81 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 discloses a pin-by-pin-level steady-state physical-thermal coupling analysis method for a liquid metal cooling fast reactor, which is based on an OpenFOAM platform, realizes fast and high-fidelity physical-thermal coupling calculation by coupling a three-dimensional steady-state neutron transport equation and a three-dimensional CFD control equation under a sub-channel grid, and realizes high-fidelity physical-thermal coupling analysis of the three-dimensional steady-state neutron transport equation and the three-dimensional CFD control equation. And three-dimensional physical and thermal phenomena such as neutron space scattering, cross flow and backflow are accurately analyzed. The method comprises the steps that firstly, OpenMC is adopted for fine modeling, and a few-group constant library is generated; secondly, geometric modeling and grid division are carried out; and then, setting an initial physical field and boundary conditions, respectively carrying out internal iteration solution of the physical thermal engineering, gradually converging the neutron flux and the coolant temperature field through external iteration between the physical thermal engineering, and finally obtaining a steady-state physical-thermal coupling characteristic. The efficient and accurate pin-by-pin level coupling analysis method is provided for the physical-thermal characteristics of the liquid metal cooled fast reactor, and effective guidance is provided for design optimization and safety analysis.

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 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 discloses a neutron transport constitutive relation and low-dimensional control equation modeling method based on sparse regression, and the method comprises the steps: enabling a state parameter of each state and a physical parameter calculated through a neutron transport program to form a neutron current matrix under a plurality of state parameters, and filling a sparse regression physical parameter matrix with the built fitting physical parameter, calculating an inverse matrix by using a least square method to obtain a coefficient matrix, and assembling to obtain a neutron transport constitutive relationship under the plurality of state parameters; according to the method, the characteristic that the calculation result of the neutron transport equation is accurate is utilized, the reason that the calculation result of the neutron diffusion equation is large in deviation in a small-scale range is grasped, the neutron transport constitutive relation is obtained in combination with sparse regression fitting, then the low-dimensional macroscopic neutron transport control equation is obtained, the theoretical derivation difficulty is small, the research progress is easier, and the method has good application prospects. And the effect of improving the calculation precision is obvious, the applicability is better, the difference from the situation under the actual reactor core working condition is small, and certain practicability is achieved.

The invention discloses a reactor transport critical simulation method and system based on a Krylov-Schur algorithm, and relates to the field of nuclear reactor core design and safety, in particular to the reactor transport critical simulation method and system based on the Krylov-Schur algorithm. The objective of the invention is to realize that the Krylov-Schur method is used for carrying out transport critical calculation to solve the maximum characteristic value, and the convergence precision of the maximum characteristic value and the characteristic vector is improved. The method is technically characterized by comprising the following steps of: discretizing a steady-state neutron transport equation and deforming in an operator form, performing multi-group approximation on an energy variable E in the steady-state neutron transport equation, performing discrete longitudinal standard discretization on an angle variable to obtain a coupled equation, expressing the coupled equation in the operator form, and giving a standard characteristic value problem; and solving the eigenvalue and eigenvector of the matrix P in the standard eigenvalue problem. The method is used for reactor transportation critical simulation.

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 an abnormal vibration control rod positioning method and system based on neutron noise signals and a medium. Relates to the technical field of control rod monitoring. According to neutron noise signal data measured by a self-powered neutron detector, whether the control rod has abnormal vibration or not is evaluated, and the cause of the abnormal vibration is determined; then calculating a transfer function between the control rod and a neutron detector by using a neutron transport equation, and determining the position of the control rod with abnormal vibration through a minimization algorithm theory; the problem that abnormal vibration of the in-reactor control rod cannot be monitored is solved, abnormal vibration of the control rod is avoided, and safety and reliability of the reactor control rod are powerfully guaranteed.

The invention discloses a nuclear reactor protection setting value correction method and a computer readable storage medium. The method comprises the following steps: S1, establishing a nuclear reactor model and an intermediate range detector model; s2, a fuel assembly in a reactor core of the nuclear reactor model is selected for neutron transport simulation calculation, and an intermediate range response factor obtained after fission neutrons of the fuel assembly are transmitted to the position of the intermediate range detector model from a reactor core active area is obtained; s3, calculating an intermediate range theoretical current value at the beginning of the life period and an intermediate range theoretical current value at the current burnup point according to the intermediate range response factor, and obtaining a correction coefficient at the current burnup point; and S4, according to the correction coefficient under the current burn-up point, correcting the protection constant value of the middle range set at the beginning of the life period to obtain the protection constant value of the middle range under the current burn-up point. The method guarantees the accuracy of the protection constant value of the nuclear reactor, reduces the risk of mistaken shutdown of the nuclear reactor and the risk of protection failure of the reactor, and gives consideration to the safety of a unit.

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 present invention relates to a method for calibrating an off-pile detector, comprising: S1: obtaining the detector response factor at each position in the pile by simulating the process of neutron transport from the inside of the pile to the off-pile detector; S2: simulating the theoretical xenon oscillation process to obtain the fission neutron production rate at each position in the pile during the xenon oscillation; S3: obtaining the response current distribution of each off-pile detector during the xenon oscillation according to the detector response factor and the fission neutron production rate; S4: calculating the calibration coefficient using the current distribution. While the present invention cancels the implementation of the on-site xenon oscillation test, it can still provide a complete set of calibration coefficients with sufficient accuracy, reduce the disturbance to the core, reduce the difficulty of controlling the reactor (especially at the end of its life), and thus reduce the pressure on the operating personnel.

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.

The invention relates to a fine power calculation method and device for researching a small irradiation component in a reactor and a readable storage medium. The fine power calculation method comprises the following steps: S1, acquiring parameter information of a structural component in a reactor core; s2, dividing the structural component into a plurality of homogenized grids and dividing a neutron transport calculation area; s3, calculating neutron fluence rate and power in each region; s4, calculating a power weight factor of each fuel rod; s5, judging whether all state calculation of the structural component is completed or not, and if yes, entering the step S7; s6, changing the state parameter information of the structure assembly, and calculating the neutron fluence rate and power, the homogenization cross section, the homogenization grid geometric information and the power weight factor; s7, calculating the neutron fluence rate and power of the three-dimensional fine net of the reactor core; and S8, calculating the three-dimensional power of each fuel rod of the irradiation small component. According to the method, the calculation resolution of the irradiation small component can be improved, and the safety analysis capability of the irradiation small component is improved.

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 invention discloses a hyperfine group rapid calculation method for a single group reaction section in Monte Carlo burnup calculation. The hyperfine group rapid calculation method comprises the following steps: (a) calculating each nuclide in each burnup region to generate a hyperfine group section; (b) carrying out Monte Carlo neutron transport calculation on each burnup region to obtain hyperfine group flux, wherein an energy group structure is matched with a hyperfine group cross section; and (c) using the hyperfine group flux as a weight spectrum, and merging hyperfine group sections into single group reaction sections for burn-up calculation and calling. When the hyperfine group cross section of each type of nuclide is calculated, a new background cross section algorithm is adopted, and only the single group absorption cross section of one type of nuclide needs to be calculated for each burnup area; according to the method, the calculation amount of Monte Carlo neutron transport is not additionally increased, the calculation of the hyperfine group flux and the background cross section of each burnup region is completed in a neutron transport process, and compared with a traditional reaction rate-flux ratio method for calculating the burnup cross section, the reaction rate counting time in the Monte Carlo neutron transport process is remarkably shortened, and the calculation efficiency is improved. The calculation speed can be increased in magnitude, and the method can be used for solving the calculation speed bottleneck problem that in Monte Carlo burnup calculation, burnup section calculation consumes much time.

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 invention discloses a continuous energy characteristic line neutron and photon transport calculation method applied to a nuclear-related device. The method comprises the following steps: 1, carrying out continuous energy deterministic theory neutron transport calculation, photon transport calculation and neutron and photon transport calculation on a nuclear-related facility; 2, generating characteristic line information on the basis of the geometric model of the nuclear facility, and performing source iteration flux initialization; 3, calculating the fission source of each energy section of the nuclear-related facility; 4, calculating a scattering source of each energy section of the nuclear-related facility; 5, calculating the average angular flux of the nuclear facility and the flux density of the neutron standard; 6, judging whether the flux of the ith energy section of the nuclear-related facility is converged or not, and if not, returning to the step 4; judging whether all energy segments are scanned or not, and if not, returning to the step 4; judging whether the flux of each energy section converges or not, and if not, returning to the step 4; and 7, calculating an effective value-added factor keff of the nuclear-related facility, judging whether the keff is converged or not, and if the keff is not converged, returning to the step 4. According to the method, neutron and photon transport calculation with a wide application range and high efficiency is realized.

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.