60 results about "Neutron transport" 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 belongs to the technical field of nuclear reactor physical calculation, and particularly discloses a multi-group steady-state neutron transport coarse mesh node global efficient solving method and system, which effectively fuse and exert the advantages of a coarse mesh node method and a Newton-Kralov method. A general coarse mesh segment transport calculation model suitable for various polyhedral segments is constructed, high-precision and high-efficiency transport scanning calculation is realized under sparse segments, the number of variables is greatly reduced, and the calculation scale of single transport scanning is reduced; a Newton-Kralov subspace global solution framework is established, a global solution variable is selected based on a variable coupling relationship of a coarse network nodal method, a linear equation corresponding to a Newton step length is solved by using a Kralov subspace method, and the Newton step length is used for updating the global solution variable to realize rapid and efficient convergence. On the premise of ensuring the calculation precision, the calculation efficiency is remarkably improved, and the calculation scale is reduced, so that the method is particularly suitable for the problem of large-scale and high-dominance-ratio neutron transportation.

The invention provides a neutron transport simulation method and device, a medium, equipment and a BNCT treatment system. The method comprises the following steps: setting a plurality of voxel grids; the voxel grids are various human tissue materials corresponding to different boron drug concentrations; performing off-line coding on each voxel grid to obtain a corresponding physical coding layer, wherein the physical coding layer is used for predicting a corresponding neutron outgoing flow and neutron flux distribution based on the neutron incoming flow of the voxel grid; acquiring a CT image of the target object, and matching the voxel grid according to the CT image and the boron medicine planned concentration to acquire a voxel model of the target object; the physical coding layers of all voxel grids in the voxel model are coupled online. According to the method, the physical coding layers of various voxel grids are generated in an off-line manner, and the corresponding neutron outgoing flow and neutron flux distribution can be quickly predicted based on the given neutron incoming flow; and coupling a plurality of voxel grids based on a patient individualized voxel model to quickly obtain neutron flux distribution in the voxel model.

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.

The invention discloses a penetration probability-based characteristic line neutron transport calculation accelerated scanning method, which comprises the following steps of: firstly, establishing characteristic line information, and dividing each characteristic line into a plurality of characteristic line segments; establishing a penetration probability coefficient from the incident boundary angle flux to the emergent boundary angle flux directly along each characteristic line; scanning the whole characteristic line in all directions, calculating an emergent boundary angle flux from the incident boundary angle flux by using the penetration probability coefficient, obtaining an incident boundary angle flux of the boundary point in the other direction from the emergent boundary angle flux, calculating the emergent boundary angle flux from the incident boundary angle flux by using the corresponding penetration probability coefficient again, and repeating the process to obtain the emergent boundary angle flux of the boundary point in the other direction. The incident boundary angular flux in all directions of the boundary point is converged; and for each feature line in all directions, only one time of feature line segment-by-segment calculation is carried out, flux information on the fine grid is obtained, and scanning is ended. According to the method, multiple times of feature line segment-by-segment calculation before convergence are avoided, the calculation amount is reduced, and the calculation efficiency of the feature line method is improved.

The invention discloses a Monte Carlo neutron transport geometric tracking method using a GPU ray tracing technology, and relates to the technical field of reactor physical particle transport simulation, and the method comprises the steps: generating an axis alignment bounding box according to spatial position recursion grouping based on triangular patch geometry, and constructing a hierarchical bounding box acceleration structure; based on the structure, the GPU ray tracing and geometry closing principle is utilized, the initial position of the particle is determined through intersection detection of random rays and a patch set, and the final geometry where the particle is located is obtained through three times of repeated positioning verification; establishing a bilateral index table, giving a unique index to lattice cells on two sides of each patch, and binding the data structure to a hierarchical bounding box acceleration structure; and completing an intersection test of the particle and the surface patch based on GPU ray tracing, accessing the index table according to an included angle between a ray and a surface patch normal to obtain geometry switching information, and completing Monte Carlo neutron transport calculation. According to the method, the efficiency bottleneck of the Monte Carlo method in UM geometry is effectively solved, and a higher calculation speed is obtained.

The invention discloses a nuclear reactor neutron flux prediction method based on global conservation constraint, and belongs to the nuclear reactor physics and machine learning crossing field, and the method comprises the following steps: constructing a depth operator network model based on global neutron conservation constraint; collecting solution data of a neutron transport equation or a neutron diffusion equation under different parameters, constructing a training data set, and training the depth operator network model to obtain an optimal model weight; and initializing the depth operator network model by using the optimal model weight, applying the depth operator network model to a neutron transport equation or a neutron diffusion equation under different parameters for solving, and predicting neutron flux distribution in a corresponding region of the nuclear reactor. According to the method, the physical constraint term for describing the global conservation of neutrons is introduced, so that the accuracy and the physical consistency of the DeepONet network for predicting the neutron transport or neutron diffusion solution can be remarkably improved, the generalization ability and the robustness of the model can be enhanced, and the method is particularly suitable for meeting the rapid and high-precision calculation requirements in the fields of reactor design, reactor core analysis and nuclear safety evaluation.

A method for tracking the lunar regolith depth of the lunar surface neutron source based on particle transport analysis, comprising: obtaining the high-energy particle spectrum of the lunar surface; establishing a particle transport model; according to the high-energy particle spectrum and the particle transport model, emitting initial particles into the lunar regolith model, and forming events and tracks after interacting with the lunar regolith after entering the lunar regolith; first forming an initial trajectory record and transmitting it to the event record set; the neutron tracks are transported step by step, and the current event record is updated according to the property change of the neutrons caused by the interaction; the neutrons perform elastic collision motion in the lunar regolith, until the neutron transport ends, forming an end track record and transmitting it to the event record set; when all the events are transported, output all the information in the event record set; according to the high-energy particle spectrum statistics, obtain the lunar regolith depth from which neutrons of different energies on the lunar surface are sourced, providing a basis for the design of the drilling depth for water exploration on the lunar surface.

The application discloses a neutron transport constitutive relation and low-dimensional control equation modeling method based on sparse regression, which comprises the following steps: a neutron flow matrix under multiple state parameters is formed by state parameters of each state and physical quantity parameters calculated through a neutron transport program; fitting physical quantity parameters are filled into a sparse regression physical quantity matrix; a coefficient matrix is obtained by using a least square method to calculate an inverse matrix; and a neutron transport constitutive relation under multiple state parameters is obtained by assembling. The application utilizes the accurate characteristics of the calculation result of the neutron transport equation, grasps the cause of the larger deviation of the calculation result of the neutron diffusion equation in a small scale range, combines sparse regression fitting to obtain the neutron transport constitutive relation, and further obtains a low-dimensional macroscopic neutron transport control equation. The theoretical derivation is less difficult, the research progress is easier, the effect of improving the calculation precision is more obvious, the applicability is better, the gap with the situation under the actual reactor core working condition is not large, and the application has certain practicality.

The invention relates to the technical field of neutron transport numerical calculation, in particular to a cross section generation method and device based on flux moment homogenization and transport correction coupling and a storage medium, and the method comprises the steps: S1, building a geometric model of a reactor assembly or lattice cell, carrying out the energy Monte Carlo calculation, and generating a conventional cross section based on statistical data; s2, extracting a spherical harmonic function flux moment and a total reaction rate moment, calculating a total cross section spherical harmonic function correction term, and introducing a conventional scattering cross section to complete flux moment homogenization correction; and S3, exporting a first-order scattering matrix, carrying out transport correction to obtain a transport correction term, and correcting the conventional cross section according to the transport correction term and the total cross section spherical harmonic function correction term to generate a final multi-group cross section. According to the method, the anisotropies of the total cross section and the scattering cross section are synergistically corrected, the calculation precision of the method in the strong anisotropy problem is improved on the premise of keeping the standard cross section data structure, and the compatibility with a widely applied reactor core calculation program is ensured.

The invention provides an energy group structure determination method for physical calculation of a reactor core of a high-temperature gas cooled reactor, and the method comprises the steps: constructing a reactor core physical model which is a reactor core model comprising multiple batches of fuel balls and graphite balls; carrying out forward neutron transport calculation on the reactor core physical model to obtain multi-group forward neutron flux density of each batch of fuel spheres and graphite spheres in the reactor core physical model; carrying out conjugate neutron transport calculation on the reactor core physical model to obtain multi-group conjugate neutron flux density of each batch of fuel spheres and graphite spheres in the reactor core physical model; determining an average flux weight in the first group partition based on the multi-group forward neutron flux density and the multi-group conjugate neutron flux density; and when the average flux weight meets a division condition, dividing the reactor core physical model based on a boundary corresponding to the average flux weight to obtain a first group energy group structure. According to the invention, the calculation memory consumption of the pebble-bed high-temperature gas cooled reactor is low, the calculation time is short, and the calculation time and calculation resources are saved.

This invention discloses a neutron transport calculation system with decoupled geometric modeling and mesh generation, including a mesh generator and a CSG geometric modeling module. The module is fully embedded into the computational framework of the neutron transport procedure using the feature line method. First, a suitable type of mesh generator is selected based on the geometry of the underlying material cell. The internal structure of the mesh generator is completed by setting the meshing parameters. Subsequently, the mesh generator is set as an attribute of the material cell. The mesh generator's built-in positioning algorithm can quickly determine the fine mesh index of the ray segment endpoints; its intersection algorithm can accurately calculate the intersection distance between the ray and the fine mesh. Thus, all feature line segment information is generated. Finally, source iteration is performed until the convergence condition is met to obtain the key physical parameters. By using this invention, fine mesh generation can be achieved, which is helpful for subsequent reactor parameter acquisition. This invention can be widely applied in the field of nuclear reactor physics numerical calculation technology.

The present disclosure provides a neutron transport simulation method, related device and computer program product, relating to the technical field of computer technology and nuclear physics. The method is implemented in the following manner: CSG modeling is performed on a reactor core to construct a CSG tree; N projection planes are generated around the reactor core, and an incident point array is generated on all the projection planes; a blank Segment list is constructed, and recursive characteristic line tracking is performed on all the incident points starting from the root node of the CSG tree, and Segment data is sequentially filled into the blank Segment list; GPU layer and GPU internal thread layer load balancing is performed; the characteristic line, FSR and material data are transmitted to each GPU; each GPU performs transport scanning calculation; the global FSR angular flux is calculated; each GPU calculates the fission moment and the fission source; the K effective value is calculated and whether it converges is verified. The technical solution of the present disclosure can break the CSG geometry construction restriction in the existing neutron transport simulation method, reduce unnecessary characteristic line generation to save computing resources, and improve the iteration solving speed to speed up the simulation.

The application provides a supercritical carbon dioxide reactor energy group structure optimization method, and relates to the technical field of nuclear reactors, which adopts a particle swarm optimization algorithm (PSO) to optimize the neutron energy group structure of a supercritical carbon dioxide reactor, so as to improve the calculation efficiency and ensure the calculation accuracy. The method firstly constructs a one-dimensional supercomponent model, and uses a Serpent Monte Carlo program and a VI TAS deterministic neutron transport program to calculate a reference solution. Then, the PSO algorithm is used to optimize the energy group structure, the position and speed of the particles are iteratively updated, and the optimal energy group structure is searched. In the optimization process, each particle in the particle swarm represents a group of possible energy group boundaries, and the effectiveness of the optimization result is verified by comparing parameters such as keff, cross section, flux, power and the like. The method of the application can improve the accuracy of the neutron transport calculation of the supercritical carbon dioxide reactor and maintain the reasonableness of the full reactor power distribution, and provides an effective technical means for the optimization of the reactor.

This invention relates to a method for neutron transport calculation or resonance calculation for dispersed fuels. The method is characterized by correcting the matrix chord length probability and average chord length used in related calculations based on the true chord length probability distribution and true average chord length of the matrix, and pre-constructing an interpolation table or fitting formula for solving the true average chord length correction factor using dimensionless parameters. The method for solving the true chord length probability distribution and true average chord length of the matrix includes the following steps: setting the geometric shape of a random medium region according to the fuel type; generating a random distribution of particle positions within the given random medium region according to a given filling rate; randomly selecting particle surfaces to isotropically emit neutrons outwards, or uniformly and isotropically emitting neutrons from the matrix, and calculating the distance to the next particle; statistically analyzing the chord lengths of all neutrons passing through the matrix, thereby obtaining the true chord length probability distribution and true average chord length of the matrix.

The application discloses a kind of special-shaped fuel geometric nested neutron transport and burnup calculation method and system, method includes: based on geometric nested model, the neutron transport calculation of special-shaped reactor is realized;Geometric nested model is obtained by constructing as follows: using the modeling of conventional assembly to construct entity geometry method, using direct acceleration geometry Monte Carlo method to model special-shaped fuel assembly, the geometry system space of direct acceleration geometry Monte Carlo geometry is embedded into the geometry system space of construct entity geometry, complete geometric nested modeling;After neutron transport calculation is completed, the burnup of geometric nested model is calculated, and the burnup calculation of special-shaped fuel reactor is realized.The application can improve the problem that Monte Carlo calculation cannot accurately and efficiently establish special-shaped fuel model using CSG method, improve the calculation efficiency using DAGMC method, and accurately and efficiently complete special-shaped fuel reactor core modeling and neutron transport calculation and burnup calculation.

This application discloses a method, apparatus, electronic device, and storage medium for calculating fuel sphere burnup, relating to the field of nuclear reactor physics and thermal calculation technology. The method includes: dividing fuel spheres into multiple batches based on their distribution; establishing neutron transport equations for each batch of fuel spheres; calculating the self-shielding factor of each batch of fuel spheres; calculating the neutron flux density of each region of the reactor core using the homogenized cross section of the pebble bed region and the homogenized cross section of the reflector layer region; calculating the power density of each region of the pebble bed based on the neutron flux density of each region of the reactor core; calculating the neutron flux density and power density of each batch of fuel spheres at the current reactor core power level based on the neutron flux density of the pebble bed region and the self-shielding factor of each batch of fuel spheres; and calculating the burnup depth of the corresponding batch of fuel spheres based on the power density of each batch of fuel spheres at the current reactor core power level, thereby improving the technical accuracy of fuel sphere burnup calculation in non-uniformly packed pebble bed reactors.

The invention discloses a lattice cell scale two-step method reactor core physical analysis method based on a first-order spherical harmonic function, which comprises the following steps of: 1, performing full-component one-step method non-uniform neutron transport calculation to obtain neutron standard flux density and neutron angle flux density in the non-uniform transport calculation; 2, calculating a homogenized few-group cross section; 3, lattice cell discontinuous factor calculation; 4, calculating a diffusion coefficient matrix; 5, solving a P1 equation; 6, calculating power and thermotechnical feedback; and 7, if the nuclear thermal coupling converges, outputting three-dimensional power distribution for subsequent safety analysis. According to the method, the first-order spherical harmonic function equation is solved repeatedly based on the angle-related section, the calculation efficiency is remarkably improved while it is guaranteed that the calculation precision is equivalent to that in the prior art, and the calculation time is shortened by 40% or above; and accurate data support is provided for design analysis of the nuclear energy device.

The present application provides a neutron transport simulation method, device, medium, equipment and BNCT treatment system. The method includes: setting a plurality of voxel grids; the voxel grids are various human tissue materials corresponding to different boron drug concentrations; performing offline encoding on each voxel grid to obtain a corresponding physical encoding layer, and the physical encoding layer is used to predict the corresponding neutron outgoing flux and neutron flux distribution based on the neutron incident flux of the voxel grid; obtaining the CT image of the target object, and matching the voxel grid according to the CT image and the planned boron drug concentration to obtain the voxel model of the target object; performing online coupling on the physical encoding layers of all the voxel grids in the voxel model. The present application generates the physical encoding layers of various voxel grids offline, and can quickly predict the corresponding neutron outgoing flux and neutron flux distribution based on the given neutron incident flux; and couples a plurality of voxel grids based on the patient's individualized voxel model to quickly obtain the neutron flux distribution in the voxel model.

The application belongs to the technical field of nuclear reactor physics numerical calculation, and discloses a symmetric boundary processing method and system of a neutron transport characteristic line method. A geometry core model is constructed through an MOC program, boundary conditions of the geometry core model are set, the geometry core model is arranged with characteristic lines to obtain characteristic line arrangement information, symmetry of the geometry core model is set, a corresponding additional module is generated, an equivalent complete module of the additional module is generated according to the symmetry based on the additional module, boundary segment angular flux of the equivalent complete module is calculated, and the equivalent complete module is calculated according to the characteristic line arrangement information and based on the boundary segment angular flux of the equivalent complete module and a known source term distribution to obtain the equivalent complete module. The application realizes reduction of the difficulty of characteristic line arrangement of the additional module, and further improves the efficiency of neutron transport calculation.

A fast fuel management model for a liquid metal fast reactor, a method for constructing the same, and an application thereof, belonging to the technical field of nuclear reactor physics. To achieve fast three-dimensional core calculations, the present invention establishes a two-dimensional single assembly model, a super-assembly model, and a radial reflector model by constructing an entity geometry method; performs assembly calculations using the Monte Carlo continuous energy point cross-section neutron transport method; for core calculations, uses the nodal diffusion method to perform core calculations and solve the core neutron diffusion equation; combines an independent burnup calculation module to construct a core diffusion and burnup calculation model, accurately solve the burnup equation, and update the nuclear density; updates the macroscopic cross-section, uses the homogenized microscopic cross-section of the assembly at the initial moment and the nuclear density updated by the assembly burnup calculation to assemble the homogenized macroscopic cross-section of the assembly at the next burnup step and perform the core calculation of the next burnup step until the end of the core life cycle. The present invention realizes fast core burnup calculations and fast fuel management.

The present invention relates to a fast-spectrum small-scale nuclear reactor core transient multi-physics field coupling method and computer program product to address the technical problems of existing nuclear reactor transient analysis, which considers a single factor, has low reliability, and cannot achieve accurate and comprehensive evaluation. The method includes: 1. Constructing a neutron microscopic cross-section database for the nuclear reactor core; constructing a neutron transport kinetics, thermal hydraulics, and mechanical calculation grid model; setting physical property parameters for thermal hydraulics and mechanical calculations; 2. Setting a preset total simulation time and determining initial and boundary conditions for the calculation; 3. Time stepping to update the current time; calculating the power density distribution, temperature distribution, density distribution of the core at the current time, and the displacement of each node of the core and reflector calculation grid, iterating until the convergence limit is met, and determining whether it is the final time. If so, the calculation ends; if not, time stepping continues until the calculation ends.

This invention belongs to the field of nuclear reactor physics computational technology, and specifically discloses a coarse-grid nodal global efficient solution method and system for multi-group steady-state neutron transport. It effectively integrates and leverages the advantages of the coarse-grid nodal method and the Newton-Krylov method, constructing a general coarse-grid nodal transport computation model applicable to various polyhedral nodals. It achieves high-precision and high-efficiency transport scan computation under sparse nodals, significantly reducing the number of variables and compressing the computational scale of a single transport scan. A Newton-Krylov subspace global solution framework is established. Global solution variables are selected based on the variable coupling relationship of the coarse-grid nodal method, and the Krylov subspace method is used to solve the linear equations corresponding to the Newton step size. The obtained Newton step size is used to update the global solution variables to achieve fast and efficient convergence. This invention significantly improves computational efficiency and reduces computational scale while ensuring computational accuracy, making it particularly suitable for large-scale, high-dominance neutron transport problems.

A reactor transport criticality simulation method and system based on the Krylov-Schur algorithm, relating to the field of nuclear reactor core design and safety, and specifically to a reactor transport criticality simulation method and system based on the Krylov-Schur algorithm. The purpose of the present invention is to realize the transport criticality calculation and solve the maximum eigenvalue using the Krylov-Schur method, and to improve the convergence accuracy of the maximum eigenvalue and the eigenvector. Technical points: The steady-state neutron transport equation is discretized and deformed in the form of an operator, the energy variable E in the steady-state neutron transport equation is approximated by multiple groups, the angle variable is discretized by vertical scale to obtain a coupled equation expressed in the form of an operator, a standard eigenvalue problem is given, and the eigenvalues ​​and eigenvectors of the matrix P in the standard eigenvalue problem are solved. The method of the present invention is used for reactor transport criticality simulation.

The present invention belongs to the field of nuclear design and safety analysis based on comprehensive neutron physics computational simulation in nuclear science, and particularly relates to a nuclear simulation analysis system and method. The system includes a geometry modeling module, a physics modeling module, a neutron transport modeling module, a neutronics multi-physics coupling modeling module, and a processing module. The significance of the present invention lies in that an integrated computational model for comprehensive simulation of different physical processes can be constructed based on a unified framework, the real physical processes can be fully considered, and the mutual correlation effects between various physical processes in dimensions such as space, energy, and time can be accurately reflected, so as to better serve nuclear simulation analysis.

The invention discloses a neutron transport node coarse net generation method based on fine net combination, which comprises the following steps of: 1, firstly traversing all radial fine grid boundary lines, and reading the length of the boundary lines, the areas of grids on two sides, material information and the like; 2, assuming that grids on two sides of the boundary line are merged, respectively calculating vector included angles between end points at two ends of the boundary line and adjacent boundary points on the grids, judging whether concave boundaries appear at the end points, and deleting the grids with the concave boundaries from the boundary line list; 3, calculating the combination weight of each boundary line based on the length of the boundary line, the area of the grids on the two sides and the material information; 4, screening the boundary line with the maximum merging weight in the boundary line list; 5, merging the grids on the two sides, deleting boundary line information, and updating a boundary line list; and 6, if residual elements exist in the boundary line list, returning to the step 2. According to the method, the convex boundary attribute of the coarse mesh is obtained, the method is suitable for any radial mesh, material information on the two sides of the mesh is considered, the method can be used for coarse mesh generation of a coarse mesh acceleration method of a neutron transport node method, and the quality of the coarse mesh is improved.

The invention discloses a neutron transport unstructured grid calculation result three-dimensional visualization system and method, and the system comprises a reading module which is used for reading the original data of the unstructured grid calculation result; the data preprocessing module is used for carrying out data partitioning on the original data, establishing a thread pool and executing multi-thread analysis in parallel to obtain analysis data; the data merging module is used for performing data merging on the analysis data to obtain complete unstructured grid calculation result analysis data; the data structure construction module is used for constructing a data structure according to the unstructured grid calculation result analysis data to obtain grid optimization data; the color mapping module is used for performing color mapping on the grid optimization data; and the rendering module is used for rendering the model passing through the color mapping module to obtain a three-dimensional visualization result of the neutron transport unstructured grid calculation result. The technical problems that in the prior art, the visualization rendering efficiency of an unstructured grid calculation result is low, and the model precision is low are solved.

The invention provides a neutron transport scanning operator optimization method based on built-in function and instruction level parallelism, and the method comprises the steps: firstly reconstructing a data layout, declaring an array according to an energy group sequence, carrying out the memory alignment, and enabling a calculation process to adapt to an SIMD vectorization architecture; then implementing explicit vectorization calculation by adopting a built-in function, performing polar angle cycle expansion and instruction recombination, and fully mining instruction-level parallel potential; meanwhile, key index calculation is optimized into vectoring polynomial approximation. Based on the optimized data flow and the parallel computing mode, efficient updating of the angular flux and the standard flux is achieved, and convergence judgment and calculation are completed. According to the method, the calculation throughput rate and the hardware resource utilization rate of the scanning operator are remarkably improved, and the method can be coordinated with coarse-grained parallel strategies such as MPI / OpenMP and is suitable for a multi-architecture high-performance calculation environment.

The invention discloses a reactor neutron noise rapid simulation method based on a Green function, and the method comprises the steps: obtaining a steady-state effective multiplication factor and a steady-state neutron flux density through a steady-state neutron transport equation according to basic parameters of a reactor; initializing a unit point source parameter and setting a unit point source, and circularly solving a Green function form of a frequency domain neutron noise equation by combining the obtained steady-state effective multiplication factor and steady-state neutron-flux density; obtaining the distribution of the frequency-domain neutron noise Green function excited by each unit point source about an energy group, a space variable, an angle variable and a frequency variable, and generating a frequency-domain neutron noise Green function database; and according to a neutron noise spectrum under excitation of any noise source, rapid calculation is carried out through linear superposition of a Green function and a noise source item. According to the method, the solution of the neutron noise spectrum under any noise source is simplified into linear superposition calculation, and the calculation efficiency of reactor neutron noise numerical simulation is improved.

The invention discloses a nuclear reactor neutron transport calculation method and device, and the method comprises the steps: carrying out the space merging of a fine network multi-group neutron transport model in each time of source iteration in a process of calculating the fine network multi-group neutron transport model through a source iteration mode, and obtaining a coarse network multi-group neutron transport model; correcting the first coarse mesh multi-group neutron standard flux through the coarse mesh multi-group correction factor to obtain a second coarse mesh multi-group neutron standard flux, and solving the coarse mesh multi-group neutron transport model to obtain a third coarse mesh multi-group neutron standard flux; performing iterative solution on the coarse mesh multi-group neutron transport model to obtain a target coarse mesh multi-group neutron standard flux so as to determine a fine mesh multi-group correction factor; and correcting the initial fine-net multi-group neutron standard flux through the fine-net multi-group correction factor to obtain a target fine-net multi-group neutron standard flux, and calculating the fine-net multi-group neutron transport model through the target fine-net multi-group neutron standard flux to improve the efficiency of nuclear reactor neutron transport calculation.