88 results about "Research reactor" patented technology

The invention provides an overpressure protection design method for an irradiation test loop, and belongs to the technical field of research reactor irradiation, and the overpressure protection design method for the irradiation test loop comprises the steps: constructing a first model of the irradiation test loop, and constructing a second model of the irradiation test loop; assuming a starting condition causing overpressure protection and a boundary condition of the overpressure protection device, when the leakage detection system is overpressure, the pressure of the leakage detection system does not exceed the pressure bearing limit of the heat insulation pipe by adjusting the sectional area of the spraying section and the spraying starting time of the overpressure protection device; the appropriate overpressure protection device is selected according to the calculated spraying sectional area size and the calculated spraying delay starting time, pressure relief is carried out on an irradiation test loop in time, the pressure of a leakage detection system is prevented from exceeding the pressure bearing limit of a heat insulation pipe, and the leakage detection reliability of the heat insulation pipe is improved. And the working safety and reliability of the irradiation test loop are ensured.

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 invention discloses a research reactor fault detection method, device and system and a storage medium, relates to the technical field of nuclear power, and aims to at least solve the problems that a fault detection result is fuzzy and inaccurate, the troubleshooting workload is large and the fault processing efficiency is low. The method is applied to a research heap system and comprises event nodes, state nodes and a preset change trend, caused by fault events corresponding to the event nodes, of the state nodes associated with the event nodes. Comprising the steps of determining a first event node associated with a target state node under the condition that the target state node is detected to be in a fault state; determining a first state variable quantity of a state variable of a first state node associated with the first event node; and selecting a second event node of which the change trend of the first state change quantity associated with the event is the same as the target preset change trend corresponding to the first state node from the first event nodes, so as to determine the fault event of the second event node as a target fault event.

The invention provides a complex reactor core collaborative design system suitable for a research reactor, relates to the technical field of reactor core design systems, aims to improve the reliability of reactor core design, and comprises a model library module, a model database module, a model database module and a model database module, the model library comprises a reactor model library and an irradiation device model library; the interactive interface module is used for realizing interactive operation; the reactor core scheme design module is used for carrying out autonomous initialization design on a reactor core, and solving the problem of fuel assembly and target position loss caused by pore channel opening operation and the area demand of a newly added beryllium aluminum assembly caused by pore channel closing operation; and the reactor core scheme evaluation module is used for evaluating the reactor core. The method has the advantages of being intelligent in design, easy to operate and reliable in design quality.

The invention relates to the technical field of isotope irradiation production, in particular to a target material arrangement device suitable for researching reactor irradiation production of iride-192. The target material arrangement device mainly comprises a cladding, a plurality of clamping blocks and iridium sheets, the clamping blocks are arranged on the inner side of the cladding and arranged around the circumference of the central axis of the cladding, and the iridium sheets are arranged on the clamping blocks; and the iridium sheet is arranged on one side, in contact with the cladding, of the clamping block. Through the structure, the target material in the target material arrangement mode of the iride-192 is designed into a pure I < r > circular sheet. The small-size target material can improve the irradiation uniformity of a single target material and can be used for manufacturing a small-size flaw detection source device, and the application range of a product is widened. The target material is defined as a natural target material and a high-enrichment-degree target material according to the difference of the content of the iridium 191 in the target material, the target material arrangement can be automatically adjusted according to the actual production requirement, and the production flexibility of the iridium 192 can be improved.

The invention relates to the technical field of research of reactor fuel irradiation, in particular to a fuel power calibration method for a loop irradiation test, which comprises the following steps: carrying out an out-of-reactor hydraulic test, and establishing a high-precision CFD (computational fluid dynamics) model of a fuel loop irradiation device based on a hydraulic test result; calculating the gamma heat release rate ratio of each material and the total heat release in the test section, and preliminarily determining the nuclear power value of the fuel assembly; based on the heat conduction differential equation and the convective heat transfer equation, the gamma heat release rate of the material and the nuclear power value of the fuel assembly are corrected; repeating the calculation process of S3 until the nuclear power value of the fuel assembly converges, and obtaining the final theoretical calculation value of the gamma heat release rate of each material and the nuclear power of the fuel assembly; and further verifying the calculation result to finally obtain the nuclear power of the fuel core body, and obtaining the burnup of the fuel assembly in the irradiation test through time integration. The method aims to solve the technical problem that the nuclear power of the fuel assembly calculated based on thermotechnical measurement data in a loop irradiation test has deviation compared with an actual value.

The invention relates to the technical field of research reactor material irradiation, and provides a variable thermal resistance structure suitable for material irradiation temperature adjustment and a design method, the variable thermal resistance structure comprises a protection tube and a to-be-tested sample arranged in the protection tube, an elastic thermal resistor is arranged between the protection tube and the to-be-tested sample, and the elastic thermal resistor is in contact with the protection tube and the to-be-tested sample at the same time; the cross section of the elastic thermal resistor is of an annular structure formed by continuous bending, inert gas is further filled between the protection tube and the sample to be tested, the elastic thermal resistor is in contact with the outer wall of the sample to be tested and the inner wall of the protection tube at the same time, heat transfer of an original air gap layer is changed, and heat transfer performance is enhanced; meanwhile, the elastic thermal resistor formed by bending has certain ductility, so that the outward heat transfer capacity of the sample to be tested in the irradiation test is enhanced, the influence of thermal expansion on heat transfer is reduced, and the irradiation test requirements of various materials are met; parameter determination is carried out through an out-of-pile verification test, and in-pile condition tests are combined, so that the elastic thermal resistance meeting the requirements is rapidly screened out.

The invention discloses a reactor core outlet temperature measuring device for a research reactor and a replacement method thereof, and belongs to the technical field of research reactor structure design. The device comprises a measuring plate and at least one group of supporting and guiding parts arranged on the measuring plate, the measuring plate is arranged at the lower part of the reactor internals, and a plurality of thermocouple thermometers are arranged on the plate surface of the measuring plate, so that the temperature of any position of a reactor core outlet can be conveniently measured; the supporting and guiding component comprises a thermocouple protection tube, a radial limiting mechanism and a thermocouple sealing assembly. According to the reactor core outlet temperature measuring device for the research reactor, the thermocouple thermometer can be integrally replaced, the structure is convenient to assemble and disassemble, and when the thermocouple thermometer breaks down and needs to be replaced, the reactor core outlet temperature measuring device and the thermocouple thermometer on the measuring plate can be integrally disassembled; the thermocouple and the thermometer can be replaced outside a research reactor, so that the temperature measurement requirement of a reactor core outlet can be continuously met.

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 present invention discloses a burnup tracking calculation method, system and medium for multi-cycle fuel management; it relates to the technical field of reactor burnup calculation; after the first-cycle core modeling, geometric grid division is performed on each burnup zone unit in the full-core model. During the cycle of the full-core model, the replacement of all reactor internal components in the full-core model is achieved by exchanging the positions of the corresponding spatial numbers in the repeated geometric grid. Combining with the multi-cycle burnup tracking calculation process, the dynamic changes of fuel burnup, control rods and beryllium followers during reactor operation and their coupling relationship with the particle transport process are accurately considered, and the simulation of the actual refueling process of the research reactor based on the Monte Carlo method is completed, realizing high-precision and high-efficiency core calculation and analysis for high-flux research reactors, and providing theoretical support for aspects such as high-quality testing of nuclear fuels and materials, optimization of core loading and irradiation resources of research reactors, and improvement of the operation safety of research reactor reactors.

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 invention relates to the technical field of research reactor material irradiation, in particular to a normal-pressure pore channel irradiation device with an autonomous rotation function, which comprises a far-end terminal, and a positioning section, a protection tube and a test section which are connected in sequence, the positioning section is configured to be matched with a normal-pressure pore channel opening, and the protection tube is configured to be matched with the normal-pressure pore channel opening; comprising a transmission shaft, a limiting ring assembly, a power transmission assembly, a servo motor and a controller, the transmission shaft is sleeved with the limiting ring assembly, the servo motor is installed in the side direction of the transmission shaft, the controller is electrically connected with the servo motor, and a far-end terminal is in communication connection with the controller. The power transmission assembly comprises a transmission shaft gear and a motor gear meshed with the transmission shaft gear, the transmission shaft gear is in coaxial key connection with the transmission shaft, the motor gear is in coaxial key connection with an output shaft of the servo motor, and lead pipelines are arranged in the axial direction of the protection pipe and the transmission shaft. A servo motor remote control mode is adopted, the automatic rotation function of the irradiation device bearing the tested piece can be achieved, and the accuracy and reliability of a test result are improved.

The invention discloses an aluminum alloy for simulating a neutron irradiation transmutation silicon effect and a manufacturing method thereof, and belongs to the field of aluminum alloys. The manufacturing method of the aluminum alloy for simulating the neutron irradiation transmutation silicon effect comprises the following steps: determining the Si element content in the alloy according to the state of the aluminum alloy to be simulated, and preparing a supersaturated solid solution raw material with corresponding components, an aluminum alloy bar is prepared through a powder metallurgy process, and an aluminum alloy structure for simulating the generation of the transmutation silicon effect is obtained through heat treatment. According to the aluminum alloy preparation method, the simulated alloy can be prepared to effectively simulate a precipitated phase structure formed by transmutation reaction after the aluminum alloy is irradiated by neutrons for a long time, and the aluminum alloy preparation method plays an important role in researching the influence of irradiation transmutation products on the performance of the aluminum alloy; the current situation that irradiation performance evaluation of a metal material with a solid transmutation product is extremely dependent on neutron irradiation resources is relieved, the research efficiency is improved, and therefore the design level and the safety performance of an aluminum alloy structural part in a research reactor are improved.

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 reactor structures, in particular to a reactor bottom structure for a research reactor and an installation method.The reactor bottom structure is installed in a reactor lower cell and comprises an anti-seismic supporting structure fixed to the bottom face of the reactor lower cell and a blow-off pipeline system fixed to the anti-seismic supporting structure; the ventilation cooling structure is fixed on the top surface of the under-reactor chamber; the control rod driving mechanism is inversely mounted; a stator coil in the control rod driving mechanism is arranged in a cylinder in the ventilation cooling structure, a pressure-resistant shell in the control rod driving mechanism is arranged in the anti-seismic supporting structure, and an anti-seismic plate in the anti-seismic supporting structure is fixed to the pressure-resistant shell. A drainage connecting pipe arranged at the bottom end of the control rod driving mechanism is connected with the blow-off pipeline system. The reactor bottom structure has a compact structural size, provides limiting and cooling for the CRDM, and is also provided with a blow-off pipeline system for regularly discharging radioactive wastes to reduce the irradiation dose of the under-reactor cell.

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 nuclear engineering, in particular to a beryllium sleeve, a reactor core assembly and a research reactor irradiation test device. The beryllium sleeve comprises a supporting piece, a beryllium block, a first fixing block and a second fixing block. The supporting piece is provided with a reactor core channel suitable for neutrons to pass through; a plurality of beryllium blocks are attached to the supporting piece, the beryllium blocks are adjacently arranged in the circumferential direction of the reactor core channel so as to surround the reactor core channel in the circumferential direction, the beryllium blocks are adjacently arranged in the extending direction of the reactor core channel so as to cover the reactor core channel in the axial direction, and aluminum pieces are arranged between the adjacent beryllium blocks in a matched mode; the first fixing block is matched with the supporting piece and the beryllium block; the second fixing block is matched with the supporting piece and the beryllium block. The beryllium sleeve is high in structural stability and reliability, neutrons can be effectively reflected to slow down quantity attenuation of the neutrons when the beryllium sleeve is combined with a reactor core assembly to be applied to a research reactor irradiation test device, neutron adsorption is reduced in cooperation with an aluminum piece, and neutron economy is improved.

The invention provides a cooling water system for a research reactor irradiation test loop and a design and use method thereof, relates to the technical field of irradiation in nuclear fuel and material reactors, and solves the technical problem of potential safety hazards caused by complete cutoff of an external water source of an existing cooling water system. A corresponding volume between a first liquid level alarm point and a second liquid level alarm point is an operation buffer volume V1, a volume between the second liquid level alarm point and a first water taking point is a minimum cooling water volume V2 at the initial stage of reactor shutdown, and a volume between the first water taking point and a second water taking point is a minimum cooling water volume V3 at the later stage of reactor shutdown; the sum of V1, V2 and V3 is the optimal volume V of the cooling water tank. According to the invention, the complexity of the system is simplified, the space occupied by equipment is reduced, the construction cost is reduced, and the reliability of the equipment and the system is ensured through the nuclear security level; the safety of the fuel element is ensured, the size design of the cooling water tank is optimized, and the space between research reactor processes is further reasonably utilized.

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 invention relates to a fine power calculation method and device for researching a small irradiation component in a reactor and a readable storage medium. The fine power calculation method comprises the following steps: S1, acquiring parameter information of a structural component in a reactor core; s2, dividing the structural component into a plurality of homogenized grids and dividing a neutron transport calculation area; s3, calculating neutron fluence rate and power in each region; s4, calculating a power weight factor of each fuel rod; s5, judging whether all state calculation of the structural component is completed or not, and if yes, entering the step S7; s6, changing the state parameter information of the structure assembly, and calculating the neutron fluence rate and power, the homogenization cross section, the homogenization grid geometric information and the power weight factor; s7, calculating the neutron fluence rate and power of the three-dimensional fine net of the reactor core; and S8, calculating the three-dimensional power of each fuel rod of the irradiation small component. According to the method, the calculation resolution of the irradiation small component can be improved, and the safety analysis capability of the irradiation small component 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 invention relates to the technical field of nuclear engineering, in particular to a beryllium sleeve, a reactor core assembly and a research reactor irradiation test device. The beryllium sleeve comprises a supporting piece, a beryllium block, a first fixing block and a second fixing block. The supporting piece is provided with a reactor core channel suitable for neutrons to pass through; the beryllium blocks are arranged on the supporting pieces in a sleeving manner, and the multiple beryllium blocks are adjacently arranged in the extending direction of the reactor core channel; the first fixing block is matched with the supporting piece and matched with the beryllium block. The first fixing block is matched with the supporting piece, the second fixing block is matched with the supporting piece and the beryllium block, and the first fixing block and the second fixing block are located on the two sides of the beryllium block respectively in the extending direction of the reactor core channel. The beryllium sleeve provided by the invention is strong in structural stability and high in reliability, and can effectively reflect neutrons to slow down quantity attenuation of the neutrons when being applied to a research reactor irradiation test device in combination with a reactor core assembly, so that the neutron economy is improved.

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 discloses a method for establishing natural circulation of a pool type research reactor and belongs to the field of nuclear power. The method comprises the following steps: under an accident condition, when a pump set of the pool type research reactor stops working, injecting pressure fluid driven in a passive form into a reactor bottom chamber of the pool type research reactor, driving a coolant in a reactor core to flow from the bottom to the top, and establishing natural circulation of the coolant. According to the method, the flowing direction of the coolant in the reactor core can be quickly reversed, coolant stagnation is avoided, the overheating risk of the reactor core is reduced, and the safety of the pool type research reactor is improved.