46 results about "Research reactor" patented technology

The invention relates to a mechanical property simulation method for researching the service life of an aluminum alloy material for a reactor. The simulation method for the life period mechanical property of the aluminum alloy material comprises the following steps: S1, preparing initial, middle and last stage aluminum alloy samples according to a < 27 > Al-< 28 > Si transmutation path of the aluminum alloy material; s2, representing the sample by using a transmission electron microscope, obtaining precipitated particle parameters, calculating a volume fraction f, fitting f and operation time t, and constructing a precipitation evolution model f (t); s3, performing high-energy ion irradiation on the sample, and performing nanoindentation experiment and load-depth curve fitting to obtain nano hardness H; s4, based on the f (t), establishing a relation model of the nano hardness H and the irradiation damage D; and S5, calculating the change quantity of the nano hardness H, converting the change quantity into the change quantity of the yield strength, associating the irradiation damage D with the service time T, and establishing a life period yield strength prediction model. According to the method, the mechanical property of the aluminum alloy in the full service cycle of the research reactor can be accurately evaluated.

The present application relates to the technical fields of research reactor material irradiation, and specifically relates to a kind of with autonomous rotating function's pressure hole irradiation device, including remote terminal and sequentially connected driving section, protection tube and test section, driving section is set in the pressure hole mouth outside, including transmission shaft, power transmission component, servo motor and controller, transmission shaft is connected with protection tube, servo motor is installed in the side direction of transmission shaft, controller is electrically connected with servo motor, remote terminal is connected with controller, power transmission component includes transmission shaft gear and the motor gear meshed with transmission shaft gear, transmission shaft gear is coaxially keyed with transmission shaft, motor gear is coaxially keyed with servo motor output shaft;Transmission shaft and protection tube are provided with the gas outlet pipe and several gas inlet pipes communicated with test section.In the present application, the mode of remote control of servo motor is used, the irradiation device of the test piece can be realized autonomous rotation and test piece temperature control function, and the accuracy and reliability of test result are improved.

The application provides a yield optimization method, system and medium for isotope irradiation production in a research reactor, and relates to the technical field of nuclear research reactors. The yield optimization method comprises the following steps: obtaining design data of the research reactor, irradiation hole usage data range and total required number of at least one type of isotope target; constructing a research reactor Monte Carlo model associated with the research reactor according to the design data; determining a flux influence factor and a multi-group microscopic cross section corresponding to the at least one type of isotope target, and a target-free neutron flux field function according to the research reactor Monte Carlo model; and taking, as a yield optimization result, layout data of each irradiation hole corresponding to a preset isotope yield target through burnup calculation according to the irradiation hole usage data range, the total required number of irradiation holes, the flux influence factor and the multi-group microscopic cross section corresponding to the at least one type of isotope target, and the target-free neutron flux field function.

The invention relates to the technical field of nuclear reactors, in particular to a control rod, a lead-based fast neutron high-flux research reactor and an experimental method thereof. Wherein the research reactor comprises a reactor container, a supporting platform, a shielding layer, a reactor lower chamber, a guide pipe and a control rod; a fuel assembly is arranged in the reactor container; one end of the guide pipe is located in the reactor container and is of a sealed structure, the length direction of the guide pipe is parallel to the length direction of the fuel assembly, the other end of the guide pipe sequentially penetrates through the reactor container, the supporting platform, the shielding layer and the under-reactor chamber and is located in the under-reactor chamber, and a driving mechanism matched with the guide pipe is arranged in the under-reactor chamber; the control rod is located in the guide pipe and can be driven by the driving mechanism to move in the length direction of the guide pipe. In a working state, the height of one end of the sealing structure of the guide pipe is not lower than that of the fuel assembly. According to the invention, the control rod can be installed, enough top space is reserved for the reactor vessel, and the reactor vessel has inherent safety characteristics.

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 relates to the technical field of nuclear reactors, in particular to a lead-based fast neutron high-flux research reactor with coolant flowing in different areas, which comprises a reactor container provided with a first coolant flow channel and a second coolant flow channel which are connected in parallel; the fuel assembly is loaded with nuclear fuel, is mounted in the reactor container and is positioned at the outlet of the first coolant flow channel; the reflecting layer assembly is a structural assembly, is mounted in the reactor container, is arranged outside the fuel assembly in a surrounding manner, and is positioned at an outlet of the second coolant runner; the lead-based cooling agent output from the second cooling agent flow channel flows to an outlet of the first cooling agent flow channel after flowing through the reflecting layer assembly, and the flow of the lead-based cooling agent flowing through the second cooling agent flow channel is smaller than that of the lead-based cooling agent flowing through the first cooling agent flow channel. The design and manufacturing difficulty of the fuel assembly and the reflecting layer assembly can be reduced.

The invention relates to the technical field of irradiation tests, and provides a regional test method and device for materials in a research reactor, and the test method comprises the following steps: dividing test regions and test stages in the research reactor, and dividing test sample batches according to specific characteristics; the test areas are divided into three types: an area I (a reactor core active section area containing a neutron field, a gamma field and a water chemical environment), an area II (a reactor core non-active section area containing a gamma field and a water chemical environment) and an area III (a reactor core outer loop inner area and only a water chemical environment); the test stages are divided into three classes: a stage I (cold-state working condition, normal temperature, less than or equal to rated pressure and less than or equal to rated flow), a stage II (hot-state normal working condition, rated temperature, rated pressure and rated flow) and a stage III (hot-state reinforced working condition, rated temperature, rated pressure and higher than rated flow). According to the method, the influence factors can be quickly locked by comparing the performance changes and defects of the same batch of samples in three areas after three-stage tests.

The application belongs to the technical field of nuclear engineering, and particularly relates to a beryllium sleeve, a core assembly and a research reactor irradiation test device. The beryllium sleeve comprises a support, a beryllium block, a first fixing block and a second fixing block. The support has a core channel suitable for the passage of neutrons. The beryllium block is sleeved on the support, and a plurality of beryllium blocks are arranged adjacently in the extension direction of the core channel. The first fixing block cooperates with the support and the beryllium block. The second fixing block cooperates with the support and the beryllium block. In the extension direction of the core channel, the first fixing block and the second fixing block are respectively located on the two sides of the beryllium block. The beryllium sleeve provided by the application has strong structural stability and high reliability. When the beryllium sleeve is applied to the research reactor irradiation test device in combination with the core assembly, the neutrons can be effectively reflected to slow down the number attenuation of the neutrons, and the neutron economy is improved.

The application provides a simulation prediction method and system for slow positron source intensity in a research reactor, comprising the following steps: S1, reading arrangement information of components in a reactor core of the research reactor; S2, constructing a Monte Carlo model of the reactor core containing a slow positron source device, and obtaining neutron multi-group angular flux and photon multi-group angular flux information near the slow positron source generating device; S3, constructing a slow positron Monte Carlo model; S4, performing energy-angle double sampling to obtain an incident neutron energy probability density distribution function f n,ig,ia , an angle probability density distribution function f n,ia , an incident photon flux energy probability density distribution function f g,ig,ia , an angle probability density distribution function f g,ia , and a neutron-photon weight factor; S5, performing function sampling of particle types, energy and angle; and S6, obtaining slow positron distribution information generated in the slow positron generating device. The application accurately considers the whole life cycle of slow positron generation and transportation through multi-stage coupling, and realizes accurate prediction of slow positron source intensity.

The invention relates to the technical field of research reactor earthquake monitoring, and provides a research reactor earthquake monitoring system and method, and the system comprises a plurality of collection ends, a plurality of digital strong motion instruments, and a terminal cabinet. Each acquisition end acquires an acceleration sensing signal of each monitoring point and transmits the acceleration sensing signal to each corresponding digital strong motion seismograph; each digital strong-motion seismograph converts the acceleration sensing signal into an acceleration digital signal, judges whether to generate single-monitoring-point alarm data according to the acceleration digital signal, and transmits the single-monitoring-point alarm data to the terminal cabinet when the single-monitoring-point alarm data is generated; and the terminal cabinet calculates a response spectrum and / or a cumulative absolute speed according to the alarm data of the single monitoring point, judges whether to generate stack stop alarm data or not according to the response spectrum and / or the cumulative absolute speed, and gives an alarm when the stack stop alarm data is generated. According to the system provided by the invention, the accuracy of earthquake monitoring is improved, and false triggering heap stop is prevented.

The invention relates to the technical field of research reactor irradiation, in particular to a fuel rod simulation heat transfer device. The sleeve assembly is arranged on the heating pipe in a sleeving mode, and a liquid cooling channel is formed in the sleeve assembly; the heat insulation layer is arranged on the heating section of the heating pipe in a sleeving mode and located between the heating pipe and the sleeve assembly; wherein a first cavity is formed between the heating pipe and the heat insulation layer; a second cavity is formed between the heat insulation layer and the sleeve assembly, and the first cavity and the second cavity are used for being filled with inert gas; and the temperature detection assembly is used for detecting the temperature of the heating pipe, the sleeve assembly and the heat insulation layer. Through a heat transfer system of the first cavity, the heat insulation layer, the second cavity, the sleeve assembly and the liquid cooling channel, the temperature gradient is increased, the outer layer temperature of the sleeve assembly is close to the coolant temperature under the condition that the heating pipe maintains the high temperature, the steady-state irradiation test of the high-temperature fuel rod in the low-temperature coolant is achieved, and it is guaranteed that new radioactive waste cannot be introduced; and a foundation is laid for smooth implementation of the irradiation test in the fuel rod pile.

The present invention discloses a rapid automated modeling method, system, and medium applicable to research reactors, and belongs to the field of thermal hydraulic technology of research reactors. The method includes: establishing position numbers for different positions of the reactor core; establishing coordinates under the position numbers; collecting layout information of components in the reactor core, establishing coordinate information of components in the reactor core; establishing component structured data corresponding to the positions; sequentially reading information of each position of the reactor core, establishing a component model of the entire reactor core, and completing rapid automated modeling of the entire core. The rapid automated modeling method applicable to research reactors provided by the present invention models and analyzes each type of component in the core (including beryllium components, fuel components, aluminum components, control rod components, and irradiation channels) separately, and finally establishes a one-to-one corresponding position assembly relationship for the core layout in a programming manner, thereby realizing three-dimensional automatic modeling.

The application provides a complex reactor core collaborative design system suitable for research reactors, relates to the technical field of reactor core design systems, and aims to improve the reliability of reactor core design, and comprises a model library module, a reactor core scheme design module, a reactor core scheme evaluation module and an interactive interface module. The model library module comprises a reactor model library and an irradiation device model library, and is established based on different research reactors and various types of irradiation test pieces in the reactors. The reactor core scheme design module is used for independently initializing and designing the reactor core, solving the problems of missing fuel assemblies and target piece positions caused by channel opening operation, and solving the regional demand of newly added beryllium aluminum assemblies caused by channel closing operation. The reactor core scheme evaluation module is used for evaluating the reactor core. The application has the advantages of intelligent design, easy operation and reliable design quality.

The invention discloses a target assembly for carrying out a research reactor irradiation production isotope test, comprising: an outer sleeve, the upper end of which is provided with a coolant inlet and the lower end of which is provided with a coolant outlet; the multiple target pieces are suitable for being detachably arranged in the outer sleeve, the axial direction of each target piece installed in the outer sleeve is parallel to the axial direction of the outer sleeve, each target piece comprises a target piece core body and a target piece wrapping shell, the target piece core body is arranged in the target piece wrapping shell, and the target piece wrapping shell is arranged in the target piece core body. Each target piece core body comprises a plurality of target piece units, and the plurality of target piece units of each target piece core body are arranged along the axial direction of the target piece. According to the embodiment of the invention, the target assembly for carrying out the research reactor irradiation production isotope test has the advantages of flexibility, diversity, high target material utilization rate, high irradiation test efficiency and the like.

The invention relates to the technical field of aging management of research heaps, and provides a self-learning method, device and equipment for a fault diagnosis model of a research heap and a medium, and the method comprises the steps: obtaining a trained initial fault diagnosis model; performing fine tuning on the initial fault diagnosis model to enable the initial fault diagnosis model to distinguish unknown faults; acquiring an equipment characteristic test parameter set, performing judgment according to the fine-tuned initial fault diagnosis model, and manually marking corresponding newly-added fault classifications for equipment characteristic test parameters belonging to unknown faults; adjusting the initial fault diagnosis model according to the newly added fault classification; and optimizing the adjusted initial fault diagnosis model by using the equipment feature training parameter set and the equipment feature test parameter set for training the initial fault diagnosis model and the corresponding fault classification to obtain an updated fault diagnosis model. According to the method provided by the invention, the unknown fault can be distinguished, and the model can be subjected to self-learning iterative updating, so that the equipment monitoring efficiency and the fault diagnosis accuracy are greatly improved.

This invention relates to the field of nuclear reactor core design optimization technology, specifically to a method and system for optimizing the loading of a research reactor core based on hierarchical cross-evolution. The method includes: acquiring core geometric parameters, irradiation mission objectives, and an initial loading scheme; constructing a hierarchical optimization model, decomposing the core loading optimization problem into a fuel assembly optimization layer and a non-fuel assembly optimization layer; performing evolutionary algorithm optimization on the fuel assembly optimization layer to generate a fuel assembly arrangement scheme; dynamically defining the decision space of the non-fuel assembly optimization layer based on the fuel assembly arrangement scheme, generating a non-fuel assembly arrangement scheme; forming a core loading scheme based on the fuel assembly and non-fuel assembly arrangement schemes; performing multi-objective fitness evaluation and iterative optimization on the core loading scheme until preset conditions are met, and outputting the final core loading scheme and related parameters. Its purpose is to solve the asymmetric core arrangement problem in multi-mission irradiation scenarios and improve the efficiency of core design optimization.

The present application relates to the technical field of nuclear reactor structure, and particularly relates to a reactor bottom structure for research reactor and a mounting method, which is mounted in a lower chamber and comprises: an anti-seismic support structure fixed on the bottom surface of the lower chamber, a blowdown pipeline system fixed on the anti-seismic support structure, a ventilation and cooling structure fixed on the top surface of the lower chamber, and a control rod drive mechanism installed upside down; a stator coil in the control rod drive mechanism is arranged in a cylinder in the ventilation and cooling structure, a pressure shell in the control rod drive mechanism is arranged in the anti-seismic support structure, an anti-seismic plate in the anti-seismic support structure is fixed on the pressure shell, and a drain connection pipe arranged at the bottom end of the control rod drive mechanism is connected with the blowdown pipeline system. The reactor bottom structure has compact structure size and provides limiting and cooling for the control rod drive mechanism, and the reactor bottom structure is further provided with the blowdown pipeline system, which regularly discharges radioactive waste to reduce the irradiation dose of the lower chamber.

This application relates to the field of nuclear engineering technology, specifically to a beryllium sleeve, a core assembly, and a research reactor irradiation testing device. The beryllium sleeve includes a support member, beryllium blocks, a first fixing block, and a second fixing block. The support member has a core channel suitable for neutron passage. Several beryllium blocks are attached to the support member, wherein, in the circumferential direction of the core channel, the beryllium blocks are arranged adjacently to enclose the core channel circumferentially, and in the extending direction of the core channel, the beryllium blocks are arranged adjacently to cover the core channel axially, with aluminum components fitting between adjacent beryllium blocks. The first fixing block fits with the support member and also fits with the beryllium blocks; the second fixing block fits with the support member and also fits with the beryllium blocks. The beryllium sleeve structure provided by this application has strong stability and high reliability. When used in conjunction with a core assembly in a research reactor irradiation testing device, it can effectively reflect neutrons to slow down neutron decay, while the aluminum components reduce neutron adsorption and improve neutron economy.

The present invention provides a cooling water system for a research reactor irradiation test loop and its design and use method, which relates to the technical field of nuclear fuel and material in-pile irradiation, and solves the technical problem that the complete interruption of the external water source in the existing cooling water system will bring about safety hazards. In the present invention, the volume corresponding to the first liquid level alarm point and the second liquid level alarm point is the operating buffer volume V1, the volume between the second liquid level alarm point and the first water intake point is the minimum cooling water volume V2 in the early stage of shutdown, and the volume between the first water intake point and the second water intake point is the minimum cooling water volume V3 in the late stage of shutdown; the sum of V1, V2 and V3 is the optimal volume V of the cooling water pool. The present invention simplifies the complexity of the system, reduces the space occupied by the equipment, and reduces the construction cost. At the same time, it ensures the reliability of the equipment and system through the nuclear safety level; ensures the safety of the fuel elements, optimizes the size design of the cooling water pool, and further reasonably utilizes the space between the research reactor process rooms.

The invention relates to the technical field of nuclear reactor design, and particularly discloses a lead-based fast neutron high-flux research reactor which comprises a container, a fuel assembly and a reactivity control system, and the fuel assembly is arranged in the container. The reactivity control system comprises an electric control device, a gas pressure control device and a gas volume control device; the electric control device, the gas pressure control device and the gas volume control device are sequentially arranged from inside to outside in the radial direction of the reactor core; the sum of the total reactivity values of the reactivity control systems is configured to be sufficient to enter and maintain the core at a prescribed subcritical depth under the worst residual reactivity conditions. Therefore, reactivity control means can be enriched, and the problem of limitation of a reactivity control mode of a nuclear reactor in a lead-based fast neutron high-flux research reactor can be solved.

The invention relates to the technical field of research reactor devices, and discloses a research reactor reaction device which comprises a pressure container, a conversion assembly, a first operation piece and a second operation piece, the top of the pressure container is provided with a first operation window and a second operation window, and the first operation window and the second operation window can open and close the pressure container. The conversion assembly is arranged in the pressure container and comprises a material storage part, under the condition that the conversion assembly rotates, the material storage part can be opposite to the second operation window, the first operation part is detachably connected with the conversion assembly, and the first operation part is used for penetrating through the first operation window to drive the conversion assembly to rotate in the circumferential direction of the pressure container; the second operation piece is used for penetrating through the second operation window so as to take loaded and unloaded materials stored on the material storage piece. An operator can take out the loaded and unloaded materials in the pressure container through a remote operation mode without opening the top cover of the pressure container.

The invention discloses a multi-physics coupling simulation method, device, equipment and medium for the combination of a research reactor deterministic theory and a probability theory. The method comprises the following steps: completing multi-cycle burnup calculation of a research reactor core under a preset boundary condition; calculating fuel fission heat release and gamma heat release of each target area of the research reactor core; determining coolant inlet flow velocity of each fuel element of the research reactor core; the minimum DNBR of the fuel element is determined and used for thermotechnical safety margin evaluation; feeding back a thermotechnical calculation result of the fuel element test loop to the physical model through a physical-thermotechnical parameter transmission channel, and obtaining a corresponding temperature calculation section based on the square root of the existing temperature section database and an adjacent temperature weighted interpolation; and iterating the section parameters and the material thermophysical property parameters to obtain a multi-physical field coupling analogue simulation result of the research reactor core. The invention belongs to the physical and thermal field of reactors. According to the method, the deviation of reactor physical design and target irradiation calculation can be reduced.

The invention relates to the technical field of irradiation tests, and provides a neutron screen structure for studying reactor instrumentation fuel stub irradiation tests, which comprises a test section, a neutron screen assembly, a neutron shielding assembly, a neutron shielding assembly, a neutron shielding assembly, a neutron shielding assembly, a neutron shielding assembly and a neutron shielding assembly, the test section is internally provided with a sample cavity, and the neutron shielding assembly is capable of absorbing thermal neutrons and transmitting fast neutrons; the neutron screen assembly is of an annular structure, a clamping structure used for loading a test sample is arranged in the neutron screen assembly, an inner side cooling flow channel is formed between the neutron screen assembly and the clamping structure, and an outer side cooling flow channel is formed in the outer side of the neutron screen assembly. Energy spectrum transformation of a local area of the fuel stub is realized, so that a fast spectrum prototype reactor environment is simulated in a thermal spectrum research reactor, and the irradiation performance of fast reactor fuel in a service environment can be better reflected.

The invention provides a method and a system for simulating and predicting the intensity of a slow positron source in a research reactor. The method comprises the following steps: S1, reading arrangement information of components in a reactor core of the research reactor; s2, constructing a research reactor core Monte Carlo model containing a slow positron source device, and obtaining neutron multi-group angle flux and photon multi-group angle flux information near a slow positron source generation device; s3, constructing a slow positron source Monte Carlo model; s4, performing energy and angle double sampling to obtain incident neutron energy probability density distribution functions fn, ig and ia, angle probability density distribution functions fn and ia, incident photon flux energy probability density distribution functions fg, ig and ia, angle probability density distribution functions fg and ia and neutron and photon weight factors; s5, function sampling of particle types, energy and angles is carried out; and S6, acquiring slow positron distribution information generated in the slow positron generation device. According to the method, the generation of the slow positron and the full life cycle of transportation are accurately considered through multi-stage coupling, and accurate prediction of the intensity of the slow positron source is realized.

The application discloses a research reactor overall design system and method, which realizes the whole-chain optimization of the research reactor overall design from demand input to parameterization and from single-discipline calculation to multi-specialty cooperation through a closed-loop mechanism of demand traction, two-way coupling of the reactor core and irradiation application, cross-disciplinary iterative optimization and comprehensive performance evaluation. Compared with the prior art, the application can significantly improve the efficiency and reliability of the research reactor overall design, ensures that the design result meets the scientific research and application tasks, the safety and economic requirements, and thus guarantees the multifunctional adaptability and global optimization of the research reactor in the early design stage.

The present application relates to the technical field of taking and placing tools, and provides a device for studying underwater taking and placing of a core block during fuel replacement of a reactor, which comprises a cylinder, a sliding rod coaxially and slidably connected in the cylinder, an actuating mechanism arranged at one end of the sliding rod and used for driving the sliding rod to move and reset, a grabbing assembly connected to the other end of the sliding rod, the grabbing assembly comprising a positioning sleeve, a grabbing hook and a jacking rod, the positioning sleeve being connected to the cylinder and used for being inserted into a circular hole arranged on the positioning block, the grabbing hook being movably connected to the positioning sleeve through a pin shaft, the jacking rod being slidably connected to the positioning sleeve, one end of the jacking rod being connected to the sliding rod, and the other end of the jacking rod abutting against the middle of the pair of grabbing hooks and used for opening or closing the grabbing hook to take or place the positioning block, the positioning sleeve is inserted into the circular hole arranged on the positioning block, the sliding rod is driven by the actuating mechanism to move or reset, the jacking rod is driven by the sliding rod to abut against the grabbing hook, the taking and placing of the positioning block are realized, the positioning operation is facilitated, the positioning block is effectively prevented from falling, and the safety and reliability of the operation are improved.

The invention relates to the technical field of nuclear reactors, and particularly discloses a lead-based fast neutron high-flux research reactor and a reflecting layer structure thereof. The plurality of reflecting layer structures are arranged around a fuel assembly of the reactor core active assembly; the box walls of the reflecting layer structures are made of high-temperature-resistant and lead-based alloy corrosion-resistant materials, and each reflecting layer structure is provided with a cavity to form a neutron reflecting layer; the upper end of the reflecting layer structure forms a completely open structure, the lower end of the reflecting layer structure is fixed on a lower grid plate of the reactor core, and at least one coolant runner is arranged at the lower end of the reflecting layer structure, so that a coolant can only flow up and down in the cavity; the shape of the inner surface of the reflecting layer structure is matched with that of the outer surface of the outermost fuel assembly of the reactor core active assembly, and the shape of the outer surface is matched with that of the inner side surface of the reactor core hanging basket. The lead-based fast neutron high-flux research reactor has a reflecting layer structure. Therefore, the reactor core coolant bypass flow is effectively reduced, the material changing operation is simplified, and the flexible arrangement requirements of positioning cylinders and irradiation devices with different sizes are met.

The application discloses a secondary supporting device suitable for a research reactor, which comprises a supporting assembly, a buffer plate, a supporting cylinder, a measuring plate and a pressing assembly, the lower side of the buffer plate is connected with the upper end of the supporting assembly, the lower end of the supporting cylinder is connected with the buffer plate, the measuring plate is connected with the upper end of the supporting cylinder, the pressing assembly is fixedly arranged on the upper side of the measuring plate, and a hanging basket assembly applies a downward force to the pressing assembly; the measuring plate is arranged on the upper end of the supporting cylinder, and the measuring plate provides arrangement space for measuring the outlet temperature of a reactor core, so that the removal and replacement of a damaged thermocouple are facilitated.

The invention relates to the technical field of research reactor irradiation, in particular to a fuel high-temperature reactor instrumentation irradiation test method and system suitable for a research reactor, and the method comprises the following steps: determining irradiation test technical index parameters, and carrying out the design of a high-temperature irradiation test piece according to the irradiation test technical index parameters; an irradiation test device is designed according to a high-temperature irradiation test piece structural design scheme, irradiation test thermal-hydraulic performance verification is carried out by combining the high-temperature irradiation test piece and the irradiation test device structural design scheme, and safety analysis under accident conditions is carried out based on irradiation test thermal-hydraulic performance verification data. And designing an irradiation test scheme based on a safety analysis result. From determination of technical index parameters to final generation of an irradiation test scheme, each stage has clear requirements and standards, so that the working efficiency is improved, human errors are reduced, subsequent verification and improvement are facilitated, accurate monitoring of a high-temperature environment can be supported, and the deviation between a test condition and an expected working condition is reduced.

The application discloses a kind of fuel assembly irradiation test methods based on high temperature high pressure test loop, it is related to the technical field of research reactor fuel irradiation, by determining the technical index parameter of fuel assembly irradiation test, the preliminary scheme of irradiation test system is designed, the safety analysis of the scheme under accident condition is carried out, and simulation is irradiated into the reactor, based on the results of safety analysis and test, determine the reactor time and the running state of device and system, finally, according to the reactor time and running state data, it is judged whether the fuel assembly in reactor core meets the technical index requirement in the simulation operation of research reactor;The method provides a set of systematic, standardized process, from the determination of technical index parameter to the final generation of irradiation test system, each has definite requirement and standard, so as to help improve work efficiency, reduce human error, and facilitate subsequent verification and improvement.