theoretical analysis, allowing for obtaining more-accurate indication signals; moreover, the present solution does not require the collection of actual measurement data under actual operating conditions, and is suited to controlling novel reactor cores lacking actual measurement data and in designing constant values of protection signals, and therefore the present solution is highly applicable.

The present application relates to a kind of nuclear power plant DCS control system, including signal acquisition unit and main control room operating platform, main control room operating platform includes the control rod control system interface operating module of intermediate variable display in control rod control system;Feedwater deaerator system interface operating module for the intermediate variable display of feedwater deaerator system;Main feedwater system interface operating module for the intermediate variable display of main feedwater system;Turbine bypass discharge system interface operating module for the intermediate variable display of turbine bypass discharge system;Output link's reactor core margin calculation system operating interface link module;Chemical volume control system interface operating module for the intermediate variable display of chemical volume control system;Boron and water supply system interface operating module for the intermediate variable display of boron and water supply system.The present application can assist operator to make judgment and intervention in advance to system state by increasing intermediate variable display;More intuitively master system state, reduce human error.

The present invention belongs to the instrumentation and control technology of nuclear power plants, and specifically relates to a safe and efficient method for testing the protection channel of a nuclear power plant turbine, including optimizing the transmission of the trip protection channel and designing the test logic configuration of the trip protection channel; optimizing the transmission of the trip protection channel means changing all trip protection signals across platforms and controllers from network transmission to hard-wired transmission; designing the test logic configuration of the trip protection channel means adding a test logic function design to the digital configuration. This method greatly reduces the risk of human error during the channel test, avoids degradation of the protection channel, reduces the manpower and time costs in the work, and saves the critical path of overhaul. Starting from the optimization of the transmission mode of the trip protection channel and the design of the channel test logic, the first turbine protection channel test of a domestic nuclear power project was realized.

Disclosed herein is a stellarator comprising two sets of coils, namely a set of encircling coils which encircle the plasma axis, and a set of shaping coils which do not encircle any other coil or the plasma. In some embodiments, the encircling coils include a structural element to maintain their shape under magnetic forces. In some embodiments, the shaping coils are mounted onto one or more structural elements which, together with the shaping coils, constitute a field shaping unit. Also disclosed is a controller which may modify the electrical current flowing in one or more subsets of the coils in order to achieve target plasma parameters. Also disclosed is a method of designing a set of shaping coils by discretizing a surface dipole or current potential distribution.

The application discloses a nuclear power plant rotating equipment intelligent fault diagnosis method and system based on digital twinning, relates to the technical field of intelligent nuclear power plants, and realizes mutual complementation of digital space and real space data through deep learning, solves the problem of insufficient nuclear power plant rotating equipment fault data, realizes multi-source data integration of a diagnosis model based on the idea of data fusion, and can improve the accuracy of diagnosis results; the application also proposes an updating method of a digital twinning model, so that the updating of the digital twinning model can be realized when new working conditions are generated in the operation process of the nuclear power plant rotating equipment or the trained twinning model cannot meet the requirements, thereby guaranteeing the consistency of the digital twinning model and the physical space model.

The invention relates to the technical field of mechanical maintenance, in particular to a method and system for correcting the average temperature parameter of a primary loop temperature probe, a medium and equipment. The method for correcting the average temperature parameter of the loop temperature probe comprises the following steps: acquiring the average temperature of the probe during standard thermal shutdown to obtain the first average temperature of the probe; acquiring a probe average temperature at a preset temperature to obtain a second probe average temperature; judging whether the average temperature of the second probe is consistent with the preset temperature; if not, correcting an average temperature parameter in an average temperature linear formula in the temperature probe by using the average temperature of the first probe and the average temperature of the second probe; and substituting the corrected average temperature parameter into an average temperature linear formula to obtain a correct average temperature linear formula. According to the invention, the average temperature parameter is corrected by using the temperature of the first probe as a standard, so that the accuracy of the temperature probe is improved, and the safety and stability of the unit in the operation process are improved.

The invention relates to an array type monitoring device for a loose part of a nuclear power station. The array type monitoring device comprises a channel selection array module and a channel switching array module which are integrated in a processor. The channel selection array screens target signals from multiple sound channels and playback channels through a preset logic (based on signal intensity and key areas), and supports single-channel extraction or multi-channel fusion; the channel switching array realizes smooth interference-free switching between real-time signals and between real-time and historical playback signals based on a special algorithm. The device adopts a 2U rack type design and is internally provided with a dual-power-amplifier system; a D-type power amplifier drives left and right sound channel loudspeakers by a high-frequency switch technology, and the power consumption is less than or equal to 75W; the class AB power amplifier drives an earphone interface through linear amplification, and the distortion rate is lt; 6%. And the integrated storage module supports historical sound data backtracking analysis. The problems that existing equipment is insufficient in multi-channel processing, inflexible in switching and poor in power amplifier adaptability are solved, and the comprehensiveness, timeliness and accuracy of nuclear power station loose part monitoring are comprehensively improved.