33 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 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 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.

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.

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 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 invention relates to the technical field of nuclear reactors, in particular to a neutron energy spectrum one-way regulation and control device and a lead-based fast neutron high-flux research reactor, and the regulation and control device comprises a first structure layer which comprises at least two neutron moderation elements, and a cylindrical irradiation cavity defined by the multiple neutron moderation elements is used for accommodating an irradiation target; the second structural layer comprises at least two first adaptive elements, and the multiple first adaptive elements can form a first cylindrical structure on the periphery of the first structural layer; the third structural layer comprises at least two second adaptive elements, and the multiple second adaptive elements can form a second cylindrical structure on the periphery of the second structural layer; wherein the plurality of first adaptive elements and the plurality of second adaptive elements are loaded with at least one of a low-energy thermal neutron absorbing material and a high-energy fast neutron strengthening material, and the energy of high-energy fast neutrons generated by nuclear fission of the reactor core can be reduced to a specific energy interval. The research stack is loaded with the regulation and control device.

A method for Mo-99 irradiation production and energy spectrum optimization based on reactor activation is proposed. This method establishes a three-dimensional high-fidelity neutronics and burnup coupling model of the reactor and irradiation channels. The full neutron energy range is divided into 238 energy regions, and perturbations are applied to the flux of each individual energy region. This yields the importance indicators and curves of Mo-99 production on flux changes in each energy region, thereby identifying positively and negatively important performance regions. For negatively important performance regions, filter nuclides are screened from the absorption cross-section characteristics of natural isotopes, and burnup calculations are performed at different doping levels to obtain production gains. The isotope scheme is then converted into a natural abundance material scheme to determine the optimal filter material and loading method. This invention, through integrated energy spectrum optimization—refined energy spectrum analysis, filter nuclide screening, determination of the optimal energy spectrum control scheme, and impurity assessment—can increase Mo-99 production without adding extra post-processing steps and reduce the risk of radioactive impurity introduction. It is applicable to the industrial and regional supply of Mo-99 for various reactor platforms, including commercial and research reactors.

The invention belongs to the technical field of fast neutron research reactor irradiation, and particularly relates to an irradiation device, a connection system and a method suitable for a lead-bismuth irradiation loop. According to the device, an air gap sleeve is placed in a heat insulation pipe, and a gas heat insulation layer is formed between the heat insulation pipe and the air gap sleeve; the pressure pipe is placed in the air gap sleeve, a high-temperature gas layer is formed between the air gap sleeve and the pressure pipe, the flow dividing pipe is placed in the pressure pipe, the test section is placed in the flow dividing pipe, the bottom of the flow dividing pipe is open, a lead-bismuth downward flow channel is formed between the pressure pipe and the flow dividing pipe, and the lead-bismuth downward flow channel returns from the bottom and enters the flow dividing pipe. The heat of the test section is taken away. The device is used for a lead-bismuth loop irradiation test of a lead-bismuth research reactor, provides a technological process and a connection system for connection / decoupling of the lead-bismuth loop and the device, solves the problem of decoupling of the lead-bismuth irradiation device and the loop after shutdown and before discharge of the device, and can also be suitable for decoupling of the lead-bismuth loop and the device of a water-cooling research reactor.

A research reactor test loop element damage monitoring system and method, relating to the field of irradiation technology for research reactors, includes a test heat exchanger, a first control valve, a delayed neutron monitor, a total gamma monitor, and an online nuclide monitor connected in sequence. The cooling inlet of the test heat exchanger is connected to the test loop, and the cooling outlet of the test heat exchanger is connected to the inlet of the delayed neutron monitor via the first control valve. The outlet of the delayed neutron monitor is connected to the inlet of the total gamma monitor. The outlet of the total gamma monitor is connected to the inlet of the online nuclide monitor. The outlet of the online nuclide monitor is connected to the regeneration inlet of the test heat exchanger via a first flow regulating valve. The regeneration outlet of the test heat exchanger is connected to the test loop. This system can accurately determine whether fuel elements have been damaged.

The present application relates to research reactor device technical field, disclose a kind of research reactor device, the research reactor device includes: pressure vessel, conversion component, first operating member and second operating member, the top of pressure vessel is equipped with first operating window and second operating window, first operating window and second operating window can open and close pressure vessel, conversion component is located in pressure vessel, conversion component includes material storage part, under the condition that conversion component rotates, material storage part can be opposite to second operating window, first operating member is separable with conversion component Connection, first operating member is used to pass through first operating window, to drive conversion component rotation along the circumference of pressure vessel, second operating member is used to pass through second operating window, to take the loading and unloading material stored on material storage part. Operating personnel can be taken out by remote operation mode, without opening the top cover of pressure vessel, the loading and unloading material in its interior.

The invention relates to the technical field of nuclear reactors, in particular to a passive residual heat removal system and a lead-based fast neutron high-flux research reactor. The residual heat removal system comprises a reflecting layer structural part which is of a hollow structure with through holes formed in the two ends, the reflecting layer structural part is vertically arranged on the outer side of the fuel assembly in a surrounding mode in the working state, and an inner cavity and a cooling runner of the fuel assembly are arranged in parallel; the lower end of the closed air cooling container is provided with an air inlet channel, the upper end of the closed air cooling container is provided with an exhaust channel, and the air inlet channel can be opened and closed and is hermetically sleeved outside the reaction area of the reactor container in a working state; the discharge chimney is vertically arranged, and the lower end of the discharge chimney is connected with the exhaust channel, so that the residual fission heat and decay heat of the reactor core can be efficiently exported under the working condition of long-term normal shutdown or accident shutdown of the reactor, and the damage of in-reactor fuel elements and in-reactor components caused by overheating of the reactor core is avoided; and the inherent safety of the lead-based fast neutron high-flux research reactor is obviously improved. The research stack is loaded with the waste heat removal system.

The invention discloses a thermotechnical device and method for a uranium hydrogen zirconium reactor liquid metal irradiation test loop. The thermotechnical device comprises a detachable dry well, an in-reactor arrangement environment and an out-reactor liquid metal loop. The detachable dry well is located in the in-reactor arrangement environment and connected with the oil cold-heat exchanger, the calibration barrel, the liquid metal flowmeter, the driving pump, the driving pump bypass and the preheating heating section through pipelines to form an irradiation loop. The detachable dry well for flowing liquid metal is arranged in the research reactor, the liquid metal irradiation loop environment is set up, the dry well and the research reactor are decoupled, and the dry well and the research reactor can be independently disassembled and assembled and transformed according to needs; by building the out-of-pile liquid metal loop and connecting the out-of-pile liquid metal loop with the detachable dry well flange, flowing of liquid metal is achieved, and the out-of-pile liquid metal loop and the detachable dry well are separately overhauled and assembled. According to the design method and device, the liquid metal irradiation environment can be built, and the pressure boundary integrity of the reactor pool is fully considered and researched.

The invention relates to the technical field of research reactor earthquake monitoring instrument verification, and provides a verification system and method for a research reactor earthquake monitoring instrument. A touch display screen receives earthquake signal parameters and interface type parameters of the earthquake monitoring instrument; the microprocessor calls a seismic waveform database according to the seismic signal parameter to generate a seismic generation signal and form an expected response of the seismic generation signal, and converts the seismic generation signal into a seismic output signal according to the interface type parameter; the target test cable transmits an earthquake output signal to the earthquake monitoring instrument and receives response data transmitted by the earthquake monitoring instrument; the microprocessor collects response data and determines a verification result of the earthquake monitoring instrument according to the response data and expected response; and the touch display screen displays the verification result. The system provided by the invention can complete the performance and function verification work of the research reactor earthquake monitoring instrument during the operation period of the research reactor.

The invention relates to the technical field of neutron irradiation test, and provides a fuel stub thermal resistance medium packaging device and a test method.The fuel stub thermal resistance medium packaging device comprises a test section, a first clamping block and a second clamping block are stacked in the test section in the flowing direction of cooling water, a first test capsule is limited in the first clamping block, and a second test capsule is limited in the second clamping block; and the cooling water channel, the air pipe assembly and the thermocouple assembly are matched to construct a modularized packaging structure, so that the high-heat-release fuel short rod can be stably borne, the temperature can be monitored in real time, and the safe development and data reliability of a low-temperature irradiation test in a reactor are ensured. A systematized in-reactor verification scheme for the composite thermal resistance layer is formed by constructing an initial test environment, increasing and researching reactor power step by step, replacing gas in batches, and comparing whether the temperature influence of elastic thermal resistance exists or not based on collected temperature data, and accurate parameter support is provided for subsequent engineering application.