146 results about "Nuclear reactor core" patented technology

The invention relates to the field of reactor data processing, and particularly discloses a numerical calculation method for solving a neutron transport equation. The method comprises the following steps: 1, describing the geometry of a reactor core of a nuclear reactor by utilizing a construction entity geometric modeling method, and subdividing a curved surface grid along a curved surface structure; 2, directly expanding a neutron flux distribution function in each curved surface grid by using a multivariate polynomial; 3, deriving a neutron transport equation and an integral residual equation of a boundary condition under each curved surface grid; 4, solving geometric integral parameters of each curved surface grid; step 5, establishing an adaptive linear system; and step 6, iteratively calculating and solving a maximum characteristic value and an unknown expansion coefficient.

The invention discloses a reactor core nuclear thermal coupling calculation method based on an embedded physical knowledge neural network. The method comprises the following steps: 1, dividing reactor core nodes of a nuclear reactor; 2, initializing neutron physical field parameters and thermal hydraulic parameters of the reactor core; 3, building, training and storing an embedded physical knowledge neural network model; 4, predicting the neutron number density of each node of the reactor core by using the stored model to obtain neutron flux distribution; 5, calculating reactor core power distribution according to neutron flux distribution; 6, calculating and updating the thermal hydraulic parameters of the reactor core at the tn + 1 moment by using a system analysis program; 7, calculating and updating reactivity of each node of the reactor core; and 8, judging whether the appointed calculation time is reached, if so, ending, otherwise, returning to the step 4. The method is based on an embedded physical knowledge neural network method, an accurate and efficient calculation method is provided for solving distribution of the neutron physical field and the thermal field in the reactor core, and the method has important significance on reactor design optimization.

An exemplary system and method provide nuclear reactors with optimized detector placement. The exemplary system and method include a nuclear reactor model, Markov decision process, and reward function, where reinforcement learning can be used to iteratively generate placements of detectors to candidate positions within the nuclear reactor.

A rapid inference method for a neutron flux distribution change of a reactor core of a nuclear reactor. In an offline training stage, a transient neutron transport equation is solved by means of a full-order solver, so as to obtain spatial distribution data of neutron flux and a delayed neutron precursor density changing over time, a reduced-order basis and a reduced-order coefficient data set of a training data set are obtained by means of proper orthogonal decomposition, and a reduced-order model is obtained by means of least squares fitting into the reduced-order coefficient data set; and in an online prediction stage, after the reduced-order model is solved on the basis of parameters of a predicted working state, the reduced-order basis is applied to a solved reduced-order coefficient, so as to reconstruct a physical field, which includes a neutron flux distribution of a reactor core, a delayed neutron precursor density distribution, etc. changing over time. In the present invention, data of a high-fidelity full-order model is used to construct a rapid prediction model, and a high-resolution change in the neutron flux of a reactor core over time is finely and efficiently described in a neutronics simulation, which plays an important role in reactor safety analysis, reactor data assimilation and inverse problem solving.

A nuclear power measurement value correction method and apparatus, and a computer device, a storage medium and a computer program product. The nuclear power measurement value correction method comprises: receiving a nuclear power measurement value and a thermal power measurement value of a nuclear reactor core in real time (S200); when the nuclear reactor core operates to a full-power steady state for the first time, acquiring a first nuclear power measurement value and a first thermal power measurement value of the nuclear reactor core, and when the nuclear reactor core operates to a low-power steady state for the first time, acquiring a second nuclear power measurement value and a second thermal power measurement value of the nuclear reactor core (S400); on the basis of the first nuclear power measurement value, the first thermal power measurement value, the second nuclear power measurement value, and the second thermal power measurement value, determining a correction coefficient (S600); and on the basis of the correction coefficient, correcting a subsequently received nuclear power measurement value, so as to obtain a corrected nuclear power value (S800). Using the nuclear power measurement value correction method can improve the efficiency of correcting a nuclear power measurement value.

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.

The invention provides a nuclear fuel loading and unloading mode determination method and device, equipment and a storage medium. The nuclear fuel loading and unloading mode determination method comprises the following steps: acquiring a unit number of an adjacent reactor core unit corresponding to a target reactor core unit, wherein the unit number indicates a position relationship between the adjacent reactor core unit and the target reactor core unit; obtaining nuclear fuel storage states of the adjacent reactor core units based on the unit numbers; according to the nuclear fuel storage state and the unit number, determining a target reactor core offset mode and a target loading auxiliary tool guiding mode when the target reactor core unit carries out nuclear fuel loading and unloading; and determining a target loading and unloading mode corresponding to the target reactor core unit according to the target reactor core offset mode and the target loading auxiliary tool guide mode. According to the embodiment of the invention, the efficiency and accuracy of determining the nuclear fuel loading and unloading mode of the nuclear reactor core can be improved.

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.

Flow channels are formed into a nuclear reactor core receptacle and provide a flow path for leaking primary coolant. The flow path redirects the leaking coolant from an upwardly vertical direction, in which the mass flow would impart a lift off force to a core assembly seated within the receptacle, to a horizontal flow path direction which maintains hydraulic balance of the core assembly. The flow channels in the nuclear reactor core receptacle provide a passive measure to allow hydraulic hold down of the core assemblies even in the absence of mechanical hold downs.

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.

A cutting device and method for cutting a plurality of parallel fingers of at least one core component in a plane orthogonal to the longitudinal axis of the fingers, the fingers being adapted to be introduced into the fuel assembly of a nuclear reactor core. The cutting device includes: a first bearing plate including a plurality of abutments respectively for a finger to be cut. The cutting device also includes a rotatable cutting plate including a corresponding cutting blade for each abutment in the first bearing plate, the rotatable cutting plate being rotatable about a rotating axis between a starting position and a final position in a circumferential direction, the first bearing plate and the cutting plate being parallel to each other, the cutting blades including a plurality of groups of cutting blades. For each group of cutting blades the angle between the cutting blades and the corresponding abutments is different to an angle between the cutting blades and the corresponding abutments of any other group of cutting blades.

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.

The present invention provides a nuclear reactor core operation analysis method and system, which includes: establishing a solid core geometric model; meshing the geometric model and setting physical parameters and boundary conditions of the solid core; establishing a neutron nuclear database, performing neutron physics calculation based on the database and the solid core geometric model to obtain the power size and power distribution results of the solid core; performing thermal analysis calculation based on the solid core power size and power distribution results to obtain the temperature distribution results of the solid core; performing mechanical analysis calculation based on the core temperature distribution results of the solid core to obtain deformation geometry calculation results; iteratively performing neutron physics calculation based on the core temperature distribution results of the solid core and the deformation geometry calculation results until the results of the iterative calculation meet specific convergence conditions; and obtaining neutron physics parameters, temperature distribution parameters and stress-strain distribution parameters of the fixed core under steady-state conditions based on the results of the iterative calculation.

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 invention discloses a visual test device for simulating high-temperature and high-pressure working conditions of a nuclear reactor core rod bundle channel prototype, the device is provided with a core test layer and a multi-stage depressurization layer, and a core test layer window on the side surface of the core test layer is sealed and insulated by using a sealing ring and an insulating spacer; an electric heating rod bundle is arranged in the core test layer; the multi-stage pressure reduction layer is formed by coaxially arranging a plurality of cylindrical pressure reduction layer windows outside the core test layer; circulating cooling water with stable pressure is respectively introduced into the core test layer and each pressure reduction layer, the pressure of each path of circulating cooling water is different, and the pressure difference of the two sides of each window is within the window material pressure bearing range; circulating cooling water in each pressure reduction layer is kept at a single-phase low temperature, so that visual observation is facilitated, and the temperature of the sealing material is ensured to be lower than the temperature resistance limit. According to the invention, the key technical problems of material strength limitation and thermal expansion mismatch in the high-temperature and high-pressure visualization test are solved, and a reliable test platform is provided for the high-temperature and high-pressure visualization of the reactor core rod bundle channel prototype.

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.

The present application belongs to the technical field of nuclear reactor core thermal safety analysis, and particularly relates to a thermal safety envelope curve generation method for a candu reactor core irradiation target. The method comprises the following steps: step 1, data collection and arrangement; step 2, calculation of irradiation heat release rate; step 3, establishment of a CFD analysis model; step 4, preparation of temperature limit criteria; step 5, preparation of a thermal calculation sequence; step 6, two-dimensional large-scale calculation screening; step 7, three-dimensional detailed calculation statistics; step 8, generation of an envelope curve; and step 9, envelope curve-based evaluation. The CFD calculation model used is more accurate than the theoretical calculation or numerical calculation method used in conventional methods, and fully considers external fluid flow heat transfer, non-uniform power distribution on the circumference of the VFD component outer surface, and internal asymmetric heat transfer processes.

A nuclear reactor comprises a pressure vessel containing a nuclear reactor core. A reactor core cooling system comprises a standpipe including a plurality of orifices in fluid communication with a refueling water storage tank (RWST) to drain water from the RWST into the standpipe, and an injection line configured to drain water from the standpipe to the pressure vessel. In some embodiments the standpipe is disposed in the RWST, while in other embodiments the standpipe is disposed outside of the RWST and cross-connection pipes connect the plurality of orifices with the RWST. The reactor core cooling system may further comprise a valve configured to control flow through one orifice of the plurality of orifices in fluid communication with the RWST based on water level in the standpipe. The valve may comprise a float valve having its float disposed in the standpipe.

The present invention discloses a target component, belonging to the technical field of the core of a nuclear reactor, which includes a first end plug, a clamping tube, a second end plug, a connecting tube, a third end plug and a target rod. The first end plug is used to be connected with a pressing system, and the connection between the first end plug and the clamping tube is a movable connection. A limiting member is provided inside the connecting tube. The limiting member can move axially to be connected with the first end plug so that the first end plug is connected to the clamping tube in position, and the limiting member can move axially away from the first end plug so that the first end plug and the clamping tube are detachably connected. In the present invention, the connection between the first end plug and the clamping tube is a detachable connection, and locking and unlocking are achieved through a mechanical clamping structure, which is convenient for installation and disassembly, enables structural components such as the pressing system to be reused, reduces the cost of the radiation target, reduces the generation of radioactive waste, and at the same time, no debris or falling parts will be generated during the underwater disassembly process of the target component, and it will not affect the water in the spent fuel pool.

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 invention relates to a method for chemically conditioning a heat transfer fluid in a primary circuit of a nuclear power plant, the circuit comprising a nuclear reactor core (NRC) and a steam generator (SG), each made of a metal alloy comprising nickel, the method comprising the steps of: A) injecting reducing and acid-base species into the heat transfer fluid in order to define, at a given temperature, a reference hydrogen potential-redox potential pair; B) injecting a reducing species at the inlet (MP1) of the steam generator in order to decrease the redox potential of the heat transfer fluid, thereby decreasing the solubility of the various phases of the nickel in the heat transfer fluid; C) injecting an oxidizing species at the inlet (MP2) of the core in order to compensate for the effects of the reducing species injected in step B), thereby increasing the redox potential of the heat transfer fluid and the solubility of the various phases of the nickel in the heat transfer fluid.

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 present invention discloses a method for automatically suppressing xenon oscillations and providing auxiliary warning prompts based on the 3AO method, belonging to the technical field of nuclear reactor core monitoring. First, a variety of high-precision sensors are arranged at key positions of the nuclear reactor to timely capture signs of xenon oscillations. By analyzing sensor data, the system can accurately identify early signals of xenon oscillations, providing a basis for formulating subsequent control strategies. The core of the control strategy is the integration of the 3AO method and the deep Q-network algorithm. This algorithm not only comprehensively considers the immediate state of the core but also optimizes control decisions by learning historical data, thereby calculating the optimal control rod position and boron concentration adjustment plan. Subsequently, according to the control instructions generated by the intelligent algorithm, the system automatically prompts the operator to adjust the control rods and boron concentration to respond in real time to the changing requirements of the nuclear reactor. By providing real-time feedback on the adjustment effects, an efficient closed-loop control system is formed, which can continuously improve the accuracy and response speed of the control strategy.

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.