268 results about "Thermal hydraulics" patented technology

The invention provides a three-dimensional thermal hydraulic analysis method and system for a reactor core. The method comprises the following steps: analyzing the type of the reactor core, dividing an outer-layer grid and a computational grid, and establishing a conservation mapping relation and a transfer equation set; the friction effect caused by the viscosity of the coolant is decomposed, and a source term is used to characterize the exchange of physical quantities caused by turbulence mixing; establishing a three-dimensional control equation set formed by mass, momentum and energy conservation equations for flow and heat transfer phenomena in the coolant channel, and forming a complete matrix equation set in combination with the transfer equation set; setting physical field boundary conditions and initial conditions of the reactor core, and setting an initial field; and carrying out iterative solution on the complete matrix equation set to obtain the thermal hydraulic parameters. According to the method, cross flow distribution and finer speed distribution in a coolant channel can be solved through thermal-hydraulic analysis, the prediction capacity of local thermal-hydraulic parameters is remarkably improved, and the method can be used for more accurate thermal-hydraulic analysis.

The invention discloses a reactor core nuclear thermal coupling calculation method based on an embedded physical knowledge neural network. The method comprises the following steps: 1, dividing reactor core nodes of a nuclear reactor; 2, initializing neutron physical field parameters and thermal hydraulic parameters of the reactor core; 3, building, training and storing an embedded physical knowledge neural network model; 4, predicting the neutron number density of each node of the reactor core by using the stored model to obtain neutron flux distribution; 5, calculating reactor core power distribution according to neutron flux distribution; 6, calculating and updating the thermal hydraulic parameters of the reactor core at the tn + 1 moment by using a system analysis program; 7, calculating and updating reactivity of each node of the reactor core; and 8, judging whether the appointed calculation time is reached, if so, ending, otherwise, returning to the step 4. The method is based on an embedded physical knowledge neural network method, an accurate and efficient calculation method is provided for solving distribution of the neutron physical field and the thermal field in the reactor core, and the method has important significance on reactor design optimization.

The invention relates to the technical field of thermal hydraulic power, in particular to a rapid prediction method for flow heat transfer characteristics of supercritical carbon dioxide of a micro-channel heat exchanger, and the method comprises the steps: dividing a training set and a test set according to a preset working condition parameter space, and generating a full-order flow field database under multiple working conditions through numerical simulation; performing eigenorthogonal decomposition on the full-order flow field database, and extracting a key mode of a dominant flow heat transfer feature to construct a reduced-order solution space; training an optimized neural network model by using the training set, and establishing a mapping relationship between the working condition parameters and the reduced-order solution space; and reconstructing flow field and temperature field information based on the mapping relation, and calculating and outputting a flow heat transfer characteristic prediction result. The method aims at solving the technical problem that the heat transfer characteristic prediction calculation efficiency of the supercritical carbon dioxide heat transfer flow of the micro-channel heat exchanger is low.

A pressurized water reactor oxidation corrosion product cross-scale nuclear-thermal-material coupling simulation method comprises the following steps: in the first stage, obtaining a supercooled boiling position through mutual loop iteration simulation of a neutronics model and a thermal hydraulic model; and in the second stage, multi-physics field coupling simulation is carried out through an oxidation corrosion product (CRUD) model, a neutronics model and a thermal hydraulic model, and high-resolution simulation under the pressurized water reactor nuclear-thermal-material coupling condition is obtained. According to the method, a scene in which the thickness range of the oxidation corrosion product is 20-200 microns is considered, multi-physical fields such as pressurized water reactor neutron physics, flow heat transfer and oxidation corrosion product deposition are calculated and analyzed, and the complex physical process can be captured more accurately.

The invention discloses a system for LOCA accident simulation irradiation test and a design method, which are characterized in that a pressurized water reactor fuel element LOCA irradiation performance test is taken as a research object, a circulation loop and a test device are established, prototype reactor operation conditions and different accident process thermal hydraulic conditions are simulated, a fuel element LOCA irradiation simulation test is carried out, and a test result is obtained. The high-temperature and high-pressure water system is divided into three areas according to temperature and pressure, the three-stage cooling circulation system is combined with the detachable plate heat exchanger, and the number of the heat exchangers and the parameter coupling complexity among the three-stage system are reduced. The mode of combining circulation loop electric heating and in-pile device electric heating is adopted, the irradiation test thermal condition is controlled, the test power of the device is increased, the device adopts a long heat exchange flow channel, the natural circulation capacity is constructed, and the safety of the test device is improved; the invention provides a test method aiming at release working conditions of trace and massive gases and fission products, and a simulated irradiation test covering various accident working condition ranges is provided, so that the application range is wide.

The invention relates to the technical field of reactor thermal hydraulics and safety, in particular to a long electric heating plate heat insulation structure and a heat transfer simulation experiment device with the long electric heating plate heat insulation structure. The long electric heating plate heat insulation structure comprises a heat insulation base, a heating plate and a cementing layer. The heat insulation base is configured to be a ceramic resin plate, and a sinking groove is formed in one plate face of the heat insulation base. The heated plate is arranged in the sinking groove so that the plate face of the heated plate can be flush with the plate face of the heat insulation base, and the surface, corresponding to the groove wall and the groove bottom of the sinking groove, of the heated plate is provided with a micropore aluminum oxide ceramic layer which is subjected to nanocrystallization treatment through an active coupling agent. And the cementing layer is arranged between the heated plate and the heat insulation base in a high-temperature vacuum pressure environment. According to the heat transfer simulation experiment device, it can be ensured that the heated plate has good single-side heat dissipation performance under the temperature working condition that the surface of the dispersion type fuel element is blistered, the experiment safety during operation of the heat transfer simulation experiment device can be ensured, and the accuracy of a measurement result in a simulation experiment can be ensured.

The application belongs to the technical field of thermal hydraulic analysis in nuclear reactors, and proposes an initial parameter generation method based on NUMAP steady-state calculation, which comprises: obtaining a thermal hydraulic simulation system model to be initialized in NUMAP, and decomposing the model into multiple single-flow single-pressure boundary chain structure sub-models; for each single-flow single-pressure boundary chain structure sub-model, setting the gas mass fraction of each in-line component and the non-condensable gas mass fraction to 0, and setting the flow to a constant value at the flow boundary; setting the temperature of each in-line component to the average of the inlet and outlet temperatures, or linearly filling the temperature of each in-line component according to the length of the component with the inlet and outlet temperatures as the endpoints; obtaining the total friction pressure drop along the line by subtracting the total gravity pressure drop along the line from the total pressure drop along the line, linearly dividing the total friction pressure drop along the line according to the parameters, and sequentially assigning the initial pressure of the in-line component from the inlet boundary. The application can make NUMAP reach steady state as soon as possible.

The invention discloses a multi-objective optimization method for performance of plate-shaped fuel. The method comprises the following steps: 1, carrying out model related setting in a fuel performance analysis program; 2, determining decision variables, optimization objectives and constraint conditions required by the multi-objective optimization method; 3, determining the number of individuals of each generation of population and the maximum number of iterations; 4, selecting a required multi-objective optimization algorithm, and performing population initialization operation; 5, introducing an elitist strategy, performing non-dominated sorting and distributing fitness; 6, calculating a non-dominated set optimization individual congestion degree and a constraint violation degree; 7, inputting the population individuals into a fuel performance analysis program, and respectively carrying out irradiation thermal hydraulic coupling calculation to obtain an optimization target value; 8, performing operations such as selection, crossover and variation on the current population individuals, and evolving to generate a filial generation population; 9, repeating the steps 5-8 until the algorithm judges that the optimization process is converged or the maximum number of iterations is reached; and 10, outputting an optimal solution set, namely design parameter information of p groups of dispersion particle diameters and pellet matrix thickness containing the optimized target value.

The invention discloses a pressure vessel lower chamber model order reduction method combining sparse grids and POD, and the method comprises the steps: adding one to the grade of the sparse grids after iteration initialization, and obtaining test grid points according to important grid points; executing the lower chamber full-order computational fluid mechanics model corresponding to the test grid points to obtain a first snapshot; reconstructing a second snapshot by using the mapping relation and the POD modal matrix; calculating a root-mean-square error of the two snapshots, and if the root-mean-square error is smaller than a threshold value, completing construction of a reduced-order model Otherwise, the grid points with the errors meeting the requirements are stored in a non-important grid point set, the grid points which do not meet the requirements are marked as important grid points, and the snapshot matrix is updated. Performing singular value decomposition on the snapshot matrix, updating the mapping relation, judging whether the maximum sparse grid grade is reached or not, and continuing iteration if the maximum sparse grid grade is not reached. According to the method, the field distribution of the thermal hydraulic parameters in the lower chamber can be quickly obtained through matrix algebraic operation, and the method can be used for application scenes such as parameterization analysis, design optimization and digital twinning.

The invention discloses a single-rod CHF test device and method, and belongs to the field of thermal hydraulic tests. The single-rod CHF test device comprises a heating rod, a thermal insulation pipe and a fixed grillwork, the heating rod coaxially penetrates through the thermal insulation pipe and is radially limited by the fixed grillwork, a cooling medium is filled between the thermal insulation pipe and the heating rod, the diameter d of the heating rod and the inner diameter D of the thermal insulation pipe meet Dh = De = (D2-d2) / d, Dh is the thermal diameter corresponding to a simulated fuel rod, and De is the thermal diameter corresponding to the simulated fuel rod. De is the hydraulic diameter corresponding to the simulated fuel rod. According to the device, the thermal hydraulic characteristics of the full-length rod bundle fuel can be accurately simulated through the single heating rod, the cost of a CHF test is effectively reduced, and the test efficiency is improved.

The invention discloses a nuclear reactor thermal hydraulic acceleration calculation method based on a modular parallel idea. The method comprises the following specific steps: 1, modeling each subsystem of a nuclear reactor system based on a sorted nuclear reactor subsystem list; 2, establishing a coupling boundary on the basis of modeling the subsystems respectively; 3, carrying out initial state parallel calculation according to a set initial boundary condition; 4, establishing a thermal hydraulic boundary information transmission strategy for the subsystem model; 5, advancing the time step, and using a calculation result in the step 4 as an initial thermal hydraulic parameter for calculation of the next time step; and 6, repeating the steps 4 and 5 until all-time parallel computing is completed. According to the method disclosed by the invention, aiming at the thermal hydraulic calculation process of the nuclear reactor, the thermal hydraulic calculation speed of the nuclear reactor system is improved while the thermal hydraulic parameters are accurately calculated.

The invention relates to the technical field of reactor thermal hydraulic power and safety, in particular to a heat transfer simulation experiment device under the surface foaming condition of a dispersion type fuel element. The device comprises a heated plate, a heat insulation base, a cover body, a medium leading-in piece, a medium leading-out piece and a conductive piece, the heated plate is provided with a vacuum bubble-shaped part; the heated plate is connected to the heat insulation base, an elastic sealing cavity is formed in the heat insulation base, the elastic sealing cavity is filled with heat insulation powder, and a temperature measuring hole is further formed in the heat insulation base; the cover body is provided with a circulation groove and connected with the heat insulation base. The medium leading-in piece is communicated with one end of the narrow slit channel; the medium leading-out piece is communicated with the other end of the narrow slit channel; the two conductive pieces are connected to the two ends of the heated plate respectively, and the vacuum bubble-shaped part is located between the two conductive pieces. According to the application, the structural state, thermal boundary and coolant flow channel geometric condition simulation of the dispersion type fuel element during bubbling can be realized, and the measurement of flow heat transfer characteristics, vapor-liquid two-phase flow field and the like near a vacuum bubble part can be carried out.

The invention provides a method and a device for inverting heat flow of a divertor based on a laser speckle image, and belongs to the technical field of nuclear fusion engineering thermal hydraulic testing. The device comprises a fusion reactor high heat flow simulation loading system, a laser speckle effect test system, a divertor water cooling simulation system and a test database system. A divertor target plate heat flow and loading surface strain corresponding database is established through the steps of adjusting cooling water flow conditions of a divertor test piece, carrying out series high heat flow loading on the divertor, measuring the divertor by adopting a laser speckle digital correlation technology and the like. In the actual reactor application, the surface strain of the same parts in the reactor is measured through the laser speckle technology, and the loaded heat flow inversion is realized through the corresponding query database. According to the method and the device for inverting the heat flow of the divertor based on the laser speckle image, quantitative inversion of the loaded heat flow of the divertor in an actual fusion reactor service environment is realized, and the operation safety of a nuclear fusion reaction device is improved.

The invention discloses a fuel rod burnout share analysis method and device and a consequence evaluation method and system under a shaft jamming accident, and the method comprises the steps: simulating a transient process of the shaft jamming accident of a nuclear power plant, and determining a transient working condition parameter corresponding to a time point of a minimum transient DNBR under the shaft jamming accident; under the condition that the minimum transient DNBR is smaller than a safety limit value, determining an initial thermal hydraulic state of a reactor core of the nuclear power plant according to the transient working condition parameters; according to the initial thermal hydraulic state and a pre-established reactor core channel model, adjusting a radial power factor of a hottest sub-channel of the sub-channel model, and determining a critical power peak factor; and according to the critical power peak factor, determining the burnout share of the fuel rod with DNB in the nuclear power plant shaft jamming accident from a fuel statistical curve. Therefore, the accuracy of analyzing the burnout share of the fuel rod under the shaft clamping accident can be improved.

The invention discloses a temperature regulation and control method and system for a main loop of a thermal hydraulic experiment system, and relates to the field of thermal hydraulic experiment research. The invention discloses a temperature regulation and control method for a main loop of a thermal hydraulic experiment system. The method comprises the steps of collecting historical data of the thermal hydraulic experiment system; training the to-be-trained valve opening prediction model to obtain a trained valve opening prediction model; the real-time experimental branch heating power, the cooler primary side inlet fluid flow, the cooler primary side inlet fluid temperature and the cooling water inlet fluid temperature serve as input parameters of the valve opening prediction model, and a valve opening feed-forward value is obtained through real-time prediction; a valve opening feedback value is obtained; and adding the valve opening feedforward value and the valve opening feedback value to obtain a valve opening value, and respectively controlling a cooling water flow regulating valve, a main loop cooler bypass flow regulating valve and a main loop cooler branch flow regulating valve to realize temperature regulation and control of the main loop of the thermal hydraulic experiment system. The problems that the main loop of the thermal hydraulic experiment system is difficult to maintain relatively stable in the experiment process and the flow fluctuation of the main loop is large when the temperature is regulated and controlled are solved.

The invention relates to a corrosion product deposition behavior numerical simulation analysis method and a related device, which are applied to a heat exchanger of a nuclear power plant residual heat removal system, and the method comprises the following steps: S1, establishing a three-dimensional geometric model corresponding to the heat exchanger based on a CFD platform; s2, performing grid division on the three-dimensional geometric model, setting boundary conditions and initial conditions of the three-dimensional geometric model, and establishing a thermal hydraulic model corresponding to the three-dimensional geometric model according to the grid division, the boundary conditions and the initial conditions; s3, calculating the thermal hydraulic model based on a FLUENT platform to obtain a temperature calculation result of a preset position of the heat exchanger; and S4, when the temperature calculation result meets the design value requirement, a corrosion product deposition and migration coupling model corresponding to the heat exchanger is established, and a sediment distribution result corresponding to the heat exchanger is calculated through the corrosion product deposition and migration coupling model based on the thermal hydraulic model. The method can predict the influence of corrosion product migration and deposition on the thermal performance of the heat exchanger.

The embodiment of the invention provides a method and equipment for measuring phase interface parameters between nuclear reactor fuel assemblies, and belongs to the technical field of thermal hydraulic analysis of nuclear power plants. The method comprises the steps that cavitation bubbles are generated in the flow channel of the special-shaped structure; images of cavitation bubbles in the flow channel of the special-shaped structure in multiple directions are collected; performing three-dimensional reconstruction according to the images of the cavitation bubble in the multiple directions to obtain the volume of the cavitation bubble; the void fraction is determined according to the volume of the void and the total volume of the fluid domain in the special-shaped structure flow channel, and the phase interface parameters comprise the volume of the void and the void fraction. According to the embodiment of the invention, the measurement precision of phase interface parameters can be improved.

The invention relates to the technical field of reactor thermal hydraulic power and safety, in particular to a dispersion type fuel element surface foaming simulation measurement system and method. The measuring system comprises a heated plate, a heat insulation base, a cover body, a medium leading-in piece, a medium leading-out piece, a conductive piece, first measuring equipment and second measuring equipment, the heated plate is provided with a vacuum bubble-shaped part; the heated plate is connected to the heat insulation base which is provided with a measuring window. The cover body is provided with a circulation groove and connected with the heat insulation base to form a narrow slit channel, and the light transmittance of the cover body is not smaller than 98%. The medium leading-in piece is communicated with one end of the narrow slit channel; the medium leading-out piece is communicated with the other end of the narrow slit channel; the two conductive pieces are connected to the two ends of the heated plate respectively; the first measuring equipment is positioned on one side of the heat-insulating base; the second measuring device is located on one side of the cover body. The thermal-hydraulic parameters of the heated plate and the medium can be conveniently measured based on the simulation system, and a reliable basis can be provided for actual thermal-hydraulic research of the dispersion type fuel element.

The invention provides a pressurized water reactor core axial power offset phenomenon prediction method and electronic equipment. The prediction method comprises the following steps: firstly, constructing a prediction model of the axial power offset phenomenon of the reactor core; wherein the prediction model is used for representing a model that the axial power offset phenomenon of the reactor core changes along with oxidation corrosion sediment CRUD and fuel burn-up on the surface of a cladding, and the prediction model comprises a neutron physical module, a thermal hydraulic module and a CRUD material module which have a coupling relationship; and then, carrying out coupling calculation on the prediction model within a preset duration, and generating prediction information of the axial power offset phenomenon of the reactor core. According to the prediction mode, the reactor core axial power offset phenomenon changing along with fuel burnup and CRUD deposition in the operation process of the pressurized water reactor is described, and therefore the prediction precision of the reactor core axial power offset phenomenon is improved.

The invention provides a nuclear power plant safety system based on causal reasoning, and relates to the technical field of nuclear power plant safety monitoring and fault diagnosis. The nuclear power plant safety system comprises a multi-source sensing layer and a causal engine layer, the multi-source sensing layer is used for collecting operation parameters of a nuclear power plant, and the causal engine layer comprises a dynamic causal network construction module and an online reasoning module. The dynamic causal network construction module is used for constructing a dynamic causal network according to the operation parameters, a continuous causal edge generated based on an equipment physical equation and a discrete event edge generated based on a probabilistic safety evaluation event tree; and the online reasoning module is used for determining a key fault propagation path according to the dynamic causal network so as to locate a fault source. According to the method, the equipment physical equation is embedded into the dynamic causal network, so that the reasoning process is ensured to accord with thermal hydraulic and safety physical rules, the problem of high false alarm rate caused by correlation misjudgment in deep learning is avoided, and the safety of the nuclear power plant is improved.

A method for determining an optimal sensor set includes identifying a set of possible faults of a thermal hydraulic system and a set of diagnostic objectives. The method also includes obtaining descriptions of sensor sets, which includes: receiving a first description of a first sensor set, and generating, based on the first description, a second description of a second sensor set. The method further includes, for each sensor set: determining a diagnostic capability of the sensor set, and calculating a score of the particular sensor set based on the particular description and the diagnostic capability, wherein a score of a sensor set is increased by a monetary cost of the sensor set and decreased by the sensor set meeting a diagnostic objective. The method also includes identifying the optimal sensor set that has a lowest score of the sensor sets, and displaying an indication of the optimal sensor set.

The invention discloses a transient analysis method and system for a multi-component working medium system of a gas cooled reactor, and relates to the technical field of safety numerical analysis of a nuclear reactor system.The transient analysis method comprises the steps that forming reasons of multi-component working media in the gas cooled reactor are analyzed, and data are collected and preprocessed; a first loop and a second loop of the system are constructed, and mathematical models for thermal hydraulic analysis are constructed respectively; and performing multi-component working medium transient analysis, and outputting the temperature of each control body. According to the method, the uniform mass conservation equation of the mixture is established for the multiple components of working media in the system, the conservation equation does not need to be independently listed for each component, the solving process is simplified, and the calculation efficiency is improved. Meanwhile, the method can be used for analyzing migration of each component working medium in the loop due to coolant flowing. Besides, in the transient analysis method of the system, the influence of the composition, concentration and thermal parameters of the multi-component working medium on the thermal hydraulic solution of the whole system in the generation and migration process is considered, so that the method accords with reality, and the accuracy of the method is guaranteed.

The invention discloses an industrial-grade printed circuit board heat exchanger design optimization method based on structured grid division and an artificial neural network, and the method comprises the steps: S1, carrying out the modeling of a heat exchanger based on a structured grid division method, and segmenting an end socket of the heat exchanger into a plurality of hexahedron structures, a flow channel of the heat exchanger is built into an agent flow channel model in the axial direction; s2, key parameters of a heat exchanger runner are extracted through a proxy runner model, and flowing and pressure drop characteristics of fluid in the runner are calculated; S3, based on the computational fluid mechanics simulation technology, a velocity field, a pressure field and flow distribution of the fluid in a heat exchanger end socket are obtained; s4, dividing control bodies for the heat exchanger based on a finite volume method, establishing a heat exchange amount and efficiency calculation program of the heat exchanger, and obtaining the efficiency of the heat exchanger; and S5, an artificial neural network training model is introduced, a multi-objective optimization algorithm is applied, the thermal hydraulic performance of the heat exchanger is optimized, a Pareto leading edge is generated, and multiple design selections are provided for designers to refer to.

The invention discloses a nuclear safety simulation verification method and device based on digital twinning and a medium, and relates to the technical field of nuclear engineering and digital twinning, and the method comprises the steps: training a physical information neural network based on a digital twinning initial model in combination with a nuclide transport equation and a thermal hydraulic equation, and outputting a nuclear safety multi-field simulation result; based on a nuclear safety multi-field simulation result, establishing a space-time diagram structure, extracting dynamic evolution characteristics by using a space-time diagram convolutional network, and outputting a nuclear accident propagation path and a risk distribution diagram; according to the method, the space-time diagram structure is constructed, the space-time diagram convolutional network is utilized to extract the dynamic evolution characteristics, the nuclear accident propagation process is modeled as cascade response in the space-time domain, and the reliability of the nuclear accident propagation process is improved. And the accuracy, the foresight and the intelligent level of nuclear security simulation verification are improved.

The invention provides a data assimilation-based reactor core thermal-hydraulic numerical solution method and system, and the method comprises the steps: building a control equation set according to a to-be-analyzed thermal-hydraulic problem, carrying out the sealing and discretization processing, and forming a complete linear control equation set; determining a solving area; establishing data transmission modes of different-scale thermal hydraulic methods; selecting a data assimilation algorithm and a CFD (computational fluid dynamics) algorithm according to the thermal hydraulic problem to be analyzed, and performing joint calculation; judging whether iteration meets a convergence requirement or not; if the CFD simulation data and / or experimental data in the current time layer exist, updating a prior covariance matrix of the temperature and the speed to obtain a posterior covariance matrix; and judging whether the numerical solution process is finished or not. According to the method, observation information of different sources and different resolutions can be fused, rapid thermal hydraulic calculation is carried out on the reactor core, the coolant flow and heat exchange model prediction capacity is improved, and the precision of the fine lattice cell sub-channel calculation method is improved.

The present application relates to a data correction method and apparatus for a nuclear reactor, a computer device and a storage medium. The method comprises: acquiring calibration reactor core data of a reactor of a nuclear power plant at a calibration moment, the calibration moment referring to the moment when a detector in the reactor is calibrated; acquiring real-time reactor core data obtained by using the detector to monitor the reactor; performing thermal hydraulic calculation on the reactor on the basis of the real-time reactor core data to obtain an estimated departure from nucleate boiling ratio of the reactor; and on the basis of a reactor core data deviation between the real-time reactor core data and the calibration reactor core data, correcting the estimated departure from nucleate boiling ratio to obtain a corrected departure from nucleate boiling ratio, wherein the corrected departure from nucleate boiling ratio is used for protection and control of the reactor. The method can improve the accuracy of reactor protection and control.

The invention discloses a thermal hydraulic numerical analysis method and equipment for a nitrogen pressure stabilization reactor, and relates to a combined analysis method of a one-dimensional system level thermal hydraulic analysis program and an equipment internal three-dimensional structure level CFD (computational fluid dynamics) program. Phenomena of nitrogen dissolution, migration, precipitation, accumulation and the like in the nitrogen pressure-stabilized reactor and thermal hydraulic effects caused by the phenomena are subjected to whole-process prediction, and the influence of the phenomena on loop flow heat transfer characteristics and key equipment working performance is further analyzed. According to the method, multi-dimensional and multi-scale computational analysis can be realized, on the basis of not consuming a large amount of computing power resources, the dissolution, migration, precipitation and accumulation behaviors of nitrogen in the coolant and the thermal hydraulic effect caused by the dissolution, migration, precipitation and accumulation behaviors are predicted, and check of a nitrogen pressure-stabilized reactor design scheme and formulation of an operation strategy are supported.

The invention discloses a subchannel division thought-based thermal hydraulic calculation method for a direct-current heat exchanger, and belongs to the technical field of fluid dynamic simulation and analysis of heat exchangers. The flowing area of the once-through heat exchanger is divided into a shell side area and a pipe side area according to different fluid characteristics of the primary side and the secondary side of the once-through heat exchanger. Dividing the shell side flow area into a plurality of adjacent sub-channels by taking the adjacent pipe walls as boundaries; the method is based on specific basic geometric parameters of a shell side sub-channel, including a sub-channel circulation area, an equivalent diameter, a wet perimeter, a heating perimeter and a gap width; establishing a heat exchanger tube side single-channel flow heat exchange basic equation; establishing a flow heat exchange basic equation in the heat exchanger shell side sub-channel; establishing a pipe wall heat conduction equation; and setting boundary conditions, and performing coupling calculation to obtain thermal parameters in sub-channels on the pipe side and the shell side of the heat exchanger. The method is used for solving the problems that one-dimensional calculation is inaccurate, and three-dimensional calculation is long in consumed time and large in calculation amount.

The invention discloses a thermal hydraulic flow field prediction method based on an evidence physical information neural network, and the method comprises the specific steps: collecting different physical quantity data sets in a thermal hydraulic system, and carrying out the preprocessing of the data; constructing data loss; constructing physical loss; constructing an evidence uncertainty modeling mechanism, and dynamically allocating evidence weights according to the reliability of different data and physical constraints; constructing a total loss function by integrating the data error, the physical residual error and the evidence reliability; and outputting the trained neural network model and prediction. According to the invention, by introducing an evidence learning mechanism, the uncertainty of the model is dynamically estimated in the network training process, and the adaptability of the model to data with inconsistent noise sensitivity degrees is significantly improved; the uncertainty of the evidence is mapped into a physical quantity weighting coefficient, a uniform normalized weighting loss function is constructed, and the subjectivity and robustness problems caused by manual setting of hyper-parameters for each physical quantity in a traditional physical information neural network are effectively avoided.

The invention relates to the field of fuel element thermal hydraulic experiments, and discloses an experimental system for testing a petal-shaped fuel element, the experimental system comprises a petal-shaped fuel element, an element container, a liquid supply unit, an electric heating device and a temperature detection device, a spiral arc-shaped convex surface and a spiral arc-shaped concave surface are formed on the peripheral wall of the petal-shaped fuel element; the element container is used for accommodating petal-shaped fuel elements; the liquid supply unit is used for conveying test liquid to the element container; the electric heating device is electrically connected with the petal-shaped fuel element and is used for heating the petal-shaped fuel element, so that the test liquid can reach a boiling critical point; the temperature detection device comprises a first temperature measurement element and a second temperature measurement element, the first temperature measurement element is used for detecting the convex surface temperature of the arc-shaped convex surface, and the second temperature measurement element is used for detecting the concave surface temperature of the arc-shaped concave surface. The experiment system can accurately detect the critical heat flux density of the petal-shaped fuel element, and can accurately obtain the circumferential position of the boiling critical point at the same time.