Nuclear reactor neutron transport calculation method and device

Abstract

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.

Description

Methods and apparatus for calculating neutron transport in nuclear reactors Technical Field

[0001] This application belongs to the field of nuclear power technology, and in particular relates to a method and apparatus for calculating neutron transport in nuclear reactors. Background Technology

[0002] Neutron transport calculation in nuclear reactors is a core component of nuclear reactor design, safety analysis, and operational optimization. Its purpose is to solve the neutron transport equations numerically to obtain the spatial distribution, energy distribution, and related physical parameters of the neutron standard flux within the reactor core. Among these, the fine-grid multi-group neutron transport model is widely used in high-precision computing scenarios because it can accurately describe the fine distribution of neutrons in both spatial and energy dimensions.

[0003] However, solving fine-network multi-group neutron transport models typically relies on source iteration methods, and the convergence speed of source iteration is significantly affected by the accuracy of the initial estimates. In related techniques, the initial estimates of neutron standard flux for the current iteration step are often directly derived from the calculation results of the previous iteration step or simply extrapolated, leading to slow iteration convergence. This is especially true in large and complex reactor models, where it significantly increases computation time and fails to meet the efficiency requirements of engineering applications. Summary of the Invention

[0004] This application provides a method and apparatus for calculating neutron transport in nuclear reactors, aiming to solve the problem of low computational efficiency in existing methods for calculating neutron transport in nuclear reactors.

[0005] In a first aspect, embodiments of this application provide a method for calculating neutron transport in a nuclear reactor, comprising: during the calculation of a fine-net multi-group neutron transport model of the nuclear reactor using a source iteration method, for each source iteration, spatially merging the fine-net multi-group neutron transport model to obtain a coarse-net multi-group neutron transport model; wherein each coarse mesh in the coarse-net multi-group neutron transport model is obtained by spatial merging based on at least two fine meshes in the fine-net multi-group neutron transport model; determining the first coarse-net multi-group neutron standard flux of each coarse mesh based on the initial fine-net multi-group neutron standard flux of the fine meshes within each coarse mesh; correcting the first coarse-net multi-group neutron standard flux of each coarse mesh using a coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron standard flux of each coarse mesh; solving the coarse-net multi-group neutron transport model using the second coarse-net multi-group neutron standard flux of each coarse mesh to obtain the third coarse-net multi-group neutron standard flux of each coarse mesh. Flux; if the flux of the third coarse-grid multi-group neutron scale of each coarse grid does not meet the first convergence condition, the flux of the third coarse-grid multi-group neutron scale of each coarse grid is taken as the flux of the first coarse-grid multi-group neutron scale of each coarse grid, and the process of correcting the flux of the first coarse-grid multi-group neutron scale of each coarse grid with the coarse-grid multi-group correction factor is returned to obtain the flux of the second coarse-grid multi-group neutron scale of each coarse grid, until the flux of the third coarse-grid multi-group neutron scale of each coarse grid satisfies the first convergence condition, and the target coarse-grid multi-group neutron scale flux of each coarse grid is obtained; the fine-grid multi-group correction factor is determined according to the target coarse-grid multi-group neutron scale flux of each coarse grid; the initial fine-grid multi-group neutron scale flux of each fine grid is corrected with the fine-grid multi-group correction factor to obtain the target fine-grid multi-group neutron scale flux of each fine grid; the fine-grid multi-group neutron transport model is calculated with the target fine-grid multi-group neutron scale flux of each fine grid.

[0006] In one embodiment of this application, before correcting the first coarse-net multi-group neutron scale flux of each coarse-net using a coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron scale flux of each coarse-net, the method further includes: performing energy merging on the coarse-net multi-group neutron transport model to obtain a coarse-net single-group neutron transport model; determining the first coarse-net single-group neutron scale flux of each coarse-net based on the first coarse-net multi-group neutron scale flux of each coarse-net; solving the coarse-net single-group neutron transport model using the first coarse-net single-group neutron scale flux of each coarse-net to obtain the second coarse-net single-group neutron scale flux of each coarse-net; in the... If the second coarse-net single-group neutron standard flux of each coarse-net does not meet the second convergence condition, the second coarse-net single-group neutron standard flux of each coarse-net is used as the first coarse-net single-group neutron standard flux of each coarse-net. The process of solving the coarse-net single-group neutron transport model using the first coarse-net single-group neutron standard flux of each coarse-net is then repeated until the second coarse-net single-group neutron standard flux of each coarse-net satisfies the second convergence condition, thus obtaining the target coarse-net single-group neutron standard flux of each coarse-net. Based on the target coarse-net single-group neutron standard flux, the coarse-net multi-group correction factor is determined.

[0007] In one embodiment of this application, determining the coarse mesh multi-group correction factor based on the sub-label flux of the target coarse mesh single group includes: determining the sum of the sub-label fluxes of the first coarse mesh multi-group of each coarse mesh as a first value; and determining the ratio of the sub-label flux of the target coarse mesh single group to the first value as the coarse mesh multi-group correction factor.

[0008] In one embodiment of this application, the step of spatially merging the fine-mesh multi-group neutron transport model to obtain a coarse-mesh multi-group neutron transport model includes: constructing the coarse-mesh multi-group neutron transport model based on the multi-group homogenized macroscopic cross-section of the coarse-mesh, wherein the multi-group homogenized macroscopic cross-section of the coarse-mesh is determined by spatial merging based on the fine-mesh multi-group neutron transport model; wherein the multi-group homogenized macroscopic cross-section of the coarse-mesh can be obtained through the following steps: obtaining the macroscopic cross-section of each fine mesh and the neutron transport model of each fine mesh according to the fine-mesh multi-group neutron transport model. The standard flux and the area of ​​each fine mesh are used to determine the total reaction rate of each coarse mesh. The reaction rate of each fine mesh is obtained by multiplying the macroscopic cross-section of the fine mesh, the standard flux of the fine mesh, and the area of ​​the fine mesh. The sum of the products of the standard flux of the fine mesh and the area of ​​the fine mesh in each coarse mesh is used to determine the total standard flux of each coarse mesh. The ratio of the total reaction rate of each coarse mesh to the total standard flux of each coarse mesh is used to determine the multi-group homogenized macroscopic cross-section of the coarse mesh.

[0009] In one embodiment of this application, the coarse-grid multi-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within the core of the nuclear reactor. The coarse-grid multi-group neutron transport model can be constructed through the following steps: determining the total net leakage rate within the core of the nuclear reactor based on the net neutron flux of each coarse grid and the grid width of each coarse grid; determining the total evacuation rate within the core of the nuclear reactor based on the multi-group homogenized macroscopic cross-section of the coarse grid and the first coarse-grid multi-group neutron standard flux of each coarse grid; and determining the total evacuation rate within the core of the nuclear reactor based on the source group neutron standard flux of each coarse grid and the homogenized scattering cross-section of each coarse grid. The total scattering generation rate within the reactor core; the normalized fission generation rate within the reactor core is determined based on the homogenized fission cross-section of each coarse grid, the number of neutrons within the reactor core, the eigenvalues, and the source energy group neutron standard flux of each coarse grid; the neutron loss within the reactor core is determined based on the total net leakage rate and the total removal rate; the neutron generation within the reactor core is determined based on the total scattering generation rate and the normalized fission generation rate; and a multi-group neutron transport model for the coarse grid is constructed based on the neutron loss and neutron generation within the reactor core.

[0010] In one embodiment of this application, the nuclear reactor includes multiple neutron energy groups; the step of energy merging of the coarse-grid multi-group neutron transport model to obtain a coarse-grid single-group neutron transport model includes: constructing the coarse-grid single-group neutron transport model based on the macroscopic cross-section of the coarse-grid single-group, wherein the macroscopic cross-section of the coarse-grid single-group is determined by energy merging based on the coarse-grid multi-group neutron transport model; wherein the macroscopic cross-section of the coarse-grid single-group can be obtained through the following steps: according to the coarse-grid multi-group neutron transport model, obtaining the reaction rate density and the first coarse-grid multi-group neutron standard flux of each neutron energy group in each coarse grid; determining the sum of the reaction rate densities of each neutron energy group in each coarse grid as the total reaction rate density of each coarse grid; determining the sum of the first coarse-grid multi-group neutron standard flux of each neutron energy group as the total neutron standard flux of each coarse grid; and determining the ratio of the total reaction rate density of each coarse grid to the total neutron standard flux of each coarse grid as the macroscopic cross-section of the coarse-grid single-group.

[0011] In one embodiment of this application, the coarse-grid single-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within the core of the nuclear reactor. The coarse-grid single-group neutron transport model can be constructed through the following steps: determining the net leakage rate within the core of the nuclear reactor based on the net neutron flux of each coarse grid and the grid width of each coarse grid; determining the neutron absorption rate within the core of the nuclear reactor based on the macroscopic cross-section of each coarse-grid single-group and the standard flux of the coarse-grid single-group neutrons of each coarse grid; and based on... The normalized fission production rate within the nuclear reactor core is determined by the homogenized fission cross-section of each coarse mesh, the standard flux of the first coarse mesh single-group neutron in each coarse mesh, and the eigenvalues. The neutron loss within the nuclear reactor core is determined based on the net leakage rate and the neutron absorption rate. Neutron generation within the nuclear reactor core is determined based on the normalized fission production rate. Finally, the coarse mesh single-group neutron transport model is constructed based on the neutron loss and neutron generation within the nuclear reactor core.

[0012] In one embodiment of this application, determining the fine mesh multi-group correction factor based on the target coarse mesh multi-group flux of each coarse mesh includes: determining the sum of the areas of the fine meshes within each coarse mesh as the total area of ​​each coarse mesh; performing a weighted summation of the initial fine mesh multi-group flux of the fine meshes within each coarse mesh and the area of ​​each fine mesh to obtain a second value for each coarse mesh; determining the product of the target coarse mesh multi-group flux of each coarse mesh and the total area of ​​each coarse mesh as a third value for each coarse mesh; and determining the ratio of the third value of each coarse mesh to the second value of each coarse mesh as the fine mesh multi-group correction factor.

[0013] In one embodiment of this application, determining the first coarse mesh multi-group neutron scale flux of each coarse mesh based on the initial fine mesh multi-group neutron scale flux of each fine mesh includes: obtaining the total neutron scale flux of each coarse mesh by multiplying the initial fine mesh multi-group neutron scale flux of each fine mesh by the area of ​​each fine mesh; and determining the ratio of the total neutron scale flux of each coarse mesh to the volume of each coarse mesh as the first coarse mesh multi-group neutron scale flux of each coarse mesh.

[0014] Secondly, embodiments of this application provide a nuclear reactor neutron transport calculation device, comprising: a first merging module, configured to, during the calculation of a fine-mesh multi-group neutron transport model of the nuclear reactor using a source iteration method, spatially merge the fine-mesh multi-group neutron transport model for each source iteration to obtain a coarse-mesh multi-group neutron transport model; wherein, each coarse mesh in the coarse-mesh multi-group neutron transport model is obtained by spatial merging based on at least two fine meshes in the fine-mesh multi-group neutron transport model; the first calculation... The system comprises three modules: a first coarse-grid multi-group neutron standard flux, a second calculation module, and a third calculation module. The first module is used to correct the first coarse-grid multi-group neutron standard flux of each coarse-grid network using a coarse-grid multi-group correction factor to obtain the second coarse-grid multi-group neutron standard flux of each coarse-grid network. Flux; The fourth calculation module is used to, when the sub-label flux of the third coarse mesh multi-group neutron in each coarse mesh does not meet the first convergence condition, take the sub-label flux of the third coarse mesh multi-group neutron in each coarse mesh as the sub-label flux of the first coarse mesh multi-group neutron in each coarse mesh, and return to execute the step of correcting the sub-label flux of the first coarse mesh multi-group neutron in each coarse mesh by the coarse mesh multi-group correction factor to obtain the sub-label flux of the second coarse mesh multi-group neutron in each coarse mesh, until the sub-label flux of the third coarse mesh multi-group neutron in each coarse mesh satisfies the first convergence condition. The system obtains the target coarse-net multi-group neutron standard flux of each coarse-net; the fifth calculation module is used to determine the fine-net multi-group correction factor based on the target coarse-net multi-group neutron standard flux of each coarse-net; the first correction module is used to correct the initial fine-net multi-group neutron standard flux of each fine-net using the fine-net multi-group correction factor to obtain the target fine-net multi-group neutron standard flux of each fine-net; the sixth calculation module is used to calculate the fine-net multi-group neutron transport model using the target fine-net multi-group neutron standard flux of each fine-net.

[0015] Thirdly, embodiments of this application provide an electronic device, including: a processor and a memory storing computer program instructions; the processor executes the computer program instructions to implement the nuclear reactor neutron transport calculation method as described in the first aspect.

[0016] Fourthly, embodiments of this application provide a computer-readable storage medium storing computer program instructions, which, when executed by a processor, implement the nuclear reactor neutron transport calculation method as described in the first aspect.

[0017] Fifthly, embodiments of this application provide a computer program product in which instructions, when executed by a processor of an electronic device, cause the electronic device to perform the nuclear reactor neutron transport calculation method as described in the first aspect.

[0018] The proposed method and apparatus for calculating neutron transport in nuclear reactors, in each source iteration of the nuclear reactor neutron transport simulation, firstly spatially merges the fine-mesh multi-group neutron transport model to construct a coarse-mesh multi-group neutron transport model, significantly reducing the model size and computational complexity. Based on the initial flux of each fine mesh within the coarse mesh, the first coarse-mesh multi-group neutron standard flux is determined. This first coarse-mesh multi-group neutron standard flux is then corrected using a coarse-mesh multi-group correction factor to obtain the second coarse-mesh multi-group neutron standard flux. Finally, the coarse-mesh multi-group neutron transport model is solved using the second coarse-mesh multi-group neutron standard flux to obtain the third coarse-mesh multi-group neutron transport model. The sub-coarse mesh neutron standard flux is iteratively calculated until it meets the first convergence condition, thus obtaining the target coarse mesh neutron standard flux. This effectively compensates for the approximation error introduced by spatial merging and ensures computational accuracy at the coarse mesh scale. Furthermore, a fine mesh neutron standard flux correction factor is determined based on the target coarse mesh neutron standard flux, and the initial fine mesh neutron standard flux is corrected to obtain the target fine mesh neutron standard flux. This target fine mesh neutron standard flux is then used to calculate the neutron transport model in the fine mesh neutron transport model. Therefore, by spatially merging the fine mesh, the solution cell size is significantly reduced. Combined with the coarse and fine mesh neutron standard flux correction factors, iterative calculations significantly improve the computational efficiency of the neutron transport equation while maintaining computational accuracy. Attached Figure Description

[0019] To more clearly illustrate the technical solutions of the embodiments of this application, the accompanying drawings used in the embodiments of this application will be briefly introduced below. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0020] Figure 1 is a flowchart illustrating a nuclear reactor neutron transport calculation method provided in an embodiment of this application; Figure 2 is a flowchart illustrating the implementation of the nuclear reactor neutron transport calculation method provided in an embodiment of this application; Figure 3 is a schematic diagram illustrating the iteration deviation during calculation using a nuclear reactor neutron transport calculation related method provided in an embodiment of this application; Figure 4 is a structural schematic diagram of a nuclear reactor neutron transport calculation device provided in an embodiment of this application; Figure 5 is a structural schematic diagram of an electronic device provided in an embodiment of this application. Detailed Implementation

[0021] The features and exemplary embodiments of various aspects of this application will be described in detail below. To make the objectives, technical solutions, and advantages of this application clearer, the application will be further described in detail below with reference to the accompanying drawings and specific embodiments. It should be understood that the specific embodiments described herein are only intended to explain this application and not to limit it. For those skilled in the art, this application can be implemented without some of these specific details. The following description of the embodiments is merely to provide a better understanding of this application by illustrating examples.

[0022] It should be noted that, in this document, relational terms such as "first" and "second" are used merely to distinguish one entity or operation from another, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Furthermore, the terms "comprising," "including," or any other variations thereof are intended to cover non-exclusive inclusion, such that a process, method, article, or apparatus that comprises a list of elements includes not only those elements but also other elements not expressly listed, or elements inherent to such a process, method, article, or apparatus. Without further limitations, an element defined by the phrase "comprising..." does not exclude the presence of additional identical elements in the process, method, article, or apparatus that includes said element.

[0023] In all specific embodiments of this application, when processing data related to user identity or characteristics, such as user information, user behavior data, user historical data, and user location information, user permission or consent is obtained first. Furthermore, the collection, use, and processing of this data comply with relevant laws, regulations, and standards. Additionally, when embodiments of this application require access to sensitive personal information, separate permission or consent from the user is obtained through pop-ups or redirects to confirmation pages. Only after obtaining the user's separate permission or consent is the necessary user-related data required for the proper functioning of these embodiments obtained.

[0024] In nuclear reactor transport calculations, it is assumed that neutron scattering is isotropic, meaning that the scattering intensity of neutrons is consistent in all directions. Given a macroscopic cross-section of the neutron reaction, the neutron transport problem for a nuclear reactor can be expressed by Equation 1: (1); where, Indicates the position within the solution domain Spatial angle The energy is Neutron angle flux at the location; This represents the neutron standard flux; it should be noted that... The term on the left side of the equation is actually the last sub-term of the first term on the right side of equation 1. and the last sub-item in the second item The reason why E' is used instead of E in Equation 1 is to distinguish E on the left side of the equal sign, so as to express different values ​​of the energy independent variable E.

[0025] Indicates a transport leakage item; Indicates a disappearing item; Represents the scattering source term; Indicates eigenvalues Normalized fission source term.

[0026] Furthermore, the sub-terms in equation (1) above can be simplified. Specifically, the transport leakage term and the disappearance term can be simplified into a single operator. Action on angular flux Above, that is The scattering source term can be simplified to The fission source term can be simplified to: .

[0027] In related technologies, the source iteration method is usually used to solve Equation 1 above. The calculation steps of the general source iteration algorithm are as follows: 1) Before initializing the iteration, the neutron standard flux is calculated. and eigenvalues An initial estimate is made, which can be obtained through various techniques (such as an initial guess given by the program or solving a similar diffusion problem).

[0028] 2) Given a fixed source, the transport calculation is performed using the latest neutron standard flux to calculate the scattering source and the fission source, i.e., solving the right-hand side of Equation 1 above. Then, based on the given scattering source and fission source, the neutron transport is calculated, i.e., solving the left-hand side of Equation 1 above, as shown in Equation 2 below: (2); where, This represents the neutron standard flux in the (n+1)th iteration. Let c represent the neutron standard flux in the nth iteration, and c represent the number of fission neutrons. This represents the eigenvalue of the nth iteration. Indicates the source term.

[0029] Based on the definition of standard flux, equation 2 above can be rewritten as equation 3 below: (3); Furthermore, equation 3 above can be simplified to equation 4 below: (4); where A represents the linear mapping from the flux of the nth iteration to the flux of the (n+1)th iteration.

[0030] Furthermore, by taking the difference of equation (4) above, we obtain equation (5) below: (5); 3) Update eigenvalues Standard flux And check the convergence criteria. The eigenvalue convergence criterion can be expressed by Equation 6 below, and the standard flux convergence criterion can be expressed by Equation 7 below: (6); (7); Exit the iterative calculation when both the eigenvalue convergence criterion and the scalar flux convergence criterion are satisfied; otherwise, return to step 2.

[0031] The aforementioned general methods face a major problem: when using source iteration methods, neutron transport calculations (such as the characteristic line method MOC, discrete ordinate method SN, etc.) become very time-consuming when multiple iterations are required. This problem exists in... This is especially evident in cases where the medium is strongly scattering. Furthermore, in the middle and later stages of the source iteration calculation, and The convergence of the deviations between them becomes slower, which in turn significantly increases the computational cost.

[0032] To address the problems of the prior art, this application provides a method and apparatus for calculating neutron transport in a nuclear reactor. The method for calculating neutron transport in a nuclear reactor provided in this application is described below.

[0033] Figure 1 shows a flowchart of a nuclear reactor neutron transport calculation method according to an embodiment of this application. As shown in Figure 1, the nuclear reactor neutron transport calculation method provided in this embodiment includes the following steps 101-108, wherein: Step 101, during the calculation of the fine-net multi-group neutron transport model of the nuclear reactor through source iteration, for each source iteration, the fine-net multi-group neutron transport model is spatially merged to obtain a coarse-net multi-group neutron transport model; wherein, each coarse mesh in the coarse-net multi-group neutron transport model is obtained by spatial merging based on at least two fine meshes in the fine-net multi-group neutron transport model.

[0034] In this step, spatial merging can be achieved by merging adjacent fine meshes into a single coarse mesh. The size of the coarse mesh can be dynamically adjusted according to the required computational accuracy to reduce the computational scale.

[0035] Step 102: Determine the first coarse mesh multigroup neutron standard flux of each coarse mesh based on the initial fine mesh multigroup neutron standard flux of each coarse mesh.

[0036] In this step, the initial fine mesh multi-group sub-standard flux can be obtained from the previous source iteration of this source iteration, or it can be set according to the actual situation, without specific limitations here.

[0037] The first coarse mesh multi-group neutron standard flux can be the normalized value of the neutron standard flux in the coarse mesh cell according to energy grouping. Specifically, it can be obtained by the area-weighted average of the fine mesh multi-group neutron standard flux, which is used to characterize the neutron standard flux distribution of the coarse mesh cell.

[0038] Step 103: Correct the first coarse mesh multigroup neutron standard flux of each coarse mesh using the coarse mesh multigroup correction factor to obtain the second coarse mesh multigroup neutron standard flux of each coarse mesh.

[0039] In this step, the coarse-grid single-group neutron transport model can be obtained by energy merging of the coarse-grid multi-group neutron transport model. Then, the target coarse-grid single-group neutron standard flux can be obtained by iteratively solving the coarse-grid single-group neutron standard flux. Finally, the coarse-grid multi-group correction factor can be obtained based on the target coarse-grid single-group neutron standard flux.

[0040] The second coarse-net multigroup neutron standard flux of each coarse-net can be the product of the coarse-net multigroup correction factor and the first coarse-net multigroup neutron standard flux of each coarse-net.

[0041] Step 104: Solve the coarse mesh multi-group neutron transport model using the second coarse mesh multi-group neutron standard flux of each coarse mesh to obtain the third coarse mesh multi-group neutron standard flux of each coarse mesh.

[0042] In this step, the modified second coarse-net multi-group neutron standard flux is used as the source term (or initial value for iteration) to solve the coarse-net multi-group neutron transport model once, and the third coarse-net multi-group neutron standard flux is obtained.

[0043] Step 105: If the third coarse mesh multi-group neutron standard flux of each coarse mesh does not meet the first convergence condition, the third coarse mesh multi-group neutron standard flux of each coarse mesh is used as the first coarse mesh multi-group neutron standard flux of each coarse mesh. Then, the process of correcting the first coarse mesh multi-group neutron standard flux of each coarse mesh using the coarse mesh multi-group correction factor is returned to obtain the second coarse mesh multi-group neutron standard flux of each coarse mesh. This process continues until the third coarse mesh multi-group neutron standard flux of each coarse mesh meets the first convergence condition, thus obtaining the target coarse mesh multi-group neutron standard flux of each coarse mesh.

[0044] In this step, if the flux of the third coarse mesh multi-group neutron standard of each coarse mesh does not meet the first convergence condition, then the flux of the third coarse mesh multi-group neutron standard of each coarse mesh is taken as the flux of the first coarse mesh multi-group neutron standard of each coarse mesh, and the process returns to step 103 until the flux of the third coarse mesh multi-group neutron standard of each coarse mesh meets the first convergence condition. Then, the coarse mesh iteration loop is exited, and the flux of the third coarse mesh multi-group neutron standard at this time is taken as the flux of the target coarse mesh multi-group neutron standard.

[0045] In some implementations, the first convergence condition can be that the relative change in the sub-standard flux of the coarse mesh multi-group in adjacent iteration steps is less than a first set threshold, ensuring the reliability and stability of the calculation results. The first set threshold can be set according to actual conditions and is not specifically limited here.

[0046] Step 106: Determine the fine mesh multi-group correction factor based on the target coarse mesh multi-group flux of each coarse mesh.

[0047] In this step, the fine mesh multigroup correction factor can be determined based on the area of ​​the fine mesh in each coarse mesh, the initial fine mesh multigroup neutron flux of each fine mesh, and the target coarse mesh multigroup neutron flux.

[0048] Step 107: Correct the initial sub-label flux of each fine mesh multi-group using the fine mesh multi-group correction factor to obtain the target sub-label flux of each fine mesh multi-group.

[0049] In this step, the product of the fine mesh multigroup correction factor and the initial fine mesh multigroup neutron flux of each fine mesh can be used as the target fine mesh multigroup neutron flux of each fine mesh.

[0050] Step 108: Calculate the multi-group neutron transport model of the fine mesh using the target fine mesh multi-group neutron standard flux of each fine mesh.

[0051] In this step, the updated target fine-net multi-group neutron standard flux is used to calculate the fine-net multi-group neutron transport model to update key parameters such as neutron standard flux and eigenvalues ​​of all fine-nets and energy groups. The results of this calculation are used as the input for the next source iteration of the fine-net multi-group neutron transport model to accelerate the convergence speed of the source iteration of the fine-net multi-group neutron transport model.

[0052] In this embodiment, in each source iteration of the nuclear reactor neutron transport simulation, the fine-net multi-group neutron transport model is first spatially merged to construct a coarse-net multi-group neutron transport model, significantly reducing the model size and computational complexity. Based on the initial flux of each fine-net within the coarse-net, the first coarse-net multi-group neutron standard flux is determined. This first coarse-net multi-group neutron standard flux is then corrected using a coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron standard flux. The coarse-net multi-group neutron transport model is then solved using the second coarse-net multi-group neutron standard flux to obtain the third coarse-net multi-group neutron standard flux. If the neutron standard flux of the third coarse-mesh multi-group neutron scale does not meet the first convergence condition, iterative iteration is performed to gradually satisfy the first convergence condition, obtaining the target coarse-mesh multi-group neutron standard flux. This effectively compensates for the approximation error introduced by spatial merging and ensures the computational accuracy at the coarse-mesh scale. Furthermore, a fine-mesh multi-group correction factor is determined based on the target coarse-mesh multi-group neutron standard flux, and the initial fine-mesh multi-group neutron standard flux is corrected to obtain the target fine-mesh multi-group neutron standard flux. This target fine-mesh multi-group neutron standard flux is then used for the computation of the fine-mesh multi-group neutron transport model. Thus, by spatially merging the fine mesh, the size of the solution element is significantly reduced. By combining the coarse-mesh and fine-mesh multi-group correction factors for iterative iteration, the computational efficiency of the neutron transport equation is significantly improved while maintaining computational accuracy.

[0053] In some embodiments, before step 103, the nuclear reactor neutron transport calculation method provided in this application may further include steps 201-205: Step 201, performing energy merging on the coarse mesh multi-group neutron transport model to obtain a coarse mesh single-group neutron transport model; Step 202, determining the first coarse mesh single-group neutron standard flux of each coarse mesh based on the first coarse mesh multi-group neutron standard flux of each coarse mesh; Step 203, solving the coarse mesh single-group neutron transport model using the first coarse mesh single-group neutron standard flux of each coarse mesh to obtain the second coarse mesh single-group neutron standard flux of each coarse mesh; Step 204, in the coarse mesh multi-group neutron transport model, performing energy merging on the coarse mesh multi-group neutron transport model to obtain a coarse mesh single-group neutron transport model; Step 205, determining the first coarse mesh single-group neutron standard flux of each coarse mesh multi-group neutron transport model; Step 206, solving the coarse mesh single-group neutron transport model using the first coarse mesh single-group neutron standard flux of each coarse mesh multi-group neutron transport model; Step 207, determining the second coarse mesh single-group neutron standard flux of each coarse mesh multi-group neutron transport model. If the second coarse-net single-group neutron standard flux of the network does not meet the second convergence condition, the second coarse-net single-group neutron standard flux of each coarse network is used as the first coarse-net single-group neutron standard flux of each coarse network. Then, the process of solving the coarse-net single-group neutron transport model using the first coarse-net single-group neutron standard flux of each coarse network is returned to obtain the second coarse-net single-group neutron standard flux of each coarse network. This process continues until the second coarse-net single-group neutron standard flux of each coarse network satisfies the second convergence condition, thus obtaining the target coarse-net single-group neutron standard flux of each coarse network. Step 205: Determine the coarse-net multi-group correction factor based on the target coarse-net single-group neutron standard flux.

[0054] In this implementation, energy merging can be achieved by superimposing the reaction rate density and the first coarse-grid multi-group neutron standard flux in the coarse-grid multi-group neutron transport model according to the neutron energy group to generate a coarse-grid single-group macroscopic cross section, thereby constructing a coarse-grid single-group neutron transport model.

[0055] Furthermore, the flux of the first coarse mesh multi-group neutron standard of each coarse mesh can be summed to obtain the flux of the first coarse mesh single-group neutron standard of each coarse mesh.

[0056] The coarse-grid single-group neutron standard flux of each coarse grid is used to solve the coarse-grid single-group neutron transport model to obtain the second coarse-grid single-group neutron standard flux of each coarse grid. It is then determined whether the second coarse-grid single-group neutron standard flux of each coarse grid satisfies the second convergence condition.

[0057] If the condition is not met, the neutron standard flux of the second coarse net single group is used as the neutron standard flux of the first coarse net single group, and the process returns to step 203 until the neutron standard flux of the second coarse net single group of each coarse net satisfies the second convergence condition. The neutron standard flux of the second coarse net single group at this time is used as the target coarse net single group neutron standard flux, and the coarse net multi-group correction factor is determined based on the target coarse net single group neutron standard flux.

[0058] The second convergence condition can be achieved by ensuring that the relative change in the sub-standard flux of the coarse net single group in adjacent iteration steps is less than a second set threshold, thus guaranteeing the reliability and stability of the calculation results. The second set threshold can be set according to actual conditions and is not specifically limited here.

[0059] In this implementation, the correction factor obtained through single-group iteration accurately reflects the overall trend of sub-standard flux in multiple groups, thereby reducing the number of subsequent multiple-group iterations and improving the convergence speed of the source iteration.

[0060] Further, step 205 may include: determining the sum of the sub-label fluxes of the first coarse mesh multi-group of each coarse mesh as a first value; and determining the ratio of the sub-label flux of the target coarse mesh single-group to the first value as the coarse mesh multi-group correction factor.

[0061] Specifically, the coarse web multigroup correction factor can be expressed by the following equation 8: (8); where, This represents the coarse web multi-group correction factor. This represents the target coarse net single-group neutron standard flux of coarse net i. Let i represent the first coarse net multigroup neutron standard flux of the i-th coarse net.

[0062] in, This is denoted as the first value.

[0063] In some implementations, the sub-standard flux of the second coarse mesh multi-group neutron in each coarse mesh can be calculated using the following equation 9: (9); where, Let represent the flux of the second coarse net multigroup neutron in coarse net i.

[0064] In some implementations, step 101 may include the following: constructing a coarse mesh multi-group neutron transport model based on the coarse mesh multi-group homogenized macroscopic cross section, wherein the coarse mesh multi-group homogenized macroscopic cross section is determined by spatial merging based on the fine mesh multi-group neutron transport model.

[0065] The macroscopic cross section of the coarse mesh, which is homogenized across multiple groups, can be obtained through the following steps: Based on the multi-group neutron transport model of the fine mesh, obtain the macroscopic cross section of each fine mesh, the neutron standard flux of each fine mesh, and the area of ​​each fine mesh; determine the total reactivity of each coarse mesh by summing the reactivity rates of the fine meshes within each coarse mesh, where the reactivity rate of each fine mesh is obtained by multiplying the macroscopic cross section of the fine mesh, the neutron standard flux of the fine mesh, and the area of ​​the fine mesh; determine the total neutron standard flux of each coarse mesh by summing the products of the neutron standard flux of each fine mesh and the area of ​​the fine mesh; and determine the macroscopic cross section of the coarse mesh, which is homogenized across multiple groups, by the ratio of the total reactivity of each coarse mesh to the total neutron standard flux of each coarse mesh.

[0066] In this implementation, the macroscopic cross-section, neutron standard flux, and area data of each fine mesh are first obtained. Then, the total reaction rate is obtained by summing the reaction rates of the fine meshes within each coarse mesh, where the reaction rate of an individual fine mesh is determined by the product of its macroscopic cross-section, neutron standard flux, and area. Simultaneously, the total neutron standard flux is obtained by summing the products of the neutron standard flux and area of ​​each fine mesh within each coarse mesh. Finally, the multi-group homogenized macroscopic cross-section of the coarse mesh is calculated using the ratio of the total reaction rate to the total neutron standard flux.

[0067] Specifically, the macroscopic cross-section of the coarse mesh with multiple homogenization can be represented by the following equation 10: (10); where, This represents the macroscopic cross-section of a coarse mesh with multiple groups homogenized. This represents the macroscopic cross-section of the j-th fine mesh. This represents the area of ​​the j-th fine mesh. This represents the neutron standard flux of the j-th fine mesh; This represents the reaction rate of energy group g in response to reaction x in fine mesh j.

[0068] in, Let i be the total reaction rate of the coarse mesh. Let i be the total neutron flux of the coarse net.

[0069] In this implementation, the accurate conversion of parameters from fine to coarse mesh is achieved through the principle of conservation of physical quantities. The mathematical expressions of reaction rate conservation and energy conservation provide a rigorous theoretical foundation for constructing the coarse mesh model. This reduces computational errors introduced by spatial merging and provides reliable initial parameters for subsequent iterative calculations of the coarse mesh model.

[0070] In some implementations, a coarse-grid multi-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within a nuclear reactor core. This model can be constructed through the following steps: determining the total net leakage rate within the nuclear reactor core based on the net neutron flux and grid width of each coarse grid; determining the total evacuation rate within the nuclear reactor core based on the multi-group homogenized macroscopic cross-section of the coarse grid and the first coarse-grid multi-group neutron standard flux of each coarse grid; and determining the total evacuation rate of the nuclear reactor core based on the source group neutron standard flux and homogenized scattering cross-section of each coarse grid. The total scattering generation rate within the reactor core; the normalized fission generation rate within the reactor core is determined based on the homogenized fission cross-section of each coarse grid, the number of neutrons within the reactor core, the eigenvalues, and the source group neutron standard flux of each coarse grid; the neutron loss within the reactor core is determined based on the total net leakage rate and the total removal rate; the neutron generation within the reactor core is determined based on the total scattering generation rate and the normalized fission generation rate; and a multi-group neutron transport model for the coarse grid is constructed based on the neutron loss and neutron generation within the reactor core.

[0071] Specifically, the coarse-net multi-group neutron transport model can be represented by the following equation 11: (11); where, Indicates net neutron flow. This represents the grid width in direction l; Indicates the incident neutron flow. Indicates the emitted neutron stream; This represents the overall macroscopic cross-section of the coarse mesh after multi-group homogenization. This represents the standard flux of the neutrons in the first coarse mesh multigroup. Represents the standard flux of the neutron in the energy group. Indicates the homogenized scattering cross section; This represents the homogenized fission cross section. This represents the number of neutrons released in each fission cycle. Represents the eigenvalue.

[0072] in, This is recorded as the total net leakage rate. This is denoted as the total eviction rate. This is denoted as the total scattering generation rate; It is denoted as the normalized fission generation rate.

[0073] In this implementation, Equation 11 above accurately characterizes the generation, transport, and loss processes of neutrons in a nuclear reactor, achieving precise simulation of neutron transport in a nuclear reactor. Furthermore, by constructing a coarse-grid multi-group neutron transport model, computational complexity is significantly reduced, and computational efficiency is improved.

[0074] In some implementations, the nuclear reactor includes multiple neutron energy groups. Step 201 may include: constructing a coarse-grid single-group neutron transport model based on the macroscopic cross-section of the coarse-grid single-group, wherein the macroscopic cross-section of the coarse-grid single-group is determined by energy merging based on the coarse-grid multi-group neutron transport model; wherein the macroscopic cross-section of the coarse-grid single-group can be obtained through the following steps: obtaining the reaction rate density and the first coarse-grid multi-group neutron standard flux of each neutron energy group in each coarse grid according to the coarse-grid multi-group neutron transport model; determining the sum of the reaction rate densities of each neutron energy group in each coarse grid as the total reaction rate density of each coarse grid; determining the sum of the first coarse-grid multi-group neutron standard flux of each neutron energy group as the total neutron standard flux of each coarse grid; and determining the ratio of the total reaction rate density of each coarse grid to the total neutron standard flux of each coarse grid as the macroscopic cross-section of the coarse-grid single-group.

[0075] Specifically, the macroscopic cross-section of a single coarse mesh group can be represented by the following equation 12: (12); where, This represents the macroscopic cross-section of a single coarse mesh group. This represents the neutron standard flux of the i-th fine mesh (i.e., the neutron standard flux of the first coarse mesh multigroup).

[0076] in, Let the total reaction rate density of the coarse mesh be denoted as i. Let i be the total neutron flux of the coarse net.

[0077] In some implementations, a coarse-grid single-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within a nuclear reactor core. This model can be constructed through the following steps: determining the net leakage rate within the nuclear reactor core based on the net neutron flux and grid width of each coarse grid; determining the neutron absorption rate within the nuclear reactor core based on the macroscopic cross-section and standard flux of each coarse-grid single-group neutrons; determining the normalized fission generation rate within the nuclear reactor core based on the homogenized fission cross-section, the standard flux of the first coarse-grid single-group neutrons, and eigenvalues ​​of each coarse grid; determining the neutron loss within the nuclear reactor core based on the net leakage rate and the neutron absorption rate; determining the neutron generation within the nuclear reactor core based on the normalized fission generation rate; and constructing the coarse-grid single-group neutron transport model based on the neutron loss and neutron generation within the nuclear reactor core.

[0078] Specifically, the coarse-net single-group neutron transport model can be expressed by the following equation 13: (13); where, Indicates net leakage rate. Indicates neutron absorption rate, This represents the neutron fission production rate after eigenvalue normalization (i.e., the normalized fission production rate).

[0079] In this implementation, a coarse-grid single-group neutron transport model is constructed based on a single-group homogenized macroscopic cross section. The coarse-grid single-group neutron standard flux is obtained by solving the coarse-grid single-group neutron transport model and used for subsequent calculation of the coarse-grid multi-group correction factor. Thus, through accurate energy merging, the neutron balance relationship of the single-group model is ensured to be consistent with that of the multi-group model, thereby improving the accuracy of the coarse-grid multi-group correction factor and reducing the number of source iterations.

[0080] In some implementations, step 106 may include: determining the total area of ​​each coarse mesh by summing the areas of the fine meshes within each coarse mesh; weighting and summing the initial fine mesh multi-group sub-standard flux of each fine mesh within each coarse mesh with the area of ​​each fine mesh to obtain a second value for each coarse mesh; determining a third value for each coarse mesh by multiplying the target coarse mesh multi-group sub-standard flux of each coarse mesh with the total area of ​​each coarse mesh; and determining the ratio of the third value of each coarse mesh to the second value of each coarse mesh as the fine mesh multi-group correction factor.

[0081] Specifically, the fine-network multi-group correction factor can be expressed by the following equation 14: (14); where, Let j represent the fine-network multi-group correction factor, and j represent the fine-network j. This represents the area of ​​the j-th fine mesh. This represents the target coarse mesh multi-group neutron standard flux. Let represent the initial fine-net multi-group neutron flux of fine-net j.

[0082] in, This is denoted as the second value. This is denoted as the third value.

[0083] In some implementations, the target sub-standard flux of each fine mesh in multiple groups can be calculated using the following equation 15: (15); where, Let represent the target fine network multi-group neutron flux of the j-th fine network.

[0084] Referring to Figure 2 below, the implementation flow of the nuclear reactor neutron transport calculation method provided in this application embodiment is described as follows: For the existing source iteration steps, as shown in the red part of Figure 2, the nuclear reactor neutron transport calculation method provided in this application embodiment can generate a high-quality fine-net multi-group correction factor for the fine-net multi-group neutron transport model, so as to correct the initial fine-net multi-group neutron standard flux of this source iteration, thereby improving the computational efficiency of the source iteration of the fine-net multi-group neutron transport model. Specifically: Step 1: Perform spatial merging. The fine-net multi-group neutron transport model is spatially merged through the coarse-net homogenized multi-group macroscopic cross section (i.e., the coarse-net multi-group homogenized macroscopic cross section) as shown in Equation 10 above, so as to construct the multi-group coarse-net finite difference linear system (i.e., the coarse-net multi-group neutron transport model) as shown in Equation 11 above.

[0085] Step 2: Perform energy merging. Use the coarse mesh homogenized macroscopic cross section (i.e., coarse mesh multi-group macroscopic interface) of the single group shown in Equation 12 above to perform energy merging on Equation 11 above, so as to construct a single group coarse mesh finite difference linear system (i.e., coarse mesh single group neutron transport model) as shown in Equation 13 above.

[0086] Step 3: Obtain the single-group neutron standard flux of the first coarse net based on the multi-group neutron standard flux of the first coarse net.

[0087] Step 4: Solve the coarse-net single-group neutron standard flux using the first coarse-net single-group neutron standard flux to obtain the second coarse-net single-group neutron standard flux.

[0088] Step 5: If the flux of the second coarse net single group neutron scale does not meet the convergence criterion (i.e., the second convergence condition), repeat steps 3 to 4 until the flux of the second coarse net single group neutron scale meets the convergence criterion, obtain the flux of the target coarse net single group neutron scale, and calculate the eigenvalues.

[0089] Step 6: As shown in Equation 8 above, determine the coarse mesh multi-group correction factor based on the target coarse mesh single-group sub-standard flux.

[0090] Step 7: As shown in Equation 9 above, the first coarse mesh multi-group neutron standard flux of each coarse mesh is corrected by the coarse mesh multi-group correction factor to obtain the second coarse mesh multi-group neutron standard flux, and the estimated values ​​of fission source and eigenvalue are updated to solve the multi-group coarse mesh finite difference linear system to obtain the third coarse mesh multi-group neutron standard flux.

[0091] Step 8: If the flux of the third coarse mesh multigroup neutron scale does not meet the convergence criterion (i.e., the first convergence condition), repeat steps 3 to 7 until the flux of the third coarse mesh multigroup neutron scale meets the convergence criterion, and obtain the flux of the target coarse mesh multigroup neutron scale.

[0092] Step 9: As shown in Equation 14 above, determine the fine-net multi-group correction factor based on the target coarse-net multi-group flux.

[0093] Step 10: As shown in Equation 15 above, the initial fine-net multi-group neutron standard flux is corrected by the fine-net multi-group correction factor to obtain the target fine-net multi-group neutron standard flux. The target fine-net multi-group neutron standard flux is then substituted into the source iteration (i.e., Equations 2 to 5 above) calculated by the fine-net multi-group neutron transport equation (i.e., the fine-net multi-group neutron transport model shown in Equation 1 above), thereby accelerating the convergence process of the source iteration.

[0094] The above embodiments are illustrated below through experimental verification: In related technologies, and The deviation between them gradually decreases, as shown in Figure 3. As shown in Table 1 below, the nuclear reactor neutron transport calculation method provided in this application provides better estimates, significantly shortens the number of convergences required for source iteration, and greatly improves computational efficiency. At the same time, the computation time of the acceleration algorithm accounts for a relatively low proportion, thus improving the overall computational efficiency.

[0095] Table 1

[0096] Figure 4 shows a structural diagram of the nuclear reactor neutron transport computing device provided in an embodiment of this application. As shown in Figure 4, the nuclear reactor neutron transport computing device 400 includes: a first merging module 401, used to spatially merge the fine-mesh multi-group neutron transport model of the nuclear reactor for each source iteration during the calculation of the fine-mesh multi-group neutron transport model of the nuclear reactor through source iteration, to obtain a coarse-mesh multi-group neutron transport model; wherein, each coarse mesh in the coarse-mesh multi-group neutron transport model is obtained by spatial merging based on at least two fine meshes in the fine-mesh multi-group neutron transport model; a first calculation module 402, used for... Based on the initial fine-net multi-group neutron standard flux of each coarse mesh, the first coarse mesh multi-group neutron standard flux of each coarse mesh is determined; the second calculation module 403 is used to correct the first coarse mesh multi-group neutron standard flux of each coarse mesh using a coarse mesh multi-group correction factor to obtain the second coarse mesh multi-group neutron standard flux of each coarse mesh; the third calculation module 404 is used to solve the coarse mesh multi-group neutron transport model using the second coarse mesh multi-group neutron standard flux of each coarse mesh to obtain the third coarse mesh multi-group neutron standard flux of each coarse mesh. The fourth calculation module 405 is used to, when the sub-scale flux of the third coarse mesh multi-group neutron in each coarse mesh does not meet the first convergence condition, take the sub-scale flux of the third coarse mesh multi-group neutron in each coarse mesh as the sub-scale flux of the first coarse mesh multi-group neutron in each coarse mesh, and return to execute the step of correcting the sub-scale flux of the first coarse mesh multi-group neutron in each coarse mesh using the coarse mesh multi-group correction factor to obtain the sub-scale flux of the second coarse mesh multi-group neutron in each coarse mesh, until the sub-scale flux of the third coarse mesh multi-group neutron in each coarse mesh satisfies the first convergence condition, and obtain the... The target coarse-net multi-group neutron standard flux of each coarse-net; the fifth calculation module 406, used to determine the fine-net multi-group correction factor based on the target coarse-net multi-group neutron standard flux of each coarse-net; the first correction module 407, used to correct the initial fine-net multi-group neutron standard flux of each fine-net using the fine-net multi-group correction factor, to obtain the target fine-net multi-group neutron standard flux of each fine-net; the sixth calculation module 408, used to calculate the fine-net multi-group neutron transport model using the target fine-net multi-group neutron standard flux of each fine-net.

[0097] The nuclear reactor neutron transport calculation device 400 provided in this application embodiment can realize the various processes implemented in the aforementioned nuclear reactor neutron transport calculation method embodiment and achieve the same technical effect. To avoid repetition, it will not be described again here.

[0098] Figure 5 shows a schematic diagram of the hardware structure of the electronic device provided in an embodiment of this application.

[0099] Electronic devices may include a processor 501 and a memory 402 storing computer program instructions.

[0100] Specifically, the processor 501 may include a central processing unit (CPU), an application-specific integrated circuit (ASIC), or one or more integrated circuits that can be configured to implement the embodiments of this application.

[0101] Memory 402 may include mass storage for data or instructions. For example, and not limitingly, memory 402 may include a hard disk drive (HDD), floppy disk drive, flash memory, optical disk, magneto-optical disk, magnetic tape, or Universal Serial Bus (USB) drive, or a combination of two or more of these. Where appropriate, memory 402 may include removable or non-removable (or fixed) media. Where appropriate, memory 402 may be internal or external to the integrated gateway disaster recovery device. In a particular embodiment, memory 402 is non-volatile solid-state memory.

[0102] Memory may include read-only memory (ROM), random access memory (RAM), disk storage media devices, optical storage media devices, flash memory devices, and electrical, optical, or other physical / tangible memory storage devices. Therefore, typically, memory includes one or more tangible (non-transitory) computer-readable storage media (e.g., memory devices) encoded with software including computer-executable instructions, and when the software is executed (e.g., by one or more processors), it is operable to perform the operations described with reference to the methods according to the first or second aspect of this disclosure.

[0103] The processor 501 implements any of the information auditing methods described in the above embodiments by reading and executing computer program instructions stored in the memory 502.

[0104] In one example, the electronic device may also include a communication interface 503 and a bus 510. As shown in Figure 5, the processor 501, memory 502, and communication interface 503 are connected via the bus 510 and communicate with each other.

[0105] The communication interface 503 is mainly used to realize communication between various modules, devices, units and / or equipment in the embodiments of this application.

[0106] Bus 510 includes hardware, software, or both, that couples components of an information auditing method or verification device together. For example, and not as a limitation, the bus may include an Accelerated Graphics Port (AGP) or other graphics bus, an Enhanced Industry Standard Architecture (EISA) bus, a Front Side Bus (FSB), HyperTransport (HT) interconnect, an Industry Standard Architecture (ISA) bus, an Infinite Bandwidth Interconnect, a Low Pin Count (LPC) bus, a memory bus, a Microchannel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a PCI-Express (PCI-X) bus, a Serial Advanced Technology Attachment (SATA) bus, a Video Electronics Standards Association Local (VLB) bus, or other suitable buses, or combinations of two or more of these. Where appropriate, bus 510 may include one or more buses. Although specific buses are described and illustrated in embodiments of this application, this application contemplates any suitable bus or interconnect.

[0107] Furthermore, in conjunction with the nuclear reactor neutron transport calculation method in the above embodiments, this application embodiment can provide a computer storage medium for implementation. This computer storage medium stores computer program instructions; when these computer program instructions are executed by a processor, they implement any of the nuclear reactor neutron transport calculation methods in the above embodiments.

[0108] Alternatively, this application embodiment can provide a computer program product for implementation, wherein when the instructions in the computer program product are executed by the processor of an electronic device, the electronic device implements any one of the nuclear reactor neutron transport calculation methods in the above embodiments.

[0109] It should be clarified that this application is not limited to the specific configurations and processes described above and shown in the figures. For the sake of brevity, detailed descriptions of known methods are omitted here. In the above embodiments, several specific steps are described as examples. However, the method process of this application is not limited to the specific steps described. Those skilled in the art can make various changes, modifications, and additions, or change the order of steps, after understanding the spirit of this application.

[0110] The functional blocks shown in the above-described structural diagram can be implemented as hardware, software, firmware, or a combination thereof. When implemented in hardware, they can be, for example, electronic circuits, application-specific integrated circuits (ASICs), appropriate firmware, plug-ins, function cards, etc. When implemented in software, the elements of this application are programs or code segments used to perform the required tasks. Programs or code segments can be stored on a machine-readable medium or transmitted over a transmission medium or communication link via data signals carried on a carrier wave. "Machine-readable medium" can include any medium capable of storing or transmitting information. Examples of machine-readable media include electronic circuits, semiconductor memory devices, ROM, flash memory, erasable ROM (EROM), floppy disks, CD-ROMs, optical disks, hard disks, fiber optic media, radio frequency (RF) links, etc. Code segments can be downloaded via computer networks such as the Internet, intranets, etc.

[0111] It should also be noted that the exemplary embodiments mentioned in this application describe methods or systems based on a series of steps or apparatus. However, this application is not limited to the order of the above steps; that is, the steps can be performed in the order mentioned in the embodiments, or in a different order, or several steps can be performed simultaneously.

[0112] The aspects of this disclosure have been described above with reference to flowchart illustrations and / or block diagrams of methods, apparatus (systems), and computer program products according to embodiments of this disclosure. It should be understood that each block in the flowchart illustrations and / or block diagrams, and combinations of blocks in the flowchart illustrations and / or block diagrams, can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable data processing apparatus to produce a machine such that these instructions, executable via the processor of the computer or other programmable data processing apparatus, enable the implementation of the functions / actions specified in one or more blocks of the flowchart illustrations and / or block diagrams. Such a processor can be, but is not limited to, a general-purpose processor, a special-purpose processor, a special application processor, or a field-programmable logic circuit. It is also understood that each block in the block diagrams and / or flowcharts, and combinations of blocks in the block diagrams and / or flowcharts, can also be implemented by special-purpose hardware performing the specified functions or actions, or can be implemented by a combination of special-purpose hardware and computer instructions.

[0113] The above description is merely a specific implementation of this application. Those skilled in the art will clearly understand that, for the sake of convenience and brevity, the specific working processes of the systems, modules, and units described above can be referred to the corresponding processes in the foregoing method embodiments, and will not be repeated here. It should be understood that the protection scope of this application is not limited thereto. Any person skilled in the art can easily conceive of various equivalent modifications or substitutions within the technical scope disclosed in this application, and these modifications or substitutions should all be covered within the protection scope of this application.

Claims

1. A method for calculating neutron transport in a nuclear reactor, characterized in that, include: In the process of calculating the fine-net multi-group neutron transport model of the nuclear reactor using a source iteration method, for each source iteration, the fine-net multi-group neutron transport model is spatially merged to obtain a coarse-net multi-group neutron transport model; wherein, each coarse mesh in the coarse-net multi-group neutron transport model is obtained by spatial merging at least two fine meshes in the fine-net multi-group neutron transport model; based on the initial fine-net multi-group neutron standard flux of the fine meshes within each coarse mesh, the first coarse-net multi-group neutron standard flux of each coarse mesh is determined; the first coarse-net multi-group neutron standard flux of each coarse mesh is corrected by a coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron standard flux of each coarse mesh; the coarse-net multi-group neutron transport model is solved using the second coarse-net multi-group neutron standard flux of each coarse mesh to obtain the third coarse-net multi-group neutron standard flux of each coarse mesh; in the third coarse-net multi-group neutron transport model of each coarse mesh... If the neutron standard flux of the group does not meet the first convergence condition, the third coarse-net multi-group neutron standard flux of each coarse-net is taken as the first coarse-net multi-group neutron standard flux of each coarse-net. The process of correcting the first coarse-net multi-group neutron standard flux of each coarse-net using the coarse-net multi-group correction factor is then performed to obtain the second coarse-net multi-group neutron standard flux of each coarse-net. This process continues until the third coarse-net multi-group neutron standard flux of each coarse-net satisfies the first convergence condition, thus obtaining the target coarse-net multi-group neutron standard flux of each coarse-net. Based on the target coarse-net multi-group neutron standard flux of each coarse-net, the fine-net multi-group correction factor is determined. The initial fine-net multi-group neutron standard flux of each fine-net is corrected using the fine-net multi-group correction factor to obtain the target fine-net multi-group neutron standard flux of each fine-net. The fine-net multi-group neutron transport model is then calculated using the target fine-net multi-group neutron standard flux of each fine-net.

2. The method according to claim 1, characterized in that, Before correcting the first coarse-net multi-group neutron scale flux of each coarse-net using a coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron scale flux of each coarse-net, the method further includes: performing energy merging on the coarse-net multi-group neutron transport model to obtain a coarse-net single-group neutron transport model; determining the first coarse-net single-group neutron scale flux of each coarse-net based on the first coarse-net multi-group neutron scale flux of each coarse-net; solving the coarse-net single-group neutron transport model using the first coarse-net single-group neutron scale flux of each coarse-net to obtain the second coarse-net single-group neutron scale flux of each coarse-net; and in the first coarse-net multi-group neutron transport model of each coarse-net, the method further includes: performing energy merging on the coarse-net multi-group neutron transport model to obtain a coarse-net single-group neutron transport model; determining the first coarse-net single-group neutron scale flux of each coarse-net based on the first coarse-net multi-group neutron scale flux of each coarse-net; solving the coarse-net single-group neutron transport model using the first coarse-net single-group neutron scale flux of each coarse-net; and in the first coarse-net multi-group neutron scale flux of each coarse-net multi-group neutron transport model, the method further includes: performing energy merging on the coarse-net multi-group neutron transport model to obtain a coarse-net single-group neutron transport model to obtain a coarse-net single-group neutron scale flux of each coarse-net multi-group neutron transport model to obtain a coarse-net single-group neutron scale flux of each coarse-net multi-group neutron transport model to obtain a coarse If the neutron standard flux of a single coarse-grid single-group does not meet the second convergence condition, the second coarse-grid single-group neutron standard flux of each coarse-grid is used as the first coarse-grid single-group neutron standard flux of each coarse-grid. The process of solving the coarse-grid single-group neutron transport model using the first coarse-grid single-group neutron standard flux of each coarse-grid is then repeated until the second coarse-grid single-group neutron standard flux of each coarse-grid satisfies the second convergence condition, thus obtaining the target coarse-grid single-group neutron standard flux of each coarse-grid. Based on the target coarse-grid single-group neutron standard flux, the coarse-grid multi-group correction factor is determined.

3. The method according to claim 2, characterized in that, The step of determining the coarse mesh multi-group correction factor based on the target coarse mesh single-group sub-label flux includes: determining the sum of the first coarse mesh multi-group sub-label fluxes of each coarse mesh as a first value; and determining the ratio of the target coarse mesh single-group sub-label flux to the first value as the coarse mesh multi-group correction factor.

4. The method according to claim 1, characterized in that, The spatial merging of the fine-mesh multi-group neutron transport model to obtain the coarse-mesh multi-group neutron transport model includes: constructing the coarse-mesh multi-group neutron transport model based on the multi-group homogenized macroscopic cross-section of the coarse-mesh, wherein the multi-group homogenized macroscopic cross-section of the coarse-mesh is determined by spatial merging based on the fine-mesh multi-group neutron transport model; wherein the multi-group homogenized macroscopic cross-section of the coarse-mesh can be obtained through the following steps: obtaining the macroscopic cross-section of each fine mesh, the neutron standard flux of each fine mesh, and so on, according to the fine-mesh multi-group neutron transport model. The area of ​​each fine mesh is determined; the sum of the reaction rates of the fine meshes within each coarse mesh is determined as the total reaction rate of each coarse mesh, the reaction rate of the fine mesh is obtained by multiplying the macroscopic cross-section of the fine mesh, the neutron standard flux of the fine mesh, and the area of ​​the fine mesh; the sum of the products of the neutron standard flux of the fine meshes within each coarse mesh and the area of ​​the fine mesh is determined as the total neutron standard flux of each coarse mesh; the ratio of the total reaction rate of each coarse mesh to the total neutron standard flux of each coarse mesh is used to determine the multi-group homogenized macroscopic cross-section of the coarse mesh.

5. The method according to claim 4, characterized in that, The coarse-grid multi-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within the nuclear reactor core. The coarse-grid multi-group neutron transport model can be constructed through the following steps: determining the total net leakage rate within the nuclear reactor core based on the net neutron flux and grid width of each coarse grid; determining the total removal rate within the nuclear reactor core based on the multi-group homogenized macroscopic cross section of the coarse grid and the first coarse-grid multi-group neutron standard flux of each coarse grid; determining the total scattering generation rate within the nuclear reactor core based on the source group neutron standard flux and homogenized scattering cross section of each coarse grid; determining the normalized fission generation rate within the nuclear reactor core based on the homogenized fission cross section of each coarse grid, the number of neutrons in the nuclear reactor core, eigenvalues, and the source group neutron standard flux of each coarse grid; and determining the neutron loss within the nuclear reactor core based on the total net leakage rate and the total removal rate. Based on the total scattering generation rate and the normalized fission generation rate, the neutron generation in the nuclear reactor core is determined; based on the neutron loss and neutron generation in the nuclear reactor core, the coarse-grid multi-group neutron transport model is constructed.

6. The method according to claim 2, characterized in that, The nuclear reactor includes multiple neutron energy groups; the energy merging of the coarse-grid multi-group neutron transport model to obtain a coarse-grid single-group neutron transport model includes: constructing the coarse-grid single-group neutron transport model based on the macroscopic cross-section of the coarse-grid single-group, wherein the macroscopic cross-section of the coarse-grid single-group is determined by energy merging based on the coarse-grid multi-group neutron transport model; wherein the macroscopic cross-section of the coarse-grid single-group can be obtained through the following steps: according to the coarse-grid multi-group neutron transport model, obtaining the reaction rate density and the first coarse-grid multi-group neutron standard flux of each neutron energy group in each coarse grid; determining the sum of the reaction rate densities of each neutron energy group in each coarse grid as the total reaction rate density of each coarse grid; determining the sum of the first coarse-grid multi-group neutron standard flux of each neutron energy group as the total neutron standard flux of each coarse grid; determining the ratio of the total reaction rate density of each coarse grid to the total neutron standard flux of each coarse grid as the macroscopic cross-section of the coarse-grid single-group.

7. The method according to claim 6, characterized in that, The coarse-grid single-group neutron transport model is used to characterize the balance between neutron loss and neutron generation within the nuclear reactor core. The coarse-grid single-group neutron transport model can be constructed through the following steps: determining the net leakage rate within the nuclear reactor core based on the net neutron flux and grid width of each coarse grid; determining the neutron absorption rate within the nuclear reactor core based on the macroscopic cross-section and standard flux of each coarse-grid single-group neutrons; determining the normalized fission generation rate within the nuclear reactor core based on the homogenized fission cross-section, standard flux of the first coarse-grid single-group neutrons, and eigenvalues ​​of each coarse grid; determining the neutron loss within the nuclear reactor core based on the net leakage rate and the neutron absorption rate; determining the neutron generation within the nuclear reactor core based on the normalized fission generation rate; and constructing the coarse-grid single-group neutron transport model based on the neutron loss and neutron generation within the nuclear reactor core.

8. The method according to claim 1, characterized in that, The step of determining the fine-mesh multi-group correction factor based on the target coarse-mesh multi-group flux of each coarse-mesh includes: determining the total area of ​​each coarse-mesh by summing the areas of the fine-mesh multi-groups within each coarse-mesh; performing a weighted summation of the initial fine-mesh multi-group flux of the fine-mesh multi-groups within each coarse-mesh and the area of ​​each fine-mesh to obtain a second value for each coarse-mesh; determining a third value for each coarse-mesh by multiplying the target coarse-mesh multi-group flux of each coarse-mesh and the total area of ​​each coarse-mesh; and determining the ratio of the third value of each coarse-mesh to the second value of each coarse-mesh as the fine-mesh multi-group correction factor.

9. The method according to claim 1, characterized in that, The step of determining the first coarse-net multi-group neutron standard flux of each coarse mesh based on the initial fine mesh multi-group neutron standard flux of each fine mesh includes: obtaining the total neutron standard flux of each coarse mesh by multiplying the initial fine mesh multi-group neutron standard flux of each fine mesh by the area of ​​each fine mesh; and determining the ratio of the total neutron standard flux of each coarse mesh to the volume of each coarse mesh as the first coarse-net multi-group neutron standard flux of each coarse mesh.

10. A nuclear reactor neutron transport computing device, characterized in that, include: The first merging module is used to spatially merge the fine-net multi-group neutron transport model of the nuclear reactor for each source iteration during the calculation of the fine-net multi-group neutron transport model using the source iteration method, to obtain a coarse-net multi-group neutron transport model; wherein, each coarse mesh in the coarse-net multi-group neutron transport model is obtained by spatial merging based on at least two fine meshes in the fine-net multi-group neutron transport model; the first calculation module is used to determine the first coarse-net multi-group neutron standard flux of each coarse mesh based on the initial fine-net multi-group neutron standard flux of the fine meshes within each coarse mesh; the second calculation module is used to correct the first coarse-net multi-group neutron standard flux of each coarse mesh using a coarse-net multi-group correction factor, to obtain the second coarse-net multi-group neutron standard flux of each coarse mesh; the third calculation module is used to solve the coarse-net multi-group neutron transport model using the second coarse-net multi-group neutron standard flux of each coarse mesh, to obtain the third coarse-net multi-group neutron standard flux of each coarse mesh; the fourth calculation module is used to... If the third coarse-net multi-group neutron scale flux of each coarse mesh does not meet the first convergence condition, the third coarse-net multi-group neutron scale flux of each coarse mesh is taken as the first coarse-net multi-group neutron scale flux of each coarse mesh. The process then returns to correcting the first coarse-net multi-group neutron scale flux of each coarse mesh using the coarse-net multi-group correction factor to obtain the second coarse-net multi-group neutron scale flux of each coarse mesh. This process continues until the third coarse-net multi-group neutron scale flux of each coarse mesh satisfies the first convergence condition, thus obtaining the flux of each coarse mesh. The system includes: a target coarse-grid multi-group neutron standard flux; a fifth calculation module, used to determine the fine-grid multi-group correction factor based on the target coarse-grid multi-group neutron standard flux of each coarse-grid; a first correction module, used to correct the initial fine-grid multi-group neutron standard flux of each fine-grid using the fine-grid multi-group correction factor, to obtain the target fine-grid multi-group neutron standard flux of each fine-grid; and a sixth calculation module, used to calculate the fine-grid multi-group neutron transport model using the target fine-grid multi-group neutron standard flux of each fine-grid.

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