389results 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.

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 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 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.

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 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.

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.

A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs; and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

The present invention provides a fast reactor nuclear power plant accident diagnosis design method, comprising: (1) based on the list of accident procedure documents, eliminating the procedure documents that are not suitable for execution guided by the accident diagnosis strategy, and forming a list of procedures to be executed guided by the accident diagnosis strategy; (2) on the basis of the list of procedures to be executed guided by the accident diagnosis strategy, sorting the accident procedures; and, based on the entry symptom information of the accident procedures, forming the entry symptom information of the accident diagnosis strategy according to the screening principle, as a criterion for judging whether to start executing the accident diagnosis strategy; (3) formulating the operation flow of the accident diagnosis strategy according to the sorted list of procedures to be executed guided by the accident diagnosis strategy and the entry symptom information of the accident diagnosis strategy. The present invention can be used to guide the design and optimization of accident diagnosis content in the accident handling guidelines and accident procedure documents of fast reactor nuclear power plants, improve the accident procedures from the perspective of strategies and measures, and effectively improve the operation safety of fast reactor nuclear power plants.

The present invention relates to a method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident, comprising: (1) diagnosing a sodium-cooled fast reactor loss of heat sink accident, wherein the diagnostic strategy is as follows: among n steam generators, at least m steam generators have an outlet sodium temperature T i Rising above the threshold value θ; (2) determining the loop type of the sodium-cooled fast reactor nuclear power plant, for single-loop design or multi-loop design of the sodium-cooled fast reactor nuclear power plant, respectively using the corresponding process to handle the accident, at the same time, the complex and numerous sodium-cooled fast reactor nuclear power plant loss of heat sink type primary cause events are classified into two categories: partial loss of heat sink and total loss of heat sink, and a matching accident handling strategy is provided, thereby improving the efficiency and reliability of accident handling. The present invention can improve the degree of power plant automation operation, reduce the human influence in the power plant operation and accident diagnosis and processing process, improve the accident response speed, reduce the operator's operating task burden, and ensure the accuracy and timeliness of accident response.

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 invention discloses a configuration logic execution optimization method, system and device, a medium and a product, and relates to the technical field of data processing, and the method comprises the steps: configuring a configuration logic based on the current application demand of a DCS system of a user; optimizing the configuration logic based on a dependency relationship between algorithm blocks to obtain a plurality of optimized configuration logic data streams; on the basis of the dependency relationship between each optimized configuration logic data stream and the algorithm block, according to a data stream parallel principle, constructing a configuration logic relationship indication table in which a plurality of algorithm block names are stored; based on the configuration logic relation indication table, generating an algorithm dependency relation register corresponding to each algorithm block; and according to the algorithm dependency relationship in the algorithm dependency relationship register, carrying out parallel calling on a plurality of algorithm actuators so as to execute the configuration logic. The execution efficiency of the configuration logic is improved.

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.