181 results about "Nuclear reactor core" patented technology

The invention discloses a nuclear reactor core channel analysis method based on two fluid models, which comprises the following steps of: 1, establishing a fluid domain computational grid according to the structure of a nuclear reactor core or an assembly; 2, respectively establishing a control equation set and a heat conduction model of the fuel rod for the continuous gas-liquid phase and the discrete liquid drop phase; 3, considering physical phenomena such as flow heat exchange, resistance pressure drop and turbulence mixing in a reactor core, and implanting a corresponding constitutive model; 4, discretizing the control equation based on a second-order discrete format to obtain a discrete equation set; 5, dividing parallel tasks and parallel regions through an MPI parallel strategy to form a large coefficient matrix of a pressure equation; and 6, solving a discrete equation set and a large coefficient matrix through a pressure coupling semi-implicit algorithm and a stable biconjugate gradient method. According to the method, the key physical parameter distribution of the pressurized water reactor can be more accurately and reasonably calculated, so that technical guidance is provided for safety analysis and engineering design of the reactor.

The invention relates to the field of theoretical calculation simulation of a nuclear reactor core state, in particular to a simulation method for dynamic reactor core power distribution measurement and a related product, and the method comprises the following steps: obtaining accurate burnup data; establishing a steady-state reactor core theoretical model based on accurate burnup; solving a three-dimensional neutron kinetic equation by taking the steady-state reactor core theoretical model as an initial state and taking a relation curve of reactor core operation power and operation time as input to obtain a dynamic reactor core theoretical model taking time as a function; inputting the power and time relation corresponding to the power and time relation corresponding to the power plant test plan into the dynamic reactor core theoretical model to obtain a database; according to the method, the steady-state theoretical model is constructed by utilizing accurate burnup data, dynamic reactor core simulation is carried out through a transient reactor core neutronics calculation model and the like, a dynamic reactor core model completely consistent with an actual test state can be generated, and actually measured data of dynamic reactor core power distribution can be directly processed and accepted.

Disclosed is a nuclear reactor, comprising: a canister, a nuclear reactor core housed inside of the canister, and a controlled temperature housing attached to the canister. The nuclear reactor core comprises a control drum and a shaft extending from the control drum and out of the nuclear reactor core. Heat generated by the nuclear reactor core enters the controlled temperature housing. The controlled temperature housing comprises a proximal end attached to the canister, a first internal plate spaced from and positioned distal to the proximal end, a second internal plate spaced from and positioned distal to the first internal plate, and a motor mounted to the second internal plate. The shaft enters the controlled temperature housing through the proximal end and extends to the motor. The controlled temperature housing is to passively maintain the motor at a temperature equal to or below a predetermined temperature threshold.

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.

The invention discloses a calculation method for a nuclear reactor neutron transport equation, and belongs to the field of nuclear reactor core design and safety. According to the method, the problem that the convergence speed of an inner iteration process of a transport equation and a low-order equation in a traditional PI-CMFD method is low is solved. The method specifically comprises the following steps: step 1, reading a coarse mesh model, a fine mesh model, material information and boundary conditions of a nuclear reactor; step 2, establishing a numerical discrete model of a steady-state multi-group transport equation according to the fine grid model read in the step 1; 3, after boundary constraint and material constraint are applied to the established numerical discrete model of the steady-state multi-group transport equation, a discrete equation set is constructed by transforming a test function in the model; and 4, solving the constructed discrete equation set based on the grid mapping relation between the coarse grid model and the fine grid model by adopting a PI method and a CMFD method to obtain a neutron angular flux value and a characteristic value of the neutron transport equation. The method can be applied to calculation of the neutron transport equation of the nuclear reactor.

The invention relates to the technical field of nuclear reactor core calculation, and discloses a method and a device for calculating a physical and thermal coupling numerical value of a reactor, a medium and equipment. In the process of calculating the physical and thermal coupling numerical value of the reactor, a neutron space-time dynamic model is subjected to time discretization by adopting a diagonal implicit Runge-Kutta format; in the transient analysis process, coupling calculation is carried out through a simple operator splitting method and a Picard iteration method to obtain a calculation value, coupling global solution variables are selected based on a Newton method, a global solution residual equation is established, solving is carried out through a Krylov subspace solution method based on the coupling global solution variables and the global solution residual equation, a JFNK iteration process is constructed, and a transient analysis result is obtained. And finally, taking the calculated value as an initial value, establishing a hybrid iteration process according to the initial value and the JFNK iteration process, and calculating to obtain a reactor transient physical thermal full-coupling numerical solution. According to the method, excessive solution is avoided, the convergence speed is increased, and stable and rapid convergence calculation is realized.

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 fast neutron nuclear reactor contains a nuclear reactor core having an array of device locations. Some device locations in the nuclear reactor core contain fissile and fertile nuclear fuel assembly devices. One or more other device locations in the nuclear reactor core contain Doppler reactivity augmentation devices that amplify the negativity of the Doppler reactivity coefficient within the nuclear reactor core. In some implementations, a Doppler reactivity augmentation device can also reduce the coolant temperature coefficient within the nuclear reactor core. Accordingly, a Doppler reactivity augmentation device contributes to a more stable nuclear reactor core.

A nuclear reactor is configured with a primary coolant loop for transferring heat away from the nuclear reactor core. In a shutdown event, the primary coolant pump may stop pumping primary coolant through the reactor core, resulting in decay heat buildup within the reactor core. An inertial energy coast down system can store kinetic energy while the nuclear reactor is operating and then release the stored kinetic energy to cause the primary coolant to continue to flow through the nuclear reactor core to remove decay heat. The inertial energy coast down system may include an impeller and a flywheel having a mass. During normal reactor operation, the flowing primary coolant spins up the impeller and flywheel, and upon a shutdown event where the primary coolant pump stops pumping, the flywheel and impeller can cause the primary coolant to continue to flow during a coast down of the flywheel and impeller.

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 computer implemented method for simulating an operation of a reactor core includes determining an initial state of the reactor core; calculating a nodal target power distribution and / or the target 3D neutron flux distribution; obtaining an actual power distribution and / or the actual 3D neutron flux distribution of the nuclear reactor core; determining a difference between the target power distribution and the actual power distribution of the nuclear reactor core and / or determining a difference between the target 3D neutron flux distribution and the actual 3D neutron flux distribution of the nuclear reactor core; determining modal expansion coefficients using a Fourier modal decomposition based on the determined difference and applying a Modal Generalized Perturbation Theory to the modal expansion coefficients for determining a 3D cross-section distribution perturbation causing the determined difference; and determining a 3D adaptation distribution for the determined difference based on the determined 3D cross-section distribution perturbation.

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 present invention relates to the field of nuclear power engineering, namely the placement of unsheathed fuel assemblies in a nuclear reactor core, aiming to improve the heat dissipation of fuel elements in a nuclear reactor. The nuclear reactor core comprises a hexagonal unsheathed fuel assembly having fuel elements located on a triangular grid with a spacing greater than the diameter of the fuel element cladding. The fuel elements comprise spiral spacer ribs on the cladding and are rigidly fixed in a given angular direction so that the fuel elements are in rib-to-rib contact at a height position at which the fuel elements reach the maximum fuel element cladding temperature during reactor operation. The fuel assembly is positioned in such a way that the fuel elements located on the surface of the fuel assembly form a single triangular grid with the remaining fuel elements in the core, and ensure that all the fuel elements are in rib-to-rib contact at a height position at which the highest fuel element cladding temperature is reached during reactor operation. At the same time, incorrect placement of the fuel assembly in the core is eliminated.

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.

To provide a nuclear fuel assembly.SOLUTION: A nuclear fuel assembly for nuclear reactor cores includes at least one fuel compact including a fuel compact coating tube defining an inner volume and a plurality of fuel pins each including a pin coating tube defining an inner volume and a fissile fuel arranged in the inner volume of the pin coating tube. The plurality of fuel pins are arranged in the inner volume of the fuel compact coating tube; and the sectional shape of the fuel compact is defined by the fuel compact coating tube.SELECTED DRAWING: Figure 1

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.