58 results about "Nuclear power system" patented technology

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.

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 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.

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 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 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.

To provide a floating-type nuclear power generation system capable of advantageously resisting core meltdown.SOLUTION: The floating-type nuclear power generation system includes: a reactor; a turbine generator driven by steam generated by heat of the reactor; and a floating body in which the reactor and the turbine generator are arranged and which is moored on the sea, the floating body having, below the reactor, a core catcher filled with ballast water of a bottom ballast tank formed in the bottom portion of the floating body by a double-hull structure in the lower side.SELECTED DRAWING: Figure 8

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.

A method of operating a nuclear power system includes generating heat in a nuclear reactor core, transmitting the heat to a heat engine, generating electricity with a generator operatively coupled to the heat engine. The method further includes detecting a no-load condition, and stopping the heat engine. The method also includes transferring heat from an outer surface of the nuclear reactor to the environment through a heat transfer system if a temperature of the nuclear reactor rises above a threshold temperature. The method further includes preventing heat from transferring from the outer surface of the nuclear reactor to the environment through the heat transfer system if the temperature of the nuclear reactor is below the threshold temperature. Nuclear power systems and nuclear reactors are also disclosed.

This invention provides a multi-redundant space nuclear power source combined thermoelectric conversion system, belonging to the field of space power generation. It solves the problems of easy failure of liquid pumps in traditional space nuclear power systems and easy failure of compressors in closed Brayton systems, which in turn lead to the inability of detectors to function properly. It mainly includes a closed Brayton system and a liquid metal magnetohydrodynamic (MHD) power generation system. The closed Brayton system includes a compressor, generator, turbine, regenerator, radiator, and semiconductor thermoelectric generator. The liquid MHD power generation system includes a nuclear reactor, liquid metal pump, gas-liquid separator, MHD nozzle, gas-liquid two-phase mixing section, and MHD power generation channel. It effectively utilizes the advantages of high output power, strong system stability, and long service life of space nuclear power sources. This invention is also applicable to space missions such as deep space exploration.

The invention provides a nuclear power system simulation method and device, electronic equipment and a storage medium, and the method comprises the steps: obtaining the information of a plurality of operation conditions of a nuclear power system, the information of the operation condition comprises equipment involved in the operation condition, a connection relation between the equipment involved in the operation condition and operation parameters of the operation condition; constructing a three-dimensional simulation model corresponding to each operation condition according to the information of each operation condition in the information of the multiple operation conditions; and performing three-dimensional simulation on each operation condition through the three-dimensional simulation model corresponding to each operation condition. Due to the fact that the three-dimensional model can more visually display the operation conditions of the equipment under all the operation conditions, the simulation effect of the three-dimensional model is closer to the actual condition compared with a two-dimensional flow chart and is more visual, and then the analog simulation effect of the nuclear power system can be improved.

This invention discloses a rapid mesh mapping method for heat exchangers in nuclear power systems, comprising the following steps: 1. Modeling and meshing the internal and external fluid domains of the heat exchanger using SolidWorks software; 2. Obtaining the coordinates of all primary side meshes and calculating the maximum center-to-center distance between adjacent meshes in three directions; 3. Marking all secondary side meshes within the range centered on the primary side mesh coordinates and with the center-to-center distance equal to half the side length; 4. Assigning cyclic numbers to the primary and secondary side meshes and counting the corresponding primary and secondary side mesh counts; 5. Writing the primary and secondary side mesh marking process into the initialization macro built into Fluent software, using the predefined macro to read the mesh coordinate parameters, performing initialization operations, and then realizing the retrieval and marking of primary and secondary side meshes. This invention solves the problem of mesh mismatch when calculating the coupling between internal and external fluids in heat transfer tubes, and still provides reasonable marking results even when the internal and external fluid meshes are divided in different ways and when the number and arrangement of heat transfer tube meshes vary significantly.

This disclosure pertains to the field of nuclear power technology, specifically relating to an intelligent inspection system for nuclear power plants. The intelligent inspection system for nuclear power plants disclosed herein implements functions including system settings, basic data, inspection management, approval management, inspection tasks, inspection results, and data statistics. It addresses the daily inspection needs and challenges faced by production personnel in power plant operation, maintenance, and equipment management, optimizes equipment performance, fully utilizes inspection data, and improves production inspection efficiency through intelligent means. Furthermore, the system disclosed herein can integrate with third-party applications to achieve single sign-on and unified authentication, thereby breaking down information silos and securely and efficiently achieving integration and information sharing between the inspection system and multiple nuclear power systems.

A nuclear power system has an open volume between containment vessel and a reactor vessel containing a reactor core. Located in the open volume is a container holding a neutron-absorbing chemical in solid form. The container is configured to release the chemical in solid form directly into the open volume in response to a predetermined temperature and / or a predetermined pressure within the open volume. The released chemical can assist in maintaining the reactor core in a sub-critical state.

The invention relates to the technical field of nuclear power system energy management, and discloses a carbon dioxide energy storage device and method, and the device comprises a carbon dioxide capture unit, a compression liquefaction unit, a storage unit and an energy release power generation unit. Wherein the carbon dioxide capturing unit comprises at least two parallel absorption towers, and each absorption tower is provided with a cooling medium interface for connecting a nuclear power system cooling medium and a heating medium interface for connecting a nuclear waste heat medium; the compressing and liquefying unit is connected with the trapping unit, and the storage unit stores liquefied carbon dioxide; the energy-releasing power generation unit is connected with the storage unit and drives a power generator to generate power through liquid carbon dioxide cold energy and nuclear waste heat. According to the invention, the residual heat and cold source of the nuclear power system can be effectively utilized, continuous capture, liquefied storage and on-demand energy release of carbon dioxide are realized, and the peak clipping and valley filling requirements of small-sized or offshore nuclear power devices are met.

The application discloses a control method of a nuclear power steam atmosphere discharge system. The control method combines the pressure of three parallel main steam pipes, the pressure difference between the three main steam pipes and three steam generator water level parameters, preliminarily judges the fault type, and determines whether to trigger the safety injection action through a control model. The application avoids the phenomenon that the safety injection action is mistakenly triggered under the condition of the steam atmosphere discharge valve compressed air leakage fault, and improves the safety and stability of the nuclear power system.

The invention relates to the technical field of ocean nuclear power systems, in particular to a novel ship power system heat exchange device which comprises a shell, an air inlet is formed in the first end of the shell, and an air outlet is formed in the second end of the shell; the base body is contained in the middle section of the shell, and a plurality of tubular channels extending in the axial direction are formed in the base body; the circumferential outer surface of each heat pipe is coated with a sleeve, each heat pipe is arranged in the corresponding tubular channel, and each heat pipe is coaxial with the corresponding tubular channel; the outer diameter of each sleeve is smaller than the inner diameter of the corresponding tubular channel, so that an annular flow channel is defined between each sleeve and the inner wall of the corresponding tubular channel, and heat exchange working media can flow through the annular flow channels conveniently. Through the configuration, the heat exchange working medium flows through the circumferential outer surface of the sleeve in the axial direction, so that heat in the heat pipe pile is efficiently transferred to the high-pressure heat exchange working medium, and continuous, stable and efficient power output is provided for a ship.