87 results about "Nuclear power system" patented technology

The invention discloses a nuclear power electronic information multi-source data fusion and visual analysis method, and relates to the technical field of nuclear power, and the method comprises the steps: S1, data collection; s2, data fusion; s3, feature extraction and selection; s4, performing data analysis and modeling; s5, carrying out visual design; s6, realizing visualization; and S7, result interpretation and application. And S8, performing feedback and optimization. According to the nuclear power electronic information multi-source data fusion and visual analysis method, multi-source data fusion can be easily realized, so that related personnel are helped to better understand and analyze the operation state of a nuclear power system, the accuracy and efficiency of decision making are improved, and complex data can be further promoted to become intuitive and easy to understand in cooperation with application of data visualization. Meanwhile, the nuclear power data visualization configuration software has flexible data access and processing capabilities, can adapt to continuously changing data and business requirements, and further improves the multi-source data fusion and visualization analysis effects.

The invention relates to the technical field of nuclear power plant system design, in particular to a nuclear power system design drawing generation method and device, electronic equipment and a storage medium. According to the nuclear power system design drawing generation method provided by the embodiment of the invention, design control demand data of a target nuclear power control system is acquired; performing control key point analysis based on the design control demand data to obtain target control key point data; performing schematic diagram construction on the target nuclear power control system according to the target control key point data to obtain a target system schematic diagram; in response to a condition that the target system schematic diagram meets a system principle verification condition, performing application diagram construction on the target nuclear power control system according to the target system schematic diagram and the target control key point data to obtain a target system application diagram; and in response to the condition that the target system schematic diagram meets the system application verification condition, determining the target system application diagram as a target system design diagram of the target nuclear power control system. Therefore, the efficiency of generating the design drawing of the control system of the nuclear power plant can be improved.

The invention relates to a nuclear power system equipment nuclear safety performance evaluation method, which comprises the following steps of: obtaining a complete system list of a power plant, and screening the system list according to a nuclear safety function correlation condition to obtain a function list meeting a nuclear safety function requirement; classifying the function list based on the operation state to obtain a category system classification of the nuclear safety related system equipment meeting the operation state classification standard; performing performance index making according to the category classification to obtain a performance index making result of the nuclear safety related system equipment; and performing nuclear safety performance evaluation on the nuclear safety related equipment in the preset time period according to the performance index formulating result to obtain a nuclear safety performance evaluation result in the preset time. Through systematic equipment screening and classification, scientific and reasonable performance index formulating and dynamic performance evaluation based on actual operation data, the nuclear safety performance of the nuclear safety related equipment in the preset time period is evaluated; and refined evaluation of nuclear power plant system equipment nuclear safety performance is realized.

Reactor buildings and vessel systems are disclosed. A nuclear power system includes: a building structure that comprises at least two exterior side walls and two end walls, at least one of the exterior walls angled non-orthogonally relative to a floor of the building structure, the at least two exterior walls and two end walls defining an interior volume of the building structure; one or more nuclear reactor systems mounted at least partially in the interior volume of the building structure; and one or more heat exchanger systems mounted at least partially to at least one of the exterior walls. A nuclear reactor vessel system includes: a nuclear fission reactor; an inner vessel that defines an inner volume sized to at least partially enclose the nuclear fission reactor; and an outer vessel sized to wholly or substantially enclose the inner vessel, the inner vessel being removable from the outer vessel.

The present invention discloses an optimization method for a marine secondary loop system based on an improved multi-objective particle swarm algorithm. By establishing mathematical models of the main and auxiliary equipment of the secondary loop system and a mathematical model for system heat balance calculation, and coupling the mathematical models of each equipment, a complete mathematical model of the secondary loop system is obtained; based on parameter sensitivity analysis, parameters that have a significant impact on the equipment structure and system performance are selected as optimization variables; the optimization objectives are the lightest weight, smallest volume, and highest effective efficiency of the secondary loop system; under the condition of meeting the determined constraints, an adaptive multi-objective particle swarm algorithm based on angular penalty distance is used for multi-objective optimization design of the weight, volume, and effective efficiency of the secondary loop system, and finally a Pareto optimal solution set and a Pareto front are obtained, and the relative optimal solution is selected by the Technique for Order Preference by Similarity to an Ideal Solution (TOPSIS method). The present invention better realizes the matching modeling and optimization of the marine nuclear power secondary loop system based on the improved algorithm, and is of great significance for improving the overall performance of the nuclear power system.

The invention discloses a multi-working-condition division method and system for nuclear power data based on a TSNE-improved HBBT model and a storage medium, and belongs to the technical field of intelligent operation and maintenance of nuclear power stations. The method comprises the following steps: firstly, acquiring historical operation data of a nuclear power system through a simulation platform, and extracting a mean value, a variance, a slope, an autocorrelation coefficient and an energy characteristic after normalization processing; dividing the feature data into a training set and a test set, and endowing working condition labels; carrying out nonlinear dimension reduction on the training set by adopting T-SNE; and finally, training and classifying by using an improved histogram gradient lifting tree model. The improved HBBT innovatively adopts an adaptive feature binning technology and a sample difficulty and cluster consistency weighting strategy, so that the model precision and robustness are effectively improved. The method solves the problems that an existing unsupervised method is high in randomness and inaccurate in classification, can accurately recognize the steady-state and transition working conditions of the nuclear power system, and provides reliable support for intelligent operation and maintenance.

The invention relates to the technical field of nuclear power reactors, and discloses a nuclear power unit secondary circuit support system and a once-through steam generator nuclear power system. According to the nuclear power unit secondary loop support system, a plurality of groups of steam generator output pipelines are communicated with a main steam pipeline in parallel; the start-stop reactor main pipe is communicated with the output pipeline of the steam generator through a start-stop branch, a steam-water separator is mounted on the start-stop reactor main pipe in parallel, a steam supply branch is communicated between the start-stop reactor main pipe and the main steam pipeline, the steam supply branch is communicated with a steam exhaust pipeline, and two groups of steam exhaust outlets are formed in the tail end of the steam exhaust pipeline in parallel; and the drainage pipeline is communicated with the liquid phase outlet of the steam-water separator, and two groups of drainage outlets are arranged on the drainage pipeline in parallel. Start-stop steam exhaust of multiple groups of steam generators is treated through the shared start-stop reactor main pipe and the steam-water separator, graded recycling of working media and heat energy in the start-stop process is achieved, vapor-phase media are recycled, and waste of heat energy resources and water resources in the initial stage of start is reduced.

The present invention discloses a startup method for a nuclear power system with a heat pipe reactor coupled to a supercritical CO2 Brayton cycle. The cold startup of the nuclear power system with a heat pipe reactor coupled to a supercritical CO2 Brayton cycle includes the startup of the core heat pipe, the startup of the compressor, and the loop switching. The system is an active startup. First, the startup of the core heat pipe is regulated through the rotating drum control system to achieve the autonomous control of the core power and temperature; the stability of the compressor inlet temperature and pressure during the startup process is maintained through the control of the compressor inlet temperature and pressure; the rotational speed of the rotating shaft is controlled to increase step by step; the loop flow rate of the system is controlled by the throttle valve at the compressor inlet to achieve the regulation of the loop flow rate; the control of the system circulation loop switching is achieved through the control of the opening degrees of the turbine bypass valve and the throttle valve, so as to realize the switching of the working medium from the bypass branch to the turbine branch. The present invention realizes the ability to control the key parameters of the cold startup of the heat pipe reactor coupled supercritical CO2 Brayton system, and achieves the startup of the system under the condition of small disturbances.

The invention relates to a nuclear reactor pressure relief spraying device and a nuclear reactor pressure relief system. The nuclear reactor pressure relief system comprises the nuclear reactor pressure relief spraying device. In the nuclear reactor pressure relief spraying device, the plurality of main spraying holes are formed in the linear main pipe, and the plurality of pre-spraying holes are formed in each arc-shaped branch pipe, so that the condensation rate of high-temperature and high-pressure steam can be effectively improved, and the operation stability of a nuclear power system is ensured. Secondly, compared with the mode that steam is directly discharged through a large notch in the prior art, steam is discharged through a plurality of uniform main spraying holes and a plurality of uniformly-distributed pre-spraying holes, so that the size of bubbles formed after the steam is discharged through hole bodies is limited, and then water hammer impact is effectively reduced; and the operation stability in the pressure relief process is improved.

A nuclear power system 100 comprising: a nuclear heat source 10, a turbine 15, a heat dissipator 20, a compressor 25, and an electric heater 35, all connected in a fluid flow circuit for a working fluid, the turbine connected to the compressor and a generator 40 by a connecting shaft 30; a sensor system 55 connected to a controller 60, and configured to acquire data from the electric heater, nuclear heat source, generator, a battery 50, and a load 45, to determine: the temperature of the fluid as it exits the electric heater; the temperature, pressure, and massflow (condition) of the fluid as it exits the nuclear heat source; the energy being generated; the battery charge; and the energy demand of the load; and report this data to the controller which is configured to control the distribution of the energy generated between the load, battery, and electric heater, such that the energy output to the grid is varied as the reactivity levels within the nuclear heat source are adjusted, whilst maintaining a constant condition of the fluid entering the turbine.

The invention relates to the technical field of nuclear power generation, in particular to a nuclear power system based on energy storage and a control method thereof.The system comprises a nuclear power subsystem, an energy storage subsystem and a control unit; the nuclear power subsystem comprises a nuclear reactor, a steam generator, a pump and a steam turbine which are sequentially connected and is used for driving the steam turbine to generate power through heat energy generated by the nuclear reactor, and electric energy generated by the steam turbine is sent to a load; the energy storage subsystem comprises a fused salt energy storage unit and a super capacitor energy storage unit; the fused salt energy storage unit is thermally coupled with a steam loop between the steam generator and the steam turbine and is used for absorbing or releasing heat energy to adjust steam parameters; the super capacitor energy storage unit is electrically connected with the power output end of the steam turbine and used for storing or releasing electric energy. And the control unit is used for dynamically adjusting the energy interaction of the nuclear power subsystem and the energy storage subsystem according to the load demand to realize load tracking, and is used for controlling the power of the nuclear reactor.

This application discloses a large-scale nuclear power system reliability analysis method based on parallel computing, belonging to the field of nuclear power system reliability analysis. The method comprises the following steps: constructing a nuclear power system fault tree model T based on the functional principle, failure mode, and reliability logic relationship of the nuclear power system; using a fault tree block method to identify independently analyzable fault tree subtrees in the nuclear power system fault tree model T, and constructing a vertex directed graph of the independently analyzable fault tree subtree vertices based on the connection relationship of the fault tree subtrees; using the vertex directed graph and a topological sorting method to identify the analysis order of the fault tree subtrees, and then analyzing the nuclear power system fault tree model T in parallel according to this order to obtain the probability of occurrence of the top event of the nuclear power system fault tree model T. The present invention can be applied to the analysis of large-scale fault tree models in nuclear power systems.

The invention discloses a multi-cooling-medium parallel heat exchanger and a nuclear power system. The multi-cooling-medium parallel heat exchanger comprises a main body, channel partition plates and protruding parts. An accommodating space is formed in the main body; the channel partition plates are arranged in the containing space so as to isolate a plurality of working medium channels distributed in parallel in the containing space, the working medium channels comprise cooling working medium channels and high-temperature working medium channels, and the high-temperature working medium channels are located among the cooling working medium channels in which different cooling working media circulate; the protruding parts are arranged on the channel partition plates corresponding to the high-temperature working medium channels, and the geometric parameters of the protruding parts making contact with different cooling working media are different so that the different cooling working media can meet the working temperature intervals of the different cooling working media. According to the heat exchanger, under the condition that the macroscopic size of the channel and the material characteristics are not changed, synchronous heat transfer of one high-temperature working medium to multiple refrigerating working media can be achieved, the purpose of synchronous heat exchange through different cooling working media is achieved, the energy utilization efficiency is improved, and the operation cost is reduced.

The invention relates to the technical field of high-temperature gas cooled reactor primary helium fans, in particular to a high-temperature gas cooled reactor primary helium fan frequency converter, a nuclear power system and a method for reducing conductivity. A high-temperature gas cooled reactor primary helium fan frequency converter comprises an expansion water tank; the power assembly comprises a conveying pipeline, and fluid is suitable for flowing in the conveying pipeline; the filtering assembly comprises a filtering pipeline, a first valve and a deionization barrel, one end of the conveying pipeline is communicated with the liquid outlet end, the other end of the conveying pipeline is communicated with the filtering pipeline, the other end of the filtering pipeline is communicated with the liquid inlet end of the expansion water tank, and the filtering pipeline is sequentially provided with the first valve and the deionization barrel in the direction from the conveying pipeline to the expansion water tank; the exhaust assembly comprises an exhaust pipeline and an exhaust valve, the exhaust valve is arranged on the exhaust pipeline, and one end of the exhaust pipeline is communicated with the filtering pipeline. The invention provides a high-temperature gas cooled reactor primary helium fan frequency converter, a nuclear power system and a method for reducing conductivity, and aims to solve the problem of siphoning in the process of replacing a deionization barrel.

The application provides a voltage tracking method, a control device and a UPS system. The method is used for tracking grid voltage. The method performs the following steps in each capture cycle: obtaining a captured current voltage cycle; if a target cycle is stored in a preset storage space, adding the current voltage cycle to the target cycle to obtain a first voltage cycle; if the first voltage cycle is not in a preset range, performing an abnormal number count; if the abnormal number does not exceed a preset number, tracking voltage according to a voltage cycle used in voltage tracking of a last capture cycle; and if the abnormal number exceeds the preset number, stopping voltage tracking of the grid voltage. The application considers the influence of grid harmonics and can effectively realize grid voltage tracking. On this basis, the application can be effectively applied to the nuclear power field, effectively ensures stable operation of electrical equipment for nuclear power, and further ensures stable operation of a nuclear power system.

The invention discloses a safety assessment method and device for a nuclear power system, and the method comprises the steps: obtaining the first information of a to-be-analyzed nuclear power system, and the first information comprises the identifications of a plurality of parts in the nuclear power system, the state type, and the equipment parameters of the parts identified by the plurality of part identifications; constructing a fault tree according to the first information; according to the equipment parameters of the parts corresponding to the multiple basic events in the fault tree, a first corresponding relation is determined, and the first corresponding relation comprises multiple first operation durations of the parts corresponding to all the basic events and first failure probabilities corresponding to all the first operation durations; simulating the accident evolution process of the nuclear power system according to the first corresponding relation to obtain simulation information; according to the simulation information, the safety of the nuclear power system is evaluated to obtain the evaluation result, and the event tree and the fault tree are associated, so that the problem that the static probability and the dynamic evolution are disjointed is solved, the path probability is more practical, the accuracy of the safety evaluation result is improved, and the deviation is effectively reduced.

The invention belongs to the technical field of nuclear power source systems, aims to solve the problem of significant noise caused by a secondary loop steam turbine in an existing nuclear power system, and discloses a liquid metal reactor nuclear power source system based on temperature difference thermoelectric conversion. Comprising a liquid metal reactor island, a reactor core waste heat removal cooling system, a temperature difference thermoelectric conversion module and a temperature difference thermoelectric module cooling system, heat generated by the liquid metal reactor nuclear island is transferred to the temperature difference thermoelectric conversion module through liquid metal, redundant heat is discharged through the reactor core waste heat removal cooling system, and the liquid metal flows through the temperature difference thermoelectric conversion module to transfer heat and flow back to the liquid metal reactor nuclear island to form heat exchange circulation. The temperature difference thermoelectric conversion module obtains heat transferred from the liquid metal to generate electric energy output. According to the invention, extremely low noise in the operation process is realized, and the method is especially suitable for application scenes with harsh requirements on concealment and acoustic performance.

A nuclear power system includes a borehole nuclear reactor, including: a reactor core comprising nuclear fuel and positioned a borehole that is formed from a terranean surface to a subterranean formation; a neck that extends upward from the reactor; a heat exchanger positioned at the top of the neck and in thermal communication with the reactor core; and a pressurizer positioned in the borehole downhole of the reactor core or including a tube or pipe that rises near to the terranean surface. The pressurizer is configured to maintain a reactor pressure at a pressure of water at an operating depth in the subterranean formation. The reactor core, the neck, the heat exchanger, and the pressurizer can be connected to form a preassembled kit prior to installation in the borehole, or are configured to be assembled as parts are lowered into the borehole.

A micro integral nuclear reactor and nuclear power system including the micro integral nuclear reactor. The micro integral nuclear reactor can include a reactor core, a steam generator, primary coolant circuit, and a reactor pressure vessel. The reactor pressure vessel can houses the reactor core, the steam generator, and the primary coolant circuit. The fuel of the reactor core can be enriched to a concentration of 5 percent or greater and less than 20 percent U-235 by weight of uranium within the reactor core. When operating, the micro integral nuclear reactor can produce a thermal power output from 6 megawatts-thermal (MWt) to 60 MWt. The nuclear power system can in lude a turbine-generator assembly fluidly coupled to the steam generator and, when operating, the nuclear power system produces an electrical power output from 2 megawatts-electric (MWe) to 20 MWe.

The invention provides a ship sensing data optimization method and device, and the method comprises the steps: determining a correlation calculation coefficient of each parameter sensor based on the distribution characteristics of to-be-optimized ship data of each parameter sensor, and then decomposing the to-be-optimized ship data based on a variational mode decomposition algorithm, the correlation between the ship data to be optimized and a plurality of intrinsic mode functions is quantitatively evaluated based on correlation calculation coefficients corresponding to the parameter sensors, and finally partition optimization reconstruction is performed based on the correlation, so that multi-source ship data optimization which automatically adapts to data features and dynamically processes real-time data is realized. The method effectively overcomes the dependence of a traditional method on artificial experience, can carry out efficient and accurate denoising processing on sensor data of a marine nuclear power system under a complex background, remarkably improves the data quality, provides more reliable data input for subsequent ship state analysis, evaluation and control decision, and improves the reliability of data processing. Therefore, safe and stable operation of the nuclear power system is guaranteed.

This invention relates to the field of nuclear power system piping analysis technology, and discloses an atmospheric pressure visualized downward horizontal pipe test system, including a test section pipeline, a control pipeline, a water tank pipeline, and an infrared thermal imaging monitoring system. The test section pipeline includes a first main pipeline, with at least one first branch pipeline corresponding to it. The two ports of the first main pipeline are connected to the two ports of the control pipeline, and the outlet end of the water tank pipeline is connected to the control pipeline. The cross-sectional shape of the first main pipeline is fan-shaped or rectangular, and its cross-sectional area is smaller than that of the main pipeline of the actual research target. The water tank pipeline provides coolant at a preset temperature to the control pipeline and the test section pipeline. The control pipeline controls the temperature and flow rate of the coolant. The infrared thermal imaging monitoring system is installed on the target pipe section of the test section pipeline to acquire the temperature information of the target pipe section. This invention mainly solves the technical problem that the flow velocity in experimental pipelines is difficult to increase.

A maskless Transform framework-based intelligent identification method for operation events of a small nuclear power system comprises the following steps of: 1, acquiring multi-source heterogeneous time sequence data related to an operation state by using a sensor system or a simulation platform carried by the small nuclear power system to form an original data set; 2, preprocessing the original data set; 3, creating a linear embedded layer network, and mapping a multi-dimensional parameter vector in the sample sequence to a high-dimensional feature space by using the network to form an event data embedded vector; 4, performing position coding on the event data embedding vector, and adding time sequence information in the sample sequence into the event data embedding vector; (5) a maskless Transform encoder network formed by stacking N identical encoder layers is built, and N is a positive integer greater than or equal to 1, and N is a positive integer greater than or equal to 1; 6, the event data embedding vector is input into a maskless Transform encoder network, and a feature sequence is output; and 7, calculating the probability that the sample sequence belongs to each operation event category through a Softmax function, and judging the category with the maximum probability as a final recognition result. The method is high in recognition accuracy and strong in time sequence characteristic capturing capability.

The application relates to an ice plug pipeline impact vibration test device and belongs to the field of nuclear power ice plug pipeline impact vibration tests, which comprises clamps, vertical columns, cross clamps, magnetic attraction devices, metal balls, vibration sensors, ice plug pipelines, infrared sensors and posture sensors; the ice plug pipelines are symmetrically provided with detachable clamps on two sides; vertical columns and posture sensors with scale marks are vertically installed on the two clamps respectively, and the vertical columns are confirmed to be vertically installed on the corresponding clamps through the posture sensors; cross clamps are detachably installed on the two vertical columns; the two cross clamps are transversely connected; the magnetic attraction devices are hung between the transversely connected two cross clamps, and the metal balls are adsorbed on the magnetic attraction devices; infrared sensors are installed on the two cross clamps respectively, and the horizontal moving distance and falling height of the metal balls are confirmed through the infrared sensors; and the vibration sensors are detachably installed on the ice plug pipelines. The application realizes analysis of ice plugs in a pipeline isolated by ice plugs in a nuclear power system under different external impact loads.

The application provides a nuclear energy electric supercharged jet propulsion system and belongs to the field of aviation power. The nuclear power propulsion system of the traditional turbine engine type solves the problems that the working processes of a compressor and a turbine influence each other, the matching of the compressor and the turbine is relatively difficult, the performance of the propulsion system under variable working conditions is poor, and the arrangement of a heat exchanger in a closed cycle nuclear power system is relatively complex and the flow resistance of the heat exchanger is relatively large. The nuclear energy electric supercharged jet propulsion system comprises a regenerator, a cold end inlet connected with an outlet of a helium-xenon compressor, a cold end outlet connected with a cold end inlet of a heater, a hot end outlet of a first heat exchanger connected with a hot end inlet of a fuel heat exchanger, a second heat exchanger, a hot end inlet of the second heat exchanger connected with a circulating working medium outlet of a reactor system, a fuel tank, a discharge port of the fuel tank connected with the cold end inlet of the fuel heat exchanger through a fuel valve, and an engine shell, wherein air from the compressor outlet flows through the first heat exchanger and the second heat exchanger in sequence. The nuclear energy electric supercharged jet propulsion system is mainly used for the propulsion system of aviation equipment.

The embodiment of the invention relates to the technical field of nuclear reactor design, in particular to an arrangement structure of a reactor nuclear power system and a reactor workshop.The arrangement structure comprises a purification and separation device which is arranged to purify and separate mixed gas in a reactor to obtain gaseous fuel and foreign gas; the fuel generating and conveying device is used for generating gaseous fuel, inputting the gaseous fuel into the reactor and inputting the gaseous fuel purified and separated by the purifying and separating device into the reactor; the thermoelectric conversion device is arranged to convert heat generated by the gaseous fuel in the reactor into electrical energy. According to the embodiment of the invention, the reactor, the purification and separation device, the fuel generating and conveying device and the thermoelectric conversion device are subjected to integrated same-site same-plant design, so that the circulation management of production, conveying, reaction and purification and separation of fuel can be realized, and the comprehensive utilization of energy and the whole-process management of nuclear fuel circulation are realized; the economical efficiency of reactor operation is improved, and leakage of radioactive substances can be reduced.

The present application provides a passive residual heat removal system for a liquid metal cooled reactor. The passive residual heat removal system includes an independent heat exchanger, an air cooler, an intermediate circuit, and a plurality of Stirling thermoelectric converters. The air cooler includes a damper. The hot end of the Stirling thermoelectric converter is inserted into the air cooler for heat exchange with the intermediate circuit. The adiabatic end of the Stirling thermoelectric converter is located on the inner wall surface of the air cooler, and the cold end of the Stirling thermoelectric converter is located outside the inner wall surface of the air cooler. When an accident causes a power outage, the damper automatically opens, and electricity is generated by using the temperature difference between the cold end and the hot end of the Stirling thermoelectric converter for emergency equipment applications. By adding Stirling thermoelectric converters to the passive residual heat removal system, the present application fully utilizes the advantage of the relatively high temperature of the liquid metal coolant in the liquid metal cooled reactor, enabling the Stirling thermoelectric converter to generate electricity using the residual heat for use by emergency equipment in the nuclear power plant, and reducing the complexity of the nuclear power system.

The invention discloses a nuclear power reliability analysis system and method based on evidence Bayesian, and relates to the crossing field of nuclear energy engineering and computer-aided reliability analysis. The nuclear power reliability analysis system based on evidence Bayes comprises a system GO graph construction module, a GO graph-DBN conversion module, a common cause failure analysis module and a node importance analysis module which are sequentially connected from front to back in a data flow mode. The nuclear power reliability analysis method based on evidence Bayesian is applied to the nuclear power reliability analysis system based on evidence Bayesian. The method comprises the following steps: constructing a system GO graph; converting the GO graph into a DBN; and common cause failure modeling. The method solves the four major problems of a traditional method in reliability analysis of the nuclear power system. The method has the core advantages that an interval reliability result reflects the uncertainty of data and expert experience; the model complexity is controllable, the calculation efficiency is high, and the method is suitable for complex systems.

The present invention discloses a nuclear power system based on energy storage and ejection energy release. The input end of the reactor is connected to the Brayton cycle subsystem, and the output end of the reactor is connected to the turbine of the Brayton cycle subsystem via the ejection subsystem. The turbine is connected to the main compressor via a drive-generator motor. When the system starts and stops, the drive-generator motor operates as an electric motor, consuming electricity to drive the main compressor to start; when there is net power output, the drive-generator motor operates as a generator, converting the kinetic energy of the motor shaft into electrical energy output. The present invention can reduce the frequent adjustment of the reactor when the user's power demand fluctuates, ensure that the main compressor, recompression compressor, and turbine operate in a high-efficiency range, and improve the energy utilization efficiency of the nuclear power system under variable operating conditions.

The embodiment of the invention provides a data reconstruction method and device, electronic equipment and a storage medium, and belongs to the technical field of data processing. The method comprises the following steps: acquiring a measurement data set of a nuclear power system; determining a constraint equation set according to at least one of the law of conservation of mass and the law of conservation of energy; the constraint equation comprises at least one second parameter and at least one first parameter in the measurement data set, and the second parameter refers to an unmeasured parameter not included in the measurement data set; for each first parameter, calculating according to a parameter value of the first parameter to obtain a relative residual error of the first parameter; constructing a data reconstruction model of the nuclear power system according to the constraint equation set and the relative residual error of each first parameter; the data reconstruction model is solved, reconstruction data are obtained, and the reconstruction data comprise the reconstruction value of each first parameter and the estimation value of each second parameter. According to the embodiment of the invention, the accuracy of the measured data can be improved.