63 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 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 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 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 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.

The invention relates to the technical field of nuclear reactors, and provides an online measurement method and device for heat capacity change of a reactor pool of a pool shell type research reactor and a medium, and the method comprises the following steps: acquiring environmental parameters and operating parameters of the reactor pool of the pool shell type research reactor; based on a pool surface evaporation heat dissipation power model, the environment parameters and the operation parameters, determining pool surface evaporation heat dissipation power of the pool shell type research reactor pool; carrying out water area division on the reactor pool of the pool shell type research reactor based on the environmental parameters, and calculating energy corresponding to each water area; determining the energy loss power of the reactor pool of the pool-shell research reactor based on the energy corresponding to each water area; and determining the heat dissipation power of the reactor pool of the pool shell type research reactor based on the surface evaporation heat dissipation power and the energy loss power of the pool. According to the embodiment, the measurement accuracy of the thermal power of the reactor is improved, the difference value between the actual thermal power value and the nuclear power value of the reactor is reduced, the safety and stability of the nuclear reactor are effectively improved, and the temperature control precision of the reactor in the operation process is ensured.

An aluminum alloy simulating the effect of neutron irradiation-induced silicon transmutation and its manufacturing method belong to the field of aluminum alloys. Among them, the manufacturing method of the aluminum alloy simulating the effect of neutron irradiation-induced silicon transmutation includes the following steps: determining the Si element content in the alloy according to the aluminum alloy state to be simulated and preparing a supersaturated solid solution raw material with the corresponding composition, using powder metallurgy technology to obtain an aluminum alloy bar and obtaining an aluminum alloy structure simulating the generation of silicon transmutation effect through heat treatment. This aluminum alloy preparation method can prepare a simulated alloy to effectively simulate the precipitation phase structure formed by the transmutation reaction after the aluminum alloy is irradiated by neutrons for a long time, which plays an important role in studying the influence of irradiation transmutation products on the properties of aluminum alloys, alleviates the current situation that the irradiation performance evaluation of metal materials with solid transmutation products highly depends on neutron irradiation resources, improves the research efficiency, and thus improves the design level and safety performance of aluminum alloy structural parts in research reactors.

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.

This application belongs to the technical field of research reactor irradiation, and specifically relates to a radial mobile fuel transient irradiation device applicable to research reactors. It includes an irradiation container, an in-pile irradiation device, a moving mechanism, and a limiting mechanism; the irradiation container has an opening; one end of the in-pile irradiation device is located inside the irradiation container, and the other end is located outside the irradiation container; the moving mechanism is arranged on the irradiation container and connected to the in-pile irradiation device to drive the in-pile irradiation device to move in the opening in a preset direction; the limiting mechanism is arranged inside the irradiation container and is movably matched with the in-pile irradiation device, and the in-pile irradiation device has a moving direction along the preset direction relative to the limiting mechanism. This application can achieve the radial displacement of the in-pile irradiation device without relying on the environment of the pool-type research reactor, so that it can be applicable to the fuel transient irradiation test of pool-type / non-pool-type research reactors.

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 present invention relates to the field of research reactor material irradiation technology, and specifically to a normal-pressure channel irradiation device with autonomous rotation capability. The device comprises a remote terminal and a sequentially connected positioning section, a protective tube, and a test section. The positioning section is configured to mate with the normal-pressure channel opening, and comprises a drive shaft, a limit ring assembly, a power transmission assembly, a servo motor, and a controller. The limit ring assembly is sleeved on the drive shaft, the servo motor is mounted laterally to the drive shaft, the controller is electrically connected to the servo motor, the remote terminal is communicatively connected to the controller, the power transmission assembly comprises a drive shaft gear and a motor gear meshing with the drive shaft gear, the drive shaft gear is coaxially keyed to the drive shaft, the motor gear is coaxially keyed to the output shaft of the servo motor, and lead pipes are provided axially along the protective tube and the drive shaft. The present invention utilizes a servo motor for remote control to achieve autonomous rotation of the irradiation device carrying the test piece, thereby improving the accuracy and reliability of the test results.

The invention provides an ice plug test system and method for a control rod drive wire of an isotope production research reactor, the test system comprises a mounting bracket and an ice plug forming mechanism, and a target test pipeline is mounted on the mounting bracket; one side of the mounting bracket is provided with a pressure pump and a water tank, one end of the pressure pump is connected with the water tank through a connecting pipeline, and the other end is communicated with a target test pipeline; the ice plug forming mechanism comprises a liquid nitrogen tank and an ice plug jacket assembly, the liquid nitrogen tank is connected with the ice plug jacket assembly through a conveying pipeline, the ice plug jacket assembly comprises a first splicing piece and a second splicing piece, and the first splicing piece and the second splicing piece are detachably clamped on the outer side of a target test pipeline; a cooling cavity is formed between the ice plug jacket assembly and the target test pipeline, and liquid nitrogen conveyed by the conveying pipeline flows through the cooling cavity to form an ice plug in the target test pipeline; the connecting pipeline is provided with a flowmeter, and the flowmeter is used for detecting the flow of water flowing out of the target test pipeline so as to determine the ice plug forming state of the target test pipeline.

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.