88 results about "Thermal hydraulics" patented technology

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 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 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 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 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 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 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.

The present application relates to the technical fields of high temperature gas cooled reactor, and discloses a kind of high temperature gas cooled reactor prism core thermal hydraulic calculation method, comprising: the core is discretized into calculation unit, and the core physical structure is mapped into calculation component;Build series heat resistance network model, obtain the total heat transfer coefficient between fuel component and graphite matrix component in the same calculation unit, and then obtain real-time heat flux density;Obtain the axial heat transfer rate between the calculation components directly contacted between adjacent axial calculation units;Obtain the radial heat transfer rate between the calculation components directly contacted between adjacent radial calculation units;Obtain the convective heat transfer coefficient and convective heat transfer rate between the surface of graphite matrix component and reflection layer component and the fluid flowing through the coolant channel, obtain the net radiation heat transfer rate;Obtain the temperature of each component, the outlet temperature of the core, the pressure difference between adjacent calculation units along the core and the average reactor pressure drop by the above results.The present application accurately simulates the thermal hydraulic behavior of PMR core.

The invention relates to the technical field of nuclear power plant steam generator thermal hydraulic analysis, in particular to a nuclear power plant steam generator tube bundle area thermal hydraulic simulation system, method and equipment. The system comprises a pre-processing module, an immersion boundary solving module and a flow thermal coupling module which are in communication connection, the pre-processing module is used for determining a geometric model and a computational domain of a heat transfer tube and a supporting structure of the steam generator; the immersion boundary solving module is used for solving a fluid control equation based on the geometric field and the non-structural background grid, and realizing preset boundary conditions by introducing virtual force source items at the boundary of the heat transfer tube and the supporting structure to obtain flow field distribution; and the flow thermal coupling module is used for receiving the flow field distribution and simulating the heat exchange process between the wall surface of the heat transfer pipe and the secondary side fluid. The technical problems that in an existing steam generator tube bundle area flow field simulation technology, microcosmic flow field analysis is insufficient, supporting structure resistance characterization is distorted, and calculation efficiency and precision are difficult to consider at the same time are solved.

The invention provides a high-energy steam-water-ice three-phase jet mechanism research test device and test method, and belongs to the technical field of multiphase flow thermal hydraulic test devices. The problem that in the prior art, interference of water flowing and air convection on ice layer melting measurement is not considered, and consequently experimental data have remarkable system errors is solved. The device comprises a spray pipe assembly, a visual water tank, a water level and water temperature adjusting device, a measurement and data acquisition device and a steam supply device, the spray pipe assembly and the water level and water temperature adjusting device are both arranged in the visual water tank, and the spray pipe assembly comprises a spray pipe, a target distance adjusting structure and an ice block fixing clamp. The water level and water temperature adjusting device comprises a plurality of ice trays, the longitudinal section of each ice tray is of an inverted-L-shaped structure, and the measuring and data collecting device comprises a high-speed camera, a data collecting system, a weighing sensor, a pressure sensor and a plurality of temperature sensors. The device is mainly used for research tests of steam deicing.

The application relates to the technical field of reactor thermal hydraulic and safety, in particular to a heated plate and a heat transfer simulation experiment device with the same. The heated plate comprises a bubbling plate, a heating plate and an insulation layer; the bubbling plate is formed with grooves through a cast-pressing process; the heating plate is arranged in parallel and at intervals with the bubbling plate; the insulation layer is located between the bubbling plate and the heating plate, the insulation layer is connected with the heating plate, and the insulation layer is connected with the bubbling plate through vacuum diffusion welding to enclose the grooves on the bubbling plate into vacuum bubble-shaped parts. The application can accurately simulate the thermal power boundary when the dispersion type fuel element bubbles, realize the simulation and measurement of the flow heat transfer characteristics near the vacuum bubble-shaped parts, and the measurement result has high reduction degree and good reliability.

The invention discloses a rod bundle channel annular flow characteristic prediction method based on disturbance wave dynamic parameters, and relates to the technical field of nuclear reactor thermal hydraulic and multiphase flow measurement. The method comprises the following steps: acquiring spatio-temporal evolution data of the surface of an annular flow liquid film in a rod bundle channel by adopting a non-invasive optical measurement system, and preprocessing the spatio-temporal evolution data; extracting a disturbance wave dynamic parameter set from the preprocessed data; constructing a hybrid prediction model by taking the disturbance wave dynamic parameter set as input and taking the target annular flow characteristics as output; and predicting a to-be-predicted rod bundle channel annular flow by using the trained hybrid prediction model to obtain an annular flow characteristic prediction result. The invention can overcome the defects of intrusion interference, incomplete local description or high calculation cost of the existing rod bundle channel annular flow characteristic prediction method, and provides a non-intrusive rod bundle channel annular flow characteristic high-precision prediction method which is based on disturbance wave dynamic parameters and can reflect the global spatio-temporal evolution characteristics of a flow field.

The invention relates to a three-dimensional thermal hydraulic analysis method, system, medium and equipment for a steam generator, and the method comprises the steps: S1, obtaining a geometric model of the steam generator, and carrying out the grid division of the geometric model, and obtaining a multi-scale grid model of the steam generator; s2, determining and setting a physical model, a phase change model and boundary conditions of the multi-scale grid model; and S3, solving the control equation set by using a computational fluid mechanics method to obtain the three-dimensional thermal hydraulic parameters of the steam generator. According to the method, high-precision analysis of a flow field and a temperature field of a tube bundle area is realized, and the prediction precision of key safety parameters such as void share, density and speed between heat transfer tubes is remarkably improved; the precision bottleneck caused by secondary side calculation simplification in traditional software is overcome by achieving primary side and secondary side coupling heat exchange calculation. While the precision is ensured, the optimization of the calculation efficiency and the resource consumption is realized through the combination of a multi-scale grid strategy and an Euler homogeneous flow model.

The application discloses a kind of lead bismuth and supercritical carbon dioxide heat exchanger break accident experimental system and method, the system includes lead storage tank, argon buffer tank, experimental container, high temperature and high pressure carbon dioxide storage tank, carbon dioxide storage tank and buffer tank and other auxiliary systems, specifically to simulate lead bismuth and supercritical carbon dioxide heat exchanger break accident under the temperature, pressure and bubble evolution, lead bismuth reactor loop extreme thermal hydraulic phenomenon caused;The whole experimental system is operated by different pipeline valve control loop, the pressure required by experiment is flexibly controlled using gas booster pump, different heating temperatures are obtained by adjusting heating wire power, different sizes and thicknesses of copper film are flexibly replaced to simulate heat exchanger break condition, temperature change and pressure response during supercritical carbon dioxide is discharged into lead bismuth in experimental container are monitored, and lead bismuth and supercritical carbon dioxide heat exchanger experimental data under initial different temperature and pressure are obtained.

The disclosure provides a simulation method, a running state prediction method and system of an integrated small reactor, the simulation method comprising: establishing mathematical models of a plurality of physical fields of the integrated small reactor based on design parameters and running parameters of the integrated small reactor; wherein the mathematical models comprise at least one of a natural circulation driving force model, a two-phase flow heat transfer correlation model and an emergency cooling water natural injection model; performing simulation calculation on each physical field based on the mathematical models to obtain a first simulation result corresponding to each physical field; and obtaining a target simulation result corresponding to the integrated small reactor based on the first simulation result. The disclosure introduces mathematical models of multiple physical fields dedicated to the integrated small reactor, adds a special model library such as the natural circulation driving force model, the two-phase flow heat transfer correlation model and the emergency cooling water natural injection model, accurately reproduces the coupled thermal hydraulic characteristics under the arrangement of the integrated small reactor, and ensures the accuracy and reliability of the simulation result.

The invention relates to the field of reactor core physical calculation, in particular to a reactor core physical optimization calculation method under a nuclear power unit bias operation condition, and the method comprises the following steps: 1, constructing a geometric model and a reactor core grid model of a reactor in fluid dynamics software; 2, establishing a porous medium model in fluid dynamics software; 3, setting boundary conditions in the hydrodynamic software and carrying out thermal hydraulic calculation; 4, constructing a thermotechnical and physical coupling interface, taking a thermotechnical calculation result of the hydrodynamic software as input, and carrying out physical calculation in physical calculation software; 5, taking the three-dimensional power distribution of the physical calculation software as fluid dynamics software input, and carrying out thermal hydraulic calculation to obtain an optimized fuel assembly upper power peak factor and an optimized fuel assembly volume power peak factor; and 6, repeating the steps 3-5 until the power peak factor converges. According to the method, the physical parameters of the reactor core under the bias ring operation condition are calculated, and the control capability of safe operation of the reactor is further improved.

The application discloses a reactor sub-channel comprehensive acceleration calculation method and device and electronic equipment, and relates to the technical field of reactor thermal hydraulic numerical simulation, wherein the method comprises the following steps: abstracting a sub-channel gap topology into an undirected graph and performing edge coloring grouping, and constructing and caching a CSR sparse structure and a write position index of each gap based on the topology; then, serial scheduling is performed according to the coloring grouping, a plurality of threads in the group accumulate the coupling coefficients to a global numerical array according to the pre-calculated position, and the coefficient matrix assembly is completed; finally, the asymmetric sparse linear equation set is solved by AMG pretreatment combined with FGMRES iteration, the pressure gradient solution is obtained, and the current step update is completed. Through edge coloring and sparse storage pre-calculation coupling, the application realizes lock-free parallel assembly and solves the problem of data race in multi-core concurrent writing.

The invention relates to a method for determining the flow resistance of a gap structure of a graphite pebble bed of a high-temperature gas cooled reactor. The method comprises the following steps: constructing a three-dimensional geometric model corresponding to a reactor core cavity of the high-temperature gas cooled reactor, which reflects the characteristics of a graphite pebble bed and a gap combined structure; wherein the three-dimensional geometric model comprises a pebble bed and a gap channel in the lateral direction of the pebble bed; identifying a plurality of accumulation structures corresponding to a near-wall area close to the gap channel in the pebble bed, and determining distribution data of each accumulation structure; for each stacking structure, establishing a CFD geometric model provided with a geometric parameter combination in the near-wall area based on the distribution data of the stacking structure; based on the CFD geometric model of each stacking structure, CFD calculation and fitting are carried out, and target flow resistance characteristic parameters corresponding to each stacking structure are determined. According to the method, a scientific basis and a high-precision calculation model can be more accurately provided for thermal hydraulic design, safety analysis and nuclear power engineering simulation of the high-temperature gas cooled reactor, and the accuracy of flow resistance characteristic parameters is improved.

The invention belongs to the technical field of thermal hydraulic simulation software, and particularly relates to a pilot operated safety valve modeling method and device based on Modelica, and the method comprises the steps: obtaining a modular model library architecture after the pilot operated safety valve is decomposed; constructing a corresponding basic-level model by adopting a modeling mode matched with the model characteristics of the basic-level model; the basic-level models are combined, and two-fluid six-equation modeling is adopted to obtain a component-level model; and building a pilot operated safety valve model by adopting the component-level model, and injecting system parameters into the pilot operated safety valve model. Characteristic analysis is carried out on a modular model library framework obtained after decomposition of the pilot operated safety valve, model combination and system parameter injection are carried out by adopting two fluids and six equations, so that the problem of how to ensure an efficient and accurate numerical calculation result of two-phase flow under a multi-specialty coupling condition is solved; modularization, reusability and expansibility of pilot operated safety valve modeling are achieved so as to meet the requirement of complex engineering application.

The invention relates to the technical field of reactor thermal hydraulic tests, in particular to an advanced pressurized water reactor U-shaped pipe section water seal evolution research system and method, and the system comprises a main coolant simulation loop, a crevasse simulation assembly and a visual measurement and acquisition system, the method comprises the steps of establishing an initial working condition, simulating an accident process, collecting multi-dimensional data, constructing an evolution model and determining a conversion boundary. According to the method, a dependent relationship between different flow forms and a water seal evolution mechanism, a synergistic action mechanism of driving force and flow resistance in a water seal forming process and correlation characteristics between the flow forms and fluid dynamics are established, and technical support is provided for establishment of a water seal conversion boundary.

The invention relates to a lead-based reactor core gas-liquid two-phase nuclear thermal coupling calculation method and system, and the method comprises the steps: constructing a reactor core geometric model, and generating reaction section data; the reactor core geometric model comprises a fuel rod area and a coolant area, and the reaction cross section data comprises multiple groups, multiple temperature points and multiple void fractions; calculating reactor core power distribution based on a reactor core neutronics calculation model according to the reaction cross section data; according to the reactor core power distribution, based on a reactor core gas-liquid two-phase thermal hydraulic calculation model, thermal parameters of the current time step are obtained, and the thermal parameters comprise the fuel rod temperature, the coolant temperature and the void fraction. According to the calculation method provided by the invention, the influence of the gas phase on neutronics characteristics can be accurately simulated, and the reactor core power distribution and thermal parameters in a gas-liquid two-phase flow state can be accurately predicted, so that the power transient safety problem caused by the gas-liquid two-phase flow possibly occurring in the reactor core under a serious accident can be solved; and technical support is provided for safety analysis and accident evaluation of the lead-based reactor.

The invention discloses an experimental measurement system and an experimental measurement method for fission product aerosol behaviors under a pool type lead-bismuth reactor accident, and belongs to the field of lead-based fast reactor thermal hydraulic safety. A reaction container and a telescopic rod are arranged in the pressure container; the reaction container is arranged at the bottom end of the pressure container, the telescopic rod is arranged at the top end of the reaction container, the telescopic rod is connected with the thermocouple, and the thermocouple is inserted into an insertion opening in the top end of the pressure container and is controlled to ascend and descend through the telescopic rod. The operation environment of a pool type reactor is simulated, then sampling analysis of the fission products in the lead-bismuth coolant in the whole process is achieved, the chemical combination decomposition characteristics and physical characteristics of the fission products can be known more visually, and due to the design of the pressure stabilizing system, the gas conveying system and the ventilation system, research parameters are more comprehensive, and the research efficiency is improved. Therefore, the evaporation rule of the fission product aerosol under the accident condition can be fully mastered conveniently according to most parameters of the pool type lead bismuth reactor under the accident condition.

The application provides a simulation weightlessness environment thermal hydraulic test device and a simulation weightlessness environment thermal hydraulic test method, and relates to the technical field of heat exchange testing. The simulation weightlessness environment thermal hydraulic test device is designed to solve the problem that there is no effective simulation weightlessness environment hydraulic test method. The simulation weightlessness environment thermal hydraulic test device comprises a unmanned aerial vehicle, a falling cabin and a test cabin. The test cabin comprises a weightlessness environment thermal hydraulic test platform. The unmanned aerial vehicle is detachably connected with the weightlessness environment thermal hydraulic test platform. A parachute is arranged in the falling cabin. The simulation weightlessness environment thermal hydraulic test device can improve the economy of the simulation weightlessness environment thermal hydraulic test.

The invention discloses an experimental device and method for testing the insulation performance of a nuclear release heat electric heating simulation rod, and the experimental device comprises a testing assembly, a cooling assembly, and a data collection module. At least one heating rod to be tested is arranged in the test assembly; the test assembly supplies power to the positive pole point of the to-be-tested heating rod in the test assembly, and supplies power to the to-be-tested heating rod to the rated power of the to-be-tested heating rod; the cooling assembly cools the to-be-tested heating rod and the positive and negative electrode insulating layer points. Therefore, before the design of the electric heating reactor core of the reactor, the insulation performance experiment test can be carried out on the to-be-tested heating rod and the minimum unit which form the electric heating simulation reactor core, the insulation failure boundary of the to-be-tested heating rod in the experiment is explored, the experiment support is provided for the electric heating simulation reactor core power connection design in the reactor thermal hydraulic experiment, and the test efficiency is improved. Therefore, the safety of the reactor thermal hydraulic experiment is ensured.