35 results about "Nuclear power system" patented technology

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 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 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 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.

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.

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.

This disclosure provides a framework that integrates a simulation model and algorithm for determining whether system requirements are met across multiparameter scenarios in a complex system. The complex system may include, for example, a life support system, thermal control system, nuclear power system, control system for vehicle dynamics, telecommunications system, finance system, chemical refining system, or propulsion system. A graphical display such as a scenario result matrix may visually represent compliance / non-compliance for each criterion for each multiparameter scenario. In some implementations, perturbations such as noise or load deviations may be introduced to special scenarios, and stability assessments may be verified for each of the special scenarios in the complex system.

The invention belongs to the technical field of nuclear power simulation, and particularly relates to a thermal hydraulic system simulation model generation method based on AIGC. The method utilizes a large language model to construct an intelligent agent, realizes the requirement of automatically generating a modeling code according to a user instruction, and comprises the following steps: step 1, generating a data set; step 2, fine tuning of the language model; step 3, agent construction: two agents are constructed, a task decomposition agent is responsible for decomposing a modeling task, and a modeling agent is responsible for generating a specific modeling code; and 4, performing model deployment integration. The method has the beneficial effects that on the basis of the method, the reasoning ability of a large language model can be utilized, and automatic construction from simulation system design requirements to corresponding modeling codes is achieved. Meanwhile, the generation speed of the nuclear power system simulation model can be increased, and the accuracy of conversion from design information to the simulation model can be improved.

This application relates to the field of fault prediction and health management technology, specifically to a health assessment method for intelligent nuclear power systems based on large model technology. The method includes: extracting knowledge from historical nuclear power data to obtain triplet data; filtering similar triplet data from the remaining triplet data and constructing an undirected graph of the triplet data; obtaining path association parameters from entities in the triplet data to each node in the undirected graph; obtaining pre-filtered triplet data for each triplet data; obtaining the association redundancy of the pre-filtered triplet data; and then filtering out redundant triplet data from all triplet data extracted from the knowledge, constructing a nuclear power knowledge graph, and assessing the health status of the nuclear power system. This application aims to make the nuclear power knowledge graph more accurate and concise by filtering out redundant parts of the triplet data, thereby achieving a more precise health assessment of the nuclear power system.

The invention provides a design margin determination method and device and electronic equipment, which are applied to the technical field of nuclear power, and the method comprises the steps: determining a plurality of groups of first parameter values according to a plurality of physical parameter values of a nuclear power system, each group of first parameter values comprising an originating event occurrence frequency value and a processing measure failure probability value, the originating event is an event causing the change of the physical parameter value, and the processing measure is a measure taken to deal with the originating event; determining a plurality of radioactive substance release frequency values of the nuclear power system according to the plurality of groups of first parameter values; determining a target physical parameter value in the plurality of physical parameter values according to the plurality of radioactive substance release frequency values; and determining a design margin of the nuclear power system according to the target physical parameter value and a physical parameter design reference value of the nuclear power system. The method improves the reliability of the design margin.

The application provides a compensation loop access method, a control device and a voltage stabilizing system. The method comprises the following steps: obtaining the conduction position of a controllable switch in a current output voltage period; obtaining the output voltage frequency of a voltage stabilizer in real time, and determining a real-time counting interval according to the output voltage frequency and a preset counting number; determining the corresponding first counting value of the controllable switch in the current output voltage period based on the counting number and the conduction position; starting counting at the real-time counting interval when the current output voltage period starts, and controlling the controllable switch to conduct when the first counting value is counted. In the application, the actual access process of the compensation loop can be adaptively adjusted according to the output voltage frequency, the generation of the excitation current in the voltage stabilizing process is reduced, and the normal operation of the voltage stabilizer is ensured. On this basis, the application can be effectively applied to the nuclear power field, the stable operation of the electrical equipment for nuclear power is ensured through effective voltage stabilization, and the stable operation of the nuclear power system is ensured.

This invention belongs to the field of thermal-hydraulic simulation technology, specifically involving a joint simulation method based on Modelica for two-fluid six-equation and differential-algebraic equations. It includes the following steps: Step 1: Using the FMU model framework, the two-fluid six-equation system is converted into a function model TpfFun containing input, output, and parameter variables; Step 2: Under Modelica-based compilation rules, the function model TpfFun and the differential-algebraic equation system are compiled as a whole to obtain a sequence plan for the unified solution of the heterogeneous system of two-fluid six-equation and differential-algebraic equations; Step 3: According to the solution sequence plan, each subset of equations in the heterogeneous system of two-fluid six-equation and differential-algebraic equations is solved sequentially. The beneficial effects of this invention are: it solves the problem of coupled solution of two-fluid six-equation and differential-algebraic equations under the unified modeling framework based on Modelica, ensuring the stable solution of the multi-disciplinary coupled model of the entire nuclear power system, and guaranteeing the convergence and efficiency of equation solving.

This disclosure belongs to the field of nuclear power technology, specifically relating to a method and apparatus for creating, monitoring, and analyzing industrial control system (ICS) honeypots. The ICS honeypot creation and monitoring analysis method provided by this disclosure offers a new approach to proactive defense. Through honeypot system management software, it manages virtual machine replication, network configuration, system updates, backups, and recovery. By configuring network IPs and services, it quickly creates ICS honeypots, allowing intruders to gain full access to the system and use them as springboards to launch further network attacks, achieving a realistic simulation effect. This allows for detailed recording of intruder attack behavior. It reduces the likelihood of the real business network of the nuclear power system being detected, increases the cost, difficulty, and timeliness of attacker attacks, and effectively complements the passive protection measures existing in the nuclear power system, achieving the goal of proactive defense.

The invention discloses a waste heat removal device and method and an underwater nuclear power system, and relates to the technical field of nuclear safety. The device comprises radiation heat exchangers arranged in a shell of the underwater nuclear power device, and a liquid inlet of each radiation heat exchanger can be communicated with a liquid outlet of a reactor; a liquid outlet of each radiation heat exchanger can be communicated with a liquid inlet of the reactor; the underwater nuclear power system comprises an underwater nuclear power device and the waste heat removal device, the underwater nuclear power device comprises a shell, a reactor, a generator and a first circulating pump, the generator is provided with a heat transfer pipeline, a liquid outlet of the reactor is communicated with a first liquid inlet of the heat transfer pipeline, and a liquid inlet of the reactor is communicated with a first liquid outlet of the heat transfer pipeline. The first circulating pump is arranged between the reactor and the generator; the method uses the residual heat removal device and comprises the steps that the underwater nuclear power device stops running; cooling liquid in the reactor flows to the radiation heat exchanger and then flows to the reactor, and circularly flows; the residual heat is completely discharged; the structure is simple, and the size is smaller.

The invention discloses a nuclear power device fault diagnosis method and system based on CNN and sliding window verification, and belongs to the technical field of nuclear power system fault diagnosis. The method comprises the following steps: acquiring sensor and actuator signals under normal and fault working conditions, and converting the sensor and actuator signals into three-channel color pictures as a training set through discrete sampling and normalization processing; a multi-layer CNN network including an input layer, a 2D convolutional layer and the like is constructed, and a fault diagnosis model is obtained through training; and collecting real-time signals to generate a test set, inputting the test set into the model in batches through a sliding time window, calculating label scores in the window, and selecting the label with the highest score as a fault type. The problems of high cost, low precision and poor real-time performance in the prior art are solved, the accuracy and the real-time performance of fault diagnosis are improved, and the method is suitable for safe operation monitoring of a nuclear power device.

The invention relates to a multi-inlet parallel condenser simulation method considering condensate supercooling and bubbling heating, which comprises the following steps: establishing a mother pipe node, each module pipe side node, a condensate confluence well node, a mass transfer boundary and a heat transfer boundary, and constructing a pipe network model by adopting streamline connection; modeling the shell side, heat exchange, air and condensate falling water of the condenser module to obtain a parallel condenser model; heat exchange of a transverse tube bundle, heat conduction of a heat transfer tube and forced convection heat exchange are calculated, heat exchange of an intermediate heat exchanger is modeled, and a heat exchanger model is obtained; coupling is conducted according to the pipe network model, the parallel condenser model and the heat exchanger model, initialization is conducted, and a stable initial calculation model is obtained; and acquiring parameters of an external boundary, inputting the parameters into the stable initial calculation model, after calculation is finished, exchanging data through an internal interface, advancing a time step length, and outputting the data to the external boundary. According to the method, the system-level evaluation and analysis requirements of the ship nuclear power system can be accurately and stably met.

This invention relates to the field of marine nuclear power system technology, specifically providing an interface heat exchange device for a ship power system. The device includes a base with several axially extending tubular channels; several heat pipes, each with its circumferential outer surface covered by a sleeve; each heat pipe is arranged within a corresponding tubular channel and extends parallel to that channel; the outer diameter of each sleeve is smaller than the inner diameter of the corresponding tubular channel, thereby defining an annular flow channel between each sleeve and the inner wall of the corresponding tubular channel; each sleeve has several radially outwardly extending ribs, each rib having a rib height and a rib width, and the rib height / rib width ratio is designed to be positively correlated with the temperature rise and pressure drop of the heat exchange medium passing through the annular flow channel, respectively. Through this configuration, the heat exchange surface area between the sleeve and the heat exchange medium can be directly expanded, and a balance can be achieved between the temperature rise and pressure drop of the heat exchange medium.

The application discloses a kind of nuclear power system CFD and system program coupling method, steps are as follows:1. for steam generator U-shaped section heat transfer tube in reactor respectively establish three-dimensional geometric model belonging to CFD program OpenFOAM and one-dimensional simplified model belonging to system program RELAP5;2, using RELAP5 program input card sets initial boundary condition and initial state of calculation domain of geometric model;3, RELAP5 program is calculated to the responsible area model, obtains calculation result, realizes data conversion using intermediate program and passes calculation result to OpenFOAM, as the boundary condition of OpenFOAM this time step;4, OpenFOAM is calculated to the responsible area model, obtains calculation result, realizes data conversion using intermediate program and passes calculation result to RELAP5 program, as the boundary condition when RELAP5 program next time step calculation;5, repeat step 3, step 4, until the region to be solved completes full time period calculation.Complete to the region to be solved multi-dimension, high-fidelity coupling analysis.

The invention discloses a silicon carbide micro-channel heat exchanger based on a photocuring printing process, and belongs to the field of ceramic materials and thermal management devices. According to the heat exchanger, silicon carbide powder, photosensitive resin, a dispersing agent and a photoinitiator are mixed according to a specific proportion to prepare slurry, and the slurry is subjected to DLP photocuring printing forming, degreasing carbonization, high-temperature sintering and other processes to prepare a silicon carbide matrix internally containing an inlet section, a heat exchange section and an outlet section communicating microchannel. The preparation method optimizes the slurry formula and printing, degreasing and sintering parameters, solves the problems of difficult preparation of complex microchannels, high cost and low efficiency in the traditional process, has the characteristics of high thermal conductivity, high temperature resistance and corrosion resistance, high production efficiency and excellent performance, and is suitable for extreme environment thermal management scenes such as aerospace and nuclear energy systems.