158results about "Computer implemented nuclear control" patented technology

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.

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.

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.

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.

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.

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.

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 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 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.

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.

Example aspects are directed to a nuclear power plant. The nuclear power plant comprises one or more first nuclear sources configured to produce power to supply a utility distribution system for use, wherein each of the one or more first nuclear sources includes a first subsystem configured to provide operational functionality to a corresponding first nuclear source; and one or more second nuclear sources configured to produce power to the one or more first subsystems and to run the nuclear power plant. Such a nuclear power plant may be safer and more reliable because of incorporating small nuclear sources produce power to the one or more first subsystems for operational functionality. Furthermore, fossil fuel backup or grid backup may be removed from the nuclear power plant, which may help to decrease complexity and cost of operation and of internal electricity distribution in the nuclear power plant.

Disclosed herein is a stellarator with two sets of coils: a set of encircling coils that encircle the plasma axis, and a set of shaping coils that do not encircle any other coils or plasma. In some embodiments, the encircling coils include structural elements to maintain their shape under magnetic forces. In some embodiments, the shaping coils are attached to one or more structural elements that, together with the shaping coils, constitute a magnetic field shaping unit. Also disclosed is a controller that can modify the current through one or more subsets of the coils to achieve target plasma parameters. Also disclosed is a method for designing the set of shaping coils by discretizing the surface dipole or current potential distribution.

A method and system for an in-vessel natural circulation alkali metal reactor system, a cleanup system, and associated methods are disclosed. The reactor vessel system includes an inner vessel defining an interior volume sized to at least partially surround a nuclear reactor. The reactor includes a plurality of nuclear fuel elements at least partially surrounded by cladding, and the reactor is cooled by a liquid metal coolant in a primary coolant loop. A pool of immersion fluid occupies a volume inside the inner vessel. The reactor vessel system includes an outer vessel sized to completely or substantially surround the inner vessel. The nuclear reactor power system includes a reactor core including an active fuel region and a rotatable drum including at least one of a neutron absorbing material, a neutron leakage enhancing material, or a neutron reflecting material, the rotatable drum being positioned outside the active fuel region of the reactor core.

This invention discloses a computer apparatus whose instructions describe a process which analyzes system entropy flows, irreversible losses and Carnot reversibilities associated with heat exchangers used in thermal engines. Irreversible losses include those from shell and tube heat exchangers, and those from the shell-side of heat exchangers such as condensers. This disclosure teaches revision to the classic Carnot Engine resulting in an Exergetic Engine. The Exergetic Engine was created by recognizing true thermodynamic irreversibility associated with any heat exchanger is determined by the summation of its internal exergy flows. This leads to a correction of Sadi Carnot's THot. For the nuclear engine, his TCold is redefined as a Fixed TRef dependent on neutronic constants and reactor coolant properties. Correcting his 200 year-old teachings produce an irreversible loss and an Exergetic Reversibility applicable to any heat exchanger used in any thermal engine.

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 present disclosure relates to a stellarator (1) comprising: a plurality of non-planar, superconducting coils, for generating a magnetic field for confining a plasma inside a plasma vessel; and a coil support structure (30, 40, 50) comprising: a first structural arrangement (30) arranged between a center (20) of the stellarator and the plasma vessel and configured to balance a radial force exerted by the magnetic field on the plurality of coils; a second structural arrangement (40, 50) configured to balance local forces exerted by the magnetic field on one or more coil segments of the plurality of coils; wherein the local forces cause at least a toroidal motion and / or a toroidal deformation of the one or more segments of the plurality of coils.

Example aspects are directed to a nuclear power plant. The nuclear power plant comprises one or more first nuclear sources configured to produce power to supply a utility distribution system for use, wherein each of the one or more first nuclear sources includes a first subsystem configured to provide operational functionality to a corresponding first nuclear source; and one or more second nuclear sources configured to produce power to the one or more first subsystems and to run the nuclear power plant. Such a nuclear power plant may be safer and more reliable because of incorporating small nuclear sources produce power to the one or more first subsystems for operational functionality. Furthermore, fossil fuel backup or grid backup may be removed from the nuclear power plant, which may help to decrease complexity and cost of operation and of internal electricity distribution in the nuclear power plant.

The present application relates to a method of optimizing a nuclear core loading pattern, comprising the following steps of training a model with results of a large number of core loading simulations, where each core loading simulation comprises a unique core loading pattern of fuel assemblies, to predict key measures for each unique loading pattern, so that the training results in the predictions of the model generally corresponding to the simulations, running the trained model with key measures in order to provide feasible results, and optimizing the loading pattern with the results of the trained model.

The application discloses a kind of based on data-driven small modular prismatic high temperature gas cooled reactor core optimization design method and system, comprising: constructing training data set;Training a data-driven proxy model is used to replace the execution high-cost simulation calculation;With a multi-objective evolutionary algorithm, and with the proxy model as its fitness function evaluator, to explore design space and generate a set of Pareto optimal candidate solution.The key features of the application are an iterative validation and model updating closed loop: the candidate solution is verified by high-fidelity simulation, and if the prediction accuracy does not meet the convergence criteria, the new data points after verification are expanded to the training data set, and the proxy model is retrained.The application can significantly reduce the calculation cost, efficiently explore the complex design space, and obtain a physically reliable and more optimal reactor design scheme.

The application relates to a method for constructing and controlling turbulence morphology in a reactor pressure vessel, which includes the following steps: S1, obtaining a vortex relevant data, obtaining a temporal distribution characteristic, a spatial distribution characteristic, a power spectral density of the turbulence, and turbulence energy spectrum at the inlet region of each fuel assembly; S2, obtaining the water gap variation of the fuel assembly, and determining whether it is needed to modify the three-dimensional flow field model of the fuel assembly based on the water gap variation of the fuel assembly until the water gap of the fuel assembly does not change; S3, obtaining a core nuclear power fluctuation amplitude among the fuel assemblies; S4, determining whether the core nuclear power fluctuation amplitude falls within a target range, and if not, modifying the internal three-dimensional structure of reactor pressure vessel until the core nuclear power fluctuation amplitude falls within the target range. The application can eliminate or improve the core nuclear power fluctuation, reduce the wear of the reactor core grid, and ensure the safety of reactor operation.

An abnormality detector comprises a specific section extractor, a feature amount calculator, and an abnormality diagnoser. The specific section extractor divides, into specific section data, phase current caused to flow through an electric motor used in a control rod drive device, and extracts the specific section data. The feature amount calculator calculates a feature amount for use in diagnosis in the abnormality diagnoser. The abnormality diagnoser diagnoses a presence or absence of abnormality of the control rod drive device based on the feature amount. The feature amount calculator calculates a natural frequency of the entire control rod drive device based on an electrical parameter of the electric motor and a mechanical parameter of the control rod drive device, and calculates the feature amount by using the natural frequency and an applied voltage frequency to the electric motor.

This system prevents the failure to detect abnormalities in the drive system due to the electrical parameters of the electric motor used in the drive system and the mechanical parameters of the drive system itself. [Solution] The anomaly detection device 20 comprises a specific section extraction unit 22, a feature quantity calculation unit 24, and an anomaly diagnosis unit 25. The specific section extraction unit 22 divides and extracts the phase current (motor phase current 31) flowing through the motor 11 used in the control rod drive device 10 into specific section data (section current data 32). The feature quantity calculation unit 24 calculates feature quantities to be used for diagnosis in the anomaly diagnosis unit 25. The anomaly diagnosis unit 25 diagnoses whether or not there is an anomaly in the control rod drive device 10 based on the feature quantities. The feature quantity calculation unit 24 calculates the natural frequency of the entire control rod drive device 10 based on the electrical parameters of the motor 11 and the mechanical parameters of the control rod drive device 10, and calculates feature quantities using the natural frequency and the applied voltage frequency to the motor 11.

The invention relates to a nuclear power equipment state monitoring method based on an on-line configuration technology, and the method comprises the following steps: S1, obtaining a primitive selected by a user from a special primitive library for nuclear power, and constructing a monitoring image; s2, dynamically binding the primitive with the multi-source system data of the nuclear power plant, wherein the dynamic binding comprises binding the primitive with a selected measuring point; and S3, state analysis is carried out on the data bound with the measuring points through an integrated equipment health state evaluation model, an equipment state monitoring result is obtained, and the system operation state is visually presented in real time. According to the method, the dynamic configuration technology and the equipment health state evaluation model are combined, construction of monitoring pictures, intelligent fusion of multi-source data and diagnosis of the equipment health state are achieved, and the problem that a traditional nuclear power monitoring system is limited in visualization effect is effectively solved.