72 results about "Nuclear reactor core" patented technology

The application relates to a neutron absorption material and application thereof to a control rod for a nuclear reactor core. The neutron absorption material provided by the application is prepared by mixing ytterbium elements and holmium elements in a certain proportion, can effectively play the high neutron capture cross-section characteristics of the ytterbium elements and the holmium elements, effectively absorb thermal neutrons, and can ensure good chemical and thermal stability, so that the neutron absorption material can maintain its structure and performance stability for a long time in a high irradiation environment, effectively avoids the conversion of the neutron absorption material into a weak neutron absorption material after absorbing neutrons, and further controls the neutron absorption characteristics of the neutron absorption material. Further, the neutron absorption material provided by the application can still ensure the neutron absorption characteristics thereof and maintain the reactivity value under high burnup depth, so as to prolong the service life of the control rod for the nuclear reactor core.

A nuclear reactor core includes a plurality of fuel elements and moderator blocks to form a nuclear reactor core inner portion and a nuclear reactor core outer portion. The nuclear reactor core inner portion includes an inner moderator matrix formed of a plurality of inner holes that include a plurality of inner fuel openings with one or more fuel elements disposed therein. The plurality of inner holes further include a plurality of inner coolant passages to flow a coolant. The nuclear reactor core outer portion includes an outer moderator matrix formed of a plurality of outer holes that include a plurality of outer fuel openings with one or more fuel elements disposed therein. The plurality of outer holes further include a plurality of outer coolant passages to flow the coolant. The outer coolant passage diameter exceeds the inner coolant passage diameter.

The application discloses a kind of structural grid methods and systems of dividing triangular arrangement rod bundle assembly with wire winding, it is related to nuclear reactor core thermal-hydraulic analysis and calculation field, by using structured hexahedral grid to divide the flow channel of triangular arrangement rod bundle with wire winding, grid quality is higher, and calculation efficiency is more optimal;When establishing the structured grid of the flow cross section of triangular arrangement rod bundle assembly with wire winding, we fully consider the geometric characteristics of wire winding, thereby improve the geometric reduction degree of grid model;According to the calculation requirement, the number and distribution of grid nodes can be flexibly adjusted, and the axial length is shortened accordingly, so as to not affect the calculation precision, reduce the number of grids, save the calculation resource, and ensure the calculation quality.

This invention discloses a method, system, and medium for data mapping between a solid geometric model and an unstructured mesh, relating to the field of nuclear reactor core coupling programming technology. By converting the CSG geometric entity into a coupling intermediate mesh (unstructured mesh), and using the coupling intermediate mesh as an intermediary between the CSG geometric entity and other professional unstructured meshes, high-precision coupled data transmission and coupled calculation between the CSG geometric entity and other professional unstructured meshes can be achieved through mesh mapping and other methods, based on the geometric correspondence between the CSG geometric entity and the coupling intermediate mesh.

The application provides a nuclear reactor core device and a preparation method thereof, and belongs to the technical field of material welding. The preparation method of the nuclear reactor core device comprises the following steps: preparing a rib plate assembly, a W-shaped plate assembly and a C-shaped plate assembly; assembling multiple W-shaped plate assemblies and C-shaped plate assemblies into a whole circle and fixing the same by using inner and outer support tools, performing automatic welding by using a laser symmetrical welding method, and forming a W / C-shaped plate assembly; synchronously and alternately welding the W / C-shaped plate assembly on the inner and outer sides by using two automatic wire filling welding guns by using a double-sided layered symmetrical welding method, and realizing the connection of the W / C-shaped plate assembly with a bottom plate and a top plate; performing automatic butt welding of an outer half-ring plate by using laser welding and laser wire filling welding; and finally, welding internal fillet welds and performing nondestructive testing. By the cooperation of laser welding, automatic MIG welding, special tools and a symmetrical welding sequence, high-quality and high-efficiency automatic welding of the core containment vessel is realized, and welding deformation is effectively controlled.

The reactor core fuel assembly transfer device comprises the reactor core fuel assembly transfer trolley, the reactor core fuel assembly transfer trolley comprises a trolley body, walking wheels and guide wheels, the walking wheels and the guide wheels are arranged on the trolley body, and the trolley body comprises a bottom plate, a rear side wall, a front side wall, a left side wall and a right side wall. The vehicle body is connected with a driving mechanism, and one end, far away from the driving mechanism, of the bottom plate is provided with a hole for a loader to turn over and pass through; the travelling wheels and the guide wheels are respectively mounted on the left side wall and the right side wall; the reactor core fuel assembly transfer trolley realizes cross-rail travelling along the transfer rail through the travelling wheels and the guide wheels; the left side wall and the right side wall of the vehicle body are concentrically provided with hinge holes used for being connected with a loader respectively. No power alternation, no clamping stagnation and no power failure risk exist in the whole working process, and the transfer efficiency and safety are improved.

The invention belongs to the nuclear industry and can be used for the production of fuel pellets from uranium-molybdenum metal powder enriched with uranium-235 to 7%, which can be used as fuel elements for nuclear reactors. The sintering of the pellets is carried out in an inert argon atmosphere at a temperature in the range of 1100°C to 1155°C, as the initial powder, a particle size of 160 μm is used, with a molybdenum content of 9.0 to 10.5 wt.%. The powder is preheated at a temperature of 500°C for 10-20 hours (in an argon medium), and then cold-pressed into pellets in a mold at a pressure of up to 950 MPa. According to another method, when preparing uranium-molybdenum pellets using a binder (plasticizer), the pellets are heated in argon from 300°C to 450°C for a duration of 2-4 hours before sintering, in order to remove the binder. The invention allows to increase the uranium content in the fuel, to reduce the heat accumulated in the core of the nuclear reactor, and to reduce the energy release in the event of a violation of the normal operating conditions of the nuclear reactor, which will improve its safety and emergency stability.

This application discloses a core power prediction method and apparatus, a core monitoring method and system, relating to the field of nuclear industry technology. The core power prediction method includes: acquiring target power detector data for a target core; the target power detector data includes values ​​from multiple power detectors; processing the target power detector data using a target power prediction model to obtain a predicted power value for the target core; wherein the target power prediction model employs a data grouping processing method, trained using historical power detector training data and their corresponding first historical power values. According to embodiments of this application, this core prediction method can monitor core power using a small number of detectors while ensuring the accuracy of core power prediction.

The invention relates to the technical field of nuclear reactor cores, and provides a reactor core adopting different hydrides for combined application, the reactor core comprises a plurality of fuel assemblies which are sequentially arranged in a zoning manner along the radial direction from the center to the periphery; part of the fuel assemblies simultaneously comprise uranium zirconium hydride fuel rods filled with uranium zirconium hydride fuel and uranium yttrium hydride fuel rods filled with uranium yttrium hydride fuel. Metal hydride fuel rods are introduced into the moderator-free reactor core, the reactor core neutron energy spectrum is slowed down, the reactor core fission reaction rate is increased, the reactor core critical mass is reduced, meanwhile, the equivalent nuclear density of uranium and hydrogen in different areas of the reactor core is adjusted in the mode that different hydride types are used in a combined mode, and reactor core power distribution is further flattened.

The application provides a fuel and a method for increasing the neutron source intensity of a nuclear reactor, wherein the abundance of O in the fuel is 1% to 50% 18 O is 1% to 50%, and alpha particles and the enriched 18 O can significantly enhance the initial neutron source intensity of a nuclear reactor core in a pressurized water nuclear power plant, and help to improve the start-up performance and reduce the dependence on external neutron sources. In the passive start-up mode without using an external neutron source, the application improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant. 18 O in the fuel, and improves the isotope abundance of O in the fuel, and improves the (alpha, n) reaction neutron source in the nuclear reactor core in the pressurized water nuclear power plant.

The invention belongs to the technical field of nuclear reactor cores, and particularly relates to a fast thermal coupling metal hydride reactor core controlled by a rotating drum and a rotating drum control method, and the reactor core comprises the rotating drum, a reactor core active area, a coolant flow channel, a beryllium oxide reflecting layer and a thermal insulation shielding layer; the coolant flow channel is composed of a coolant flow channel inner barrel and a coolant flow channel outer barrel; the reactor core active area is wrapped by the coolant flow channel inner barrel, and oxide fuel rods and metal hydride fuel rods are arranged in the reactor core active area; the beryllium oxide reflecting layer is arranged on the coolant flow channel outer barrel in a sleeving manner, and the heat preservation shielding layer is arranged on the beryllium oxide reflecting layer in a sleeving manner; a plurality of rotating drums are arranged between the beryllium oxide reflecting layer and the heat preservation shielding layer in the circumferential direction. The oxide fuel rods and the metal hydride fuel rods are arranged in the reactor core active area at the same time, so that the fast thermal coupling design of a reactor core neutron energy spectrum is realized, and the reactivity control value of the rotating drum is improved.

A metal hydride fuel burn-up self-sustaining reactor core controlled by burnable poison relates to the technical field of nuclear reactor cores, and comprises a low-enrichment-degree control rod-free fuel assembly located in the center of the reactor core, and a middle-enrichment-degree control rod-free fuel assembly arranged around the periphery of the low-enrichment-degree control rod-free fuel assembly. A middle-enrichment-degree safety rod-containing fuel assembly is arranged between the fuel tank and the fuel tank; a high-enrichment-degree control rod-free fuel assembly is arranged around the periphery of the medium-enrichment-degree control rod-free fuel assembly, and a medium-enrichment-degree compensation adjusting rod-containing fuel assembly is arranged between the medium-enrichment-degree control rod-free fuel assembly and the high-enrichment-degree control rod-free The periphery of the high-enrichment-degree control rod-free fuel assembly is filled with lead bismuth; the reactor core is used for solving the problem that a traditional reactor core is difficult to reach expected burnup depth and even causes short service life.

The utility model provides a control rod assembly, and belongs to the technical field of nuclear reactor core control rod assemblys.The control rod assembly comprises a cladding, a first end plug, a second end plug and an absorber, the first end plug and the second end plug block openings in the two ends of the cladding respectively, and the absorber is limited in the cladding through the first end plug and the second end plug; the variable-diameter section is arranged on the cladding, so that the wall thickness of the variable-diameter section is increased from the first end to the second end, the wear resistance of the cladding is improved under the condition that an external structure is not increased, and the control rod cladding is prevented from continuously abrading with a reactor internal component guide structure under the action of flow-induced vibration when the control rod stays at a fixed axial position for a long time; therefore, the control rod is prevented from wear-through damage and the structural integrity of the control rod is ensured.

The invention discloses a nuclear reactor neutron fluence determination method and device, and relates to the technical field of nuclear reactor core design and safety. Comprising the following steps: acquiring a first neutron fluence rate of a first position in an axial active section of a fuel assembly in a nuclear reactor; acquiring a correction coefficient of a second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient comprises at least one of a space correction coefficient and an energy group correction coefficient, the space correction coefficient is used for correcting attenuation of the neutron fluence rate from the first position to the second position, and the energy group correction coefficient is used for correcting energy group structure deviation between a first energy group structure of the first neutron fluence rate and a second energy group structure of the second position; based on the correction coefficient, correcting the first neutron fluence rate to obtain a second neutron fluence rate of the second position; and determining the neutron fluence of the second position based on the second neutron fluence rate of the second position. According to the invention, the accuracy of neutron fluence prediction outside the axial active section of the fuel assembly can be improved.

The invention relates to an experimental device capable of being used for researching thermal hydraulic characteristics of a reactor core rod bundle of a nuclear reactor, and relates to the technical field of reactors. The experimental device capable of being used for researching the thermal hydraulic characteristics of the reactor core rod bundle of the nuclear reactor comprises a pressure-bearing shell, a lower sealing head, a measuring assembly, a flow channel, a simulation rod array and an upper sealing head, wherein a containing cavity is formed in the pressure-bearing shell; the lower sealing head is mounted at the lower end of the pressure-bearing shell; the measuring assembly is mounted at one end of the pressure-bearing shell away from the lower sealing head; the measuring assembly comprises a mounting flange and a screen module, the mounting flange is connected with the pressure-bearing shell, and the screen module is detachably mounted on the side, away from the pressure-bearing shell, of the mounting flange; the runner is positioned in the accommodating cavity, and one end is connected with the mounting flange; the simulation rod array comprises at least one simulation rod, one end of each simulation rod is fixed to the measuring assembly, and the other end of each simulation rod penetrates through the flow channel. The upper sealing head is installed on the side, away from the pressure-bearing shell, of the measuring assembly. The experimental device can meet the requirements of heat transfer characteristic experiments after water level swelling and drying in the rod bundle at the same time.

The application discloses a reactor neutron detector calibration method, system and computer equipment, relates to the nuclear reactor core monitoring technical field, and is based on the absolute neutron flux at a position of a reactor core measured by an activation method, combines the neutron flux distribution relation inside and outside the reactor, maps the absolute neutron flux at the position of the neutron detector outside the reactor, obtains the corresponding relation between the neutron detector measurement result and the absolute neutron flux at the neutron detector, and based on the corresponding relation and the neutron detector measurement result, the absolute neutron flux at the detector can be quickly and accurately obtained. The application only uses the activation method to calibrate the conversion factor of the neutron detector once, and under various conditions such as changes in the core arrangement, changes in the core control rod state, changes in the core power and the like, the absolute neutron flux at the detector can be directly obtained through the measurement result of the neutron detector, and multiple irradiation experiments are not needed, so that the measurement efficiency is improved, and the cost is saved.

Provided in the present application are a method and apparatus for determining a nuclear fuel loading and unloading mode, and a device and a storage medium. The method for determining a nuclear fuel loading and unloading mode comprises: acquiring a unit number of an adjacent core unit corresponding to a target core unit, the unit number indicating a positional relationship between the adjacent core unit and the target core unit; on the basis of the unit number, acquiring a nuclear fuel storage state of the adjacent core unit; on the basis of the nuclear fuel storage state and the unit number, determining a target core offset mode and a target loading auxiliary tool guiding mode for the target core unit when same performs nuclear fuel loading and unloading; and on the basis of the target core offset mode and the target loading auxiliary tool guiding mode, determining a target loading and unloading mode corresponding to the target core unit. By means of the embodiments of the present application, the efficiency and accuracy of determining a nuclear fuel loading and unloading mode for a nuclear reactor core can be improved.

A fast neutron nuclear reactor, in particular liquid-metal cooled, with a reactivity control system arranged on the periphery of the nuclear core. The invention relates to a fast neutron nuclear reactor (1) comprising: - a reactor vessel (10) filled with a coolant, - a reactor core (11) consisting of fuel assemblies (13), - a reactivity control system (2), on the periphery of the core, in a cold manifold zone, between the reactor vessel and the separation step between the hot and cold manifolds, such that the reactivity control assemblies or monoblock rods are traversed by the cold coolant. Figure for the abstract: Fig. 2

A multi-use incore system for use with a nuclear reactor having a reactor core including an instrumentation tube is disclosed. The multi-use incore system comprises a drive cable, a neutron detector, an electromagnetic tool, and an irradiation target. The drive cable comprises an adapter at its distal end. The drive cable is slidably receivable into the instrumentation tube. The neutron detector is removably attachable to the adapter of the drive cable to monitor core power distribution. The neutron detector is slidably receivable into the instrumentation tube by the drive cable. The electromagnetic tool is removably attachable to the adapter of the drive cable in lieu of the neutron detector. The irradiation target is to produce a radioisotope. The drive cable is selectively attachable to the irradiation target by the electromagnetic tool to insert the irradiation target into the instrumentation tube and remove the irradiation target from the instrumentation tube.

The invention relates to the technical field of nuclear reactor cores, and particularly provides a reactor core with mixed arrangement of metal hydride fuel and metal fuel, which comprises a plurality of fuel assemblies, any fuel assembly comprises a metal hydride fuel rod loaded with metal hydride fuel and a metal fuel rod loaded with metal fuel; and in any one of the fuel assemblies, the proportion of the metal hydride fuel rods is 0-100%. The high uranium density and high proliferation characteristics of metal fuel and the hydrogen atom moderation characteristic of metal hydride fuel are utilized, the metal fuel and the metal hydride fuel are arranged in a mixed mode, meanwhile, the proportion of the metal hydride fuel and the metal fuel is adjusted according to the reactor core power life period and other design requirements, and customized adjustment of the reactor core energy spectrum is achieved; and the performance of the reactor core is optimized to meet the requirements of different application scenes.

The invention relates to the technical field of nuclear reactor cores, in particular to a control rod assembly and a method for prolonging the service life of the control rod assembly. The control rod assembly comprises a connecting handle, a plurality of control rods and an anti-loosening pad, the tail end of the connecting handle is detachably connected with a plurality of control rods; an anti-loosening pad is arranged at the joint of the tail end of the connecting handle and the control rod; the connecting handle and the control rod extrude the anti-loosening pad to realize loosening prevention; a plurality of finger rods are arranged at the tail end of the connecting handle, each finger rod is provided with an internal thread, and the lower end of each finger rod is provided with two grooves close to the right; and the finger rod is in threaded connection with the control rod. The method for prolonging the service life of the control rod assembly comprises the following steps: assembling the control rod assembly with the structure in the technical scheme; after the control rod assembly is applied to a reactor, in-service inspection is carried out on the control rod assembly in the using process, and when it is found that an individual control rod is abraded and swelled, the control rod which cannot be used is replaced in time. The control rod assembly adopts a split design, so that the control rod is independently replaced, and the overall service life of the control rod assembly is prolonged.

A nuclear reactor is designed and constructed to withstand seismic events. A mechanical hold down and discriminator may be provided as cooperating structure between the core receptacle and a nozzle of a core assembly. One or more receivers may be formed on the nozzle that cooperate with corresponding protrusions in a core receptacle. As the nozzle is inserted into the receptacle, the protrusion engages with the receiver to inhibit unintentional withdrawal of the nozzle from the receiver. A first pattern of receivers on a nozzle may align with a first pattern of protrusions within a receptacle to enable the nozzle to fit within a first receptacle. A second pattern of receivers formed on a second nozzle may inhibit the second nozzle from being inserted into the first receptacle and thereby provide a discrimination feature to ensure proper core assemblies are located in the correct location within a nuclear reactor core.

This invention discloses a method and apparatus for simulating and analyzing the dynamic loads of nuclear reactor decommissioning and dismantling, comprising: constructing a refined finite element model of the reactor body; developing a custom subroutine for loading moving dynamic loads; setting and calculating multi-dimensional simulation conditions; verifying structural safety thresholds and identifying weak points; and optimizing and analyzing operational parameters. The technical solution of this invention solves the problems of existing technologies being unable to accurately simulate the mechanical response of moving dynamic loads during reactor dismantling, unable to comprehensively cover all operational conditions, and unable to provide accurate safety control parameters for dismantling operations.

The invention discloses a method, a module, a numerical simulation platform and equipment for realizing finite volume calculation, relates to the technical field of nuclear reactor core or CFD numerical program design, and solves the problem that finite element calculation and finite volume calculation of the existing nuclear reactor numerical simulation platform are incompatible. According to the technical scheme, the method is characterized in that a 0-order Monomial unit is introduced on the basis of a polynomial Monomial finite element unit, so that the order of a polynomial of the polynomial Monomial finite element unit is reduced to 0 order; and setting the variable discrete type as a 0-order Monomial unit so as to realize switching from finite element calculation to finite volume calculation on a numerical simulation platform. According to the method, expansion is carried out on the basis of the finite element Monomial polynomial unit, and the 0-order Monomial unit type is increased, so that finite volume calculation is supported, and a partial differential equation solving framework of a nuclear reactor numerical simulation platform can support the same interface module to realize finite element and finite volume calculation at the same time.

The invention belongs to the technical field of nuclear power, and particularly relates to a nuclear reactor core fuel assembly interference identification method and device, a computer readable storage medium and electronic equipment. The method includes: determining a center-to-bottom distance of a first component of a first fuel assembly of a nuclear reactor core and a center-to-bottom distance of a second component of a second fuel assembly; wherein the first fuel assembly is a fuel assembly which is being loaded into or unloaded from a reactor core of a nuclear reactor, the second fuel assembly is a fuel assembly adjacent to the first fuel assembly, and the distance from the center of the component to the bottom of the fuel assembly is the distance from the center of the component to the bottom of the fuel assembly; determining the lifting height of the first fuel assembly; according to the center bottom distance of the first component, the center bottom distance of the second component and the lifting height of the first fuel assembly, the center distance of the first component and the second component in the vertical direction is determined; and determining an interference identification result between the first component and the second component according to the center distance between the first component and the second component in the vertical direction.

The present invention relates to a method for installing a temporary neutron measurement system (TNMS) mounted on a core of an integrated nuclear reactor, comprising the steps of: (S10) arranging, inside a reactor core, a measurement assembly capable of measuring neutrons discharged from the reactor core; and (S20) connecting, to the measurement assembly, a connection assembly for supplying power to the measurement assembly and receiving a measurement signal.

The invention discloses a two-phase thermal hydraulic analysis method for a reactor core of a nuclear reactor, which comprises the following seven steps of: 1, determining sub-channel modeling parameters according to a sub-channel structure of the reactor core, establishing a sub-channel three-dimensional geometric model, and generating a computational grid of the three-dimensional geometric model by utilizing grid division software; 2, establishing a basic control equation of a vapor phase and a liquid phase, and establishing a constitutive model of the closed basic control equation; 3, establishing a fuel rod heat conduction model; 4, discretizing the fluid and solid calculation models in the steps 2 and 3 based on a finite volume discretization method; and 5, establishing a fluid and solid coupling solving process. According to the method, the calculation efficiency is considered while the channel three-dimensional geometry and the main flow characteristic resolution capability are kept, a novel calculation analysis method is provided for efficient simulation of nuclear reactor core two-phase thermal hydraulic characteristic analysis, and a technical support is provided for a high-fidelity numerical reactor simulation analysis system.

A heat exchanger for cooling a nuclear reactor core is disclosed herein. The heat exchanger can include a first stage including an input configured to receive a working fluid from an external source into the heat exchanger, and a first plenum configured to envelope a moderator heat pipe extending from the nuclear reactor core. The heat exchanger can further include a second stage including an output configured to remove a working fluid from the heat exchanger to the external source, and a second plenum configured to envelope a power heat pipe extending from the nuclear reactor core, wherein the first plenum and the second plenum are in fluid communication and configured such that the external fluid must traverse the first plenum and over the moderator heat pipe before entering the second plenum and traversing over the power heat pipe.

The invention belongs to the technical field of nuclear reactor cores, and particularly relates to a modular multipurpose lead-bismuth fast reactor core and a design method. The method comprises the following steps: step 1, flattening reactor core power distribution to determine an enrichment degree gradient and partitions; 2, designing structures of a fuel assembly, a safety rod assembly, an adjusting rod assembly and a compensation rod assembly and a control rod lifting strategy; step 3, determining radial and axial sizes and power density of the reactor core; step 4, reactor core reactivity control design; 5, designing the value of the single-beam control rod; and step 6, irradiation pore channel design. According to the nuclear reactor, high-power-level and long-life reactor core energy output can be achieved by means of nuclear fuel proliferation of the lead-bismuth fast reactor, and through the arrangement of the irradiation hole channels, the transmutation of minor actinide nuclides and long-life fission products, production of various isotopes, irradiation tests of related materials such as fuel elements and assemblies and the like are conducted by means of the high fast neutron fluence rate of the irradiation hole channels.

The invention discloses a nuclear reactor core temperature three-dimensional sensing and multistage intelligent regulation and control system and method, and relates to the technical field of intelligent regulation and control, and the method comprises the steps: constructing a reactor core three-dimensional temperature field through a distributed anti-radiation optical fiber sensor array, and achieving the comprehensive monitoring of local hot spots and abnormal regions. And the edge calculation module is combined with Kalman filtering and an LSTM prediction model, so that a future temperature field can be pre-judged in advance, and the perspectiveness of overtemperature risk response is improved. The system further introduces a digital twinborn simulation unit, carries out real-time simulation on various regulation and control strategies, selects an optimal scheme for execution, and achieves multi-stage linkage regulation and control of a coolant valve, auxiliary spraying, a safety rod and the like. Through the closed-loop design of perception-prediction-simulation-execution, a temperature monitoring blind area is effectively eliminated, the response delay is compressed to 0.5 second, and the safety and reliability of reactor core operation are remarkably improved.