176 results about "Thermal hydraulics" patented technology

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 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 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 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 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.

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 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 belongs to the technical field of reactor thermal hydraulic tests, and particularly relates to a device for measuring steam-water direct contact condensation. The device comprises an upper cylinder body, a lower cylinder body, an upper sealing head and a lower sealing head, the upper sealing head, the upper cylinder body, the lower cylinder body and the lower sealing head are connected from top to bottom to form a sealed cylinder body structure; a visual mirror is arranged on the lower cylinder body, and the axial height of the visual mirror covering the lower cylinder body is more than 1 / 2; an electrode wire measuring structure is arranged at the joint of the upper cylinder and the lower cylinder and is used for capturing the vapor-liquid interface contact area; a through hole is formed in the upper sealing head, a steam outlet of the steam inlet connecting pipe is connected into the barrel structure through the through hole, and the steam outlet is connected with a steam diffuser. The vapor-liquid interface contact area can be obtained directly through visual shooting and later graphic processing, the vapor-liquid effective contact area can be quantitatively measured through the electrode wire, and therefore the vapor-water direct contact condensation heat transmission intensity is quantified.

The application discloses a simulation processing method for CCFL phenomenon of a downcomer based on RELAP5, and specifically comprises the following steps: S1, obtaining geometric parameters and operation condition parameters of the downcomer, making a node division scheme, and obtaining node information, wherein the node information comprises a number of circumferential nodes, a number of axial nodes, and a number of single pipe nodes between multiple pipes; S2, writing a RELAP5 program on an Input card, performing a thermal hydraulic calculation, and obtaining a first parameter which changes in real time; S3, extracting a second parameter through a strip card to obtain a third parameter, wherein the third parameter comprises gas-liquid phase flow, pressure and velocity; S4, importing the third parameter into a processing software MATLAB, reading a fourth parameter of the gas-liquid phase according to the node information set in S1; and S5, performing two-dimensional cloud chart drawing on the fourth parameter in S4. The application can quickly obtain required results under different node division schemes and different geometric parameters and operation parameters, and improve analysis efficiency.

The application provides a deep learning method for predicting thermal hydraulic characteristics of a sodium-cooled fast reactor heat exchanger, and belongs to the technical field of a sodium-cooled fast reactor passive residual heat removal system; the method comprises the following steps: a first DNN is used to approximate low-fidelity flow and temperature distribution; a second DNN is used to build a nonlinear relationship between low-fidelity and high-fidelity data; and a physical information partial differential equation and a boundary condition are used as constraints to inject a third DNN, so as to optimize a prediction result; by combining the advantages of a fluid heat transfer partial differential equation and data driving, the method can realize high-fidelity prediction of flow velocity and temperature in a sodium-cooled fast reactor PCHE by using only low-fidelity data, solves the problem that physical information in a heat exchanger is difficult to measure, and has important significance for improving the inherent safety of a sodium-cooled fast reactor and the safety of a reactor core.

The application relates to the technical field of reactor thermal hydraulic and safety, in particular to a dispersion type fuel element surface bubbling simulation measurement system and method. The measurement system comprises a heated plate, an adiabatic base, a cover body, a medium inlet piece, a medium outlet piece, a conductive piece, a first measurement device and a second measurement device; the heated plate is provided with a vacuum bubble part; the heated plate is connected to the adiabatic base, and the adiabatic base is provided with a measurement window; the cover body is provided with a flow-through groove and is connected to the adiabatic base to form a narrow gap channel, the light transmittance of the cover body is not less than 98%; the medium inlet piece is communicated with one end of the narrow gap channel; the medium outlet piece is communicated with the other end of the narrow gap channel; the two conductive pieces are connected to the two ends of the heated plate respectively; the first measurement device is located on one side of the adiabatic base; and the second measurement device is located on one side of the cover body. The application can conveniently measure the thermal hydraulic parameters of the heated plate and the medium based on the simulation system, and can provide a reliable basis for the actual dispersion type fuel element thermal hydraulic research.

The present application relates to the technical field of nuclear reactor thermal hydraulic simulation analysis, and provides a thermal stratification phenomenon simulation method and apparatus, a device, a medium, and a program product. The method comprises: on the basis of an interactive model computation result in a pre-configured program interface, controlling a one-dimensional system program and a three-dimensional computational fluid mechanics program to perform coupling computation, wherein the program interface is configured between the one-dimensional system program and the three-dimensional computational fluid mechanics program, and the model computation result comprises results obtained by performing simulation computation on a liquid metal reactor by the one-dimensional system program and the three-dimensional computational fluid mechanics program; and obtaining a real-time computation result of coupling computation of the one-dimensional system program and the three-dimensional computational fluid mechanics program, wherein the real-time computation result comprises a simulation result of a thermal stratification phenomenon of the liquid metal reactor in a transient operation process. The present application can reduce the computing resource demand and the time cost and improve the efficiency while ensuring that the thermal stratification phenomenon of the liquid metal reactor is accurately simulated.

The invention provides a multifunctional flow heat exchange visualization experiment device and method, and belongs to the technical field of thermotechnical waterpower, the device comprises an electric heater, and an outer layer annular space flow channel and an inner layer narrow annular space flow channel are respectively formed among the electric heater, an inner layer quartz glass sleeve and an outer layer quartz glass container; two pairs of thermocouples are symmetrically arranged on each of three cross sections of the electric heater and used for measuring the wall surface temperature, a pair of thermocouples is mounted on each of three cross sections of the inner-layer quartz glass sleeve, and a pair of thermocouples is mounted at each of an inlet of the outer-layer annular space test section, an outlet of the outer-layer annular space test section and an outlet of the inner-layer annular space test section and used for monitoring the water temperature. Visual observation and parameter measurement of different flow states and heat exchange phenomena are achieved, meanwhile, a temperature control and feedback system is adopted to enable temperature parameters in a device system to reach set design, it is ensured that the device cannot be damaged due to heat transfer deterioration near a critical point, and automatic thermal protection is achieved.

The invention discloses a twisted pressure suppression spray pipe capable of adaptively adjusting the immersion depth, relates to the technical field of reactor thermal hydraulic power, and solves the problems of condensate water hammer and thermal stratification caused by water seal removal, pool swelling and low-speed discharge in the whole operation stage of a pressure suppression system in the prior art. The suppression spray pipe comprises a twisted inner pipe and a circular outer pipe, the twisted inner pipe is arranged at the bottom of the pressure suppression box, one end is communicated with the containment, and the other end penetrates through the water space in the pressure suppression box and is arranged above the liquid level of the water space; the upper end of the circular outer pipe is closed, and the lower end movably sleeves the twisted inner pipe; an air space in the pressure suppression box is communicated with the containment through a vacuum release valve; the twisted inner pipe is a spiral straight pipe. The pressure suppression spray pipe adopts a passive design, the immersion depth of the spray pipe is automatically adjusted by virtue of the spring, the gravity of the outer pipe and the pressure difference between the containment and the pressure suppression box, the pressure suppression spray pipe is simple and reliable, and the problems of condensate water hammer and thermal stratification are solved by virtue of a plurality of through holes formed in the limiting porous ring.

The application discloses a kind of quick calculation method for determining the air volume configuration of oil-immersed transformer cooling system, the method considers that transformer develops to "lightweight, miniaturization" direction, cooling system intensification is a kind of inevitable trend, accurately configure cooling fan diameter can both ensure efficient heat dissipation, and can avoid problems such as high cost, heavy quality and large wind loss, and is in harmony with the concept of "light" and "small". The method comprises establishing a thermal hydraulic model considering the correlation of air volume change, introducing air leakage parameters and performing conjugate heat transfer calculation between the oil domain and the air domain, obtaining characteristic parameters reflecting the dynamic behavior of the heat exchanger coupled fluid. This method is to quickly obtain the thermal characteristic factors and wind loss of the heat exchanger under oil-immersed air cooling mode during the design and development stage to ensure reasonable and efficient heat dissipation of the transformer cooling system.

The invention belongs to the technical field of nuclear reactor engineering tests, and particularly relates to a rod bundle fuel assembly testing device. The device comprises a pressure channel, a simulation rod bundle arranged in the pressure channel, and a neutron imaging system arranged outside the pressure channel. The cross section of the pressure channel is in a regular hexagon shape, a lengthened pipe assembly is designed on the upper portion of the pressure channel, the simulation rod bundle is composed of a plurality of electric heating rods, and an outlet pipeline and an inlet pipeline are arranged at the upper end and the lower end of the pressure channel respectively and used for being connected with an external liquid metal loop. According to the invention, high-heat-flux heating of the reactor core can be simulated under the condition of liquid metal cooling, so that the thermal hydraulic phenomenon in the reactor core channel can be accurately simulated.

The embodiment of the application discloses a kind of thermal test device and test method, thermal test device includes metal body, metal body includes two metal plates distributed in working medium passage two sides. A plurality of electrically conductive ends are arranged at each metal plate, and the plurality of electrically conductive ends include at least two electrically conductive initial ends and electrically conductive terminal, in at least two electrically conductive initial ends, select any one electrically conductive initial end as test initial end, and test initial end and electrically conductive terminal are respectively connected to the two poles of power supply. The thermal test device disclosed in the application, by selecting any one of at least two electrically conductive initial ends as test initial end, the heating length of the metal plate located on both sides of the working medium passage can be controlled, thereby the heating conditions of symmetrical heating and asymmetrical heating can be simulated, so that the flow and heat exchange law inside the working medium passage under non-uniform thermal load can be tested, and the thermal hydraulic parameters of the working medium passage under a specific electric power distribution can also be obtained, and the test efficiency of the heat exchange equipment can also be improved.

The utility model relates to the technical field of nuclear reactor thermal hydraulics, and particularly discloses a multi-rod-bundle critical heat flux density experimental device, which comprises a shell and heating rods, the shell comprises a first electrode cylinder, a water inlet section, an experimental section, a water outlet section and a second electrode cylinder which are sequentially arranged from left to right, and all the sections are hermetically connected through flanges; shells of the water inlet section, the experiment section and the water outlet section comprise inner shells and pressure-bearing shells, flow channels are formed in the inner spaces of the inner shells, and interlayers are formed between the inner shells and the pressure-bearing shells; the multiple heating rods are located in the flow channel, and heating rod sealing discs used for supporting and limiting the heating rods are arranged in the water inlet section and the water outlet section correspondingly. The first electrode cylinder and the second electrode cylinder are provided with conductive columns used for heating the heating rod in the axial direction of the shell, and the conductive columns extend out of the shell. The two sides of the experiment section along the axial direction of the shell are provided with outlet pipes which are respectively connected with a temperature sensor and a pressure transmitter. The utility model aims to solve the problem that the existing critical heat flux density experimental device is not suitable for experimental research in a multi-rod-bundle high-pressure environment.

The invention belongs to the technical field of reactor thermal hydraulics, and particularly relates to a test system for simulating passive safety injection pipeline fracture. The system comprises a full-pressure water replenishing tank simulation body and a crevasse pipeline, the top of the full-pressure water replenishing tank simulator is connected with the pressure vessel simulator cold pipe section through a balance pipeline, and the bottom of the full-pressure water replenishing tank simulator is communicated with the pressure vessel simulator descending ring cavity through a DVI header pipe; a first isolating valve is arranged on the balance pipeline; the two ends of the crevasse pipeline are connected to the two ends of the first isolation valve through the second isolation valve and the third isolation valve correspondingly. A fourth isolating valve is arranged on the DVI header pipe; the two ends of the crevasse pipeline are connected to the two ends of the fourth isolation valve through a fifth isolation valve and a sixth isolation valve correspondingly. Wherein the crevasse pipeline is connected in series with at least one crevasse simulation piece. The embodiment of the invention can be used for carrying out the response characteristic test of the passive safety injection system under the conditions of small crevasses and double-end breakage in a balance pipe and a DVI pipe.

The present application relates to the technical field of nuclear reactor thermal hydraulic, in particular, to a rectangular fuel assembly coupling heat transfer coupling method and system, medium. The method or system provided by the present application mainly includes obtaining the surface heat flux density distribution of the fuel plate; performing the iteration calculation of the coolant flow distribution characteristics and the fuel plate heat conduction characteristics; judging whether the iteration calculation of the fuel plate heat conduction characteristics is completed. Through the above method, for the refined analysis and calculation requirement of the rectangular fuel assembly flow heat transfer, a detailed coupling heat transfer calculation method and process are proposed, which can be used for accurate solution of the above coupling heat transfer problem. The present application can be used as the algorithm basis for the development of the core thermal hydraulic program of the advanced research reactor (such as CARR). The calculation accuracy of the fuel assembly can be provided, so as to release the thermal design margin. It plays an important role in improving the overall performance of the research reactor.

The present invention discloses a closed gas circulation flow heat exchange experimental device and method to solve the problem that the existing technology cannot provide reliable thermal hydraulic experimental data for the design of gas direct cooling reactor system. Specifically, it includes an experimental component, a preheating component, a boosting component, a high-pressure gas source group and a vacuum pump; the experimental component includes an air inlet cavity and an air outlet cavity sleeved therein; the air outlet cavity is provided with a plurality of air outlet channels connected to the interior of the air inlet cavity and the air outlet end of the air outlet cavity, and a heating rod is provided therein; a first thermocouple and a first pressure gauge are provided on the air inlet cavity; the preheating component includes a preheating gas tank and a preheating rod provided therein; the air outlet end of the boosting component is connected to the air inlet end of the preheating gas tank; the air outlet end of the preheating gas tank is connected to the air inlet end of the air inlet cavity; the air outlet end of the air outlet cavity is connected to the air inlet end of the boosting component, and a second thermocouple and a second pressure gauge are provided on the connecting pipeline; the high-pressure gas source group is used to fill the experimental gas; and the vacuum pump is used to evacuate the air.

The invention discloses a reactor core physical-thermal coupling calculation method and device and a safety evaluation method and system.The method comprises the steps that a mapping relation is determined based on reactor core structure parameters, a first arrangement rule and a second arrangement rule of a micro reactor; a function interface is constructed based on the mapping relation, the function interface is used for input and output data exchange between the reactor core neutronics model and the reactor core thermal hydraulic model, the thermal hydraulic model is compiled into a static link library in advance, and the static link library and the neutronics model are linked and integrated in the same executable program; performing data conversion and iterative coupling processing between the reactor core neutronics model and the reactor core thermal hydraulic model through a function interface to obtain a temperature distribution-power distribution output pair corresponding to each round of iteration; and determining the temperature distribution-power distribution output pair obtained by the last iteration as the target power distribution and the target temperature distribution of the reactor core. And the accuracy and efficiency of physical-thermal coupling calculation of the reactor core are improved through internal coupling.

The application relates to the technical field of reactor thermal hydraulic and safety, in particular to a long electric heating plate heat insulation structure and a heat transfer simulation experiment device with the same. The long electric heating plate heat insulation structure comprises a heat insulation base, a heating plate and a glue layer; the heat insulation base is configured as a ceramicized resin plate, one plate surface of the heat insulation base is provided with a sink; the heating plate is arranged in the sink so that the plate surface of the heating plate is flush with the plate surface of the heat insulation base, the surface of the heating plate corresponding to the sink wall and the sink bottom is provided with a microporous alumina ceramic layer which is nano-treated by using an active coupling agent; and the glue layer is arranged between the heating plate and the heat insulation base under a high-temperature vacuum pressure environment. The application can ensure that the heating plate has good single-side heat dissipation performance under the temperature condition of bubble formation on the surface of a dispersed fuel element, can guarantee the test safety of the heat transfer simulation experiment device during operation, and can ensure the accuracy of the measurement result in the simulation test.