Nuclear reactor few-group diffusion coefficient solving method based on anisotropic neutron transport
By directly solving the multigroup diffusion coefficient and performing transport corrections based on anisotropic P1 neutron transport calculations, the problem of limited accuracy in traditional methods is solved, and higher accuracy nuclear reactor core diffusion calculations are achieved.
Patent Information
- Authority / Receiving Office
- CN · China
- Patent Type
- Applications(China)
- Current Assignee / Owner
- XI AN JIAOTONG UNIV
- Filing Date
- 2026-02-02
- Publication Date
- 2026-05-01
AI Technical Summary
In the physical analysis of nuclear reactor cores, existing technologies suffer from limited accuracy of minority group diffusion coefficients due to approximation errors introduced by traditional transport correction methods, making it difficult to fully inherit the leakage and anisotropic effects of transport solutions.
A method based on anisotropic P1 neutron transport calculation is adopted, which solves the multigroup diffusion coefficient by accurate energy spectrum and performs transport correction, thereby reducing calculation assumptions and improving calculation accuracy.
This improves the calculation accuracy and engineering applicability of the minority group diffusion coefficient, and enhances the accuracy and efficiency of core diffusion calculation.
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Figure CN121959958A_ABST
Abstract
Description
A method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport. Technical Field
[0001] This invention relates to the field of nuclear reactor physics calculation and analysis technology, specifically to a method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport. Background Technology
[0002] In two-step software for nuclear reactor core physics analysis, a refined energy spectrum and flux distribution are typically obtained first through neutron transport calculations at the component level. Then, component homogenization generates minority group constants for solving the core diffusion equations. The accuracy of the minority group parameters directly determines the predictive power of core diffusion calculations for key results such as eigenvalues, power distribution, and reaction rate. Among these, the minority group diffusion coefficient is a crucial parameter characterizing neutron leakage capability and flux gradient response. In engineering practice, the multi-group diffusion coefficient is often derived from the transport cross-section under the Fick's law approximation.
[0003] To effectively reflect the impact of anisotropic scattering on neutron transport in diffusion models, existing component programs typically employ transport correction methods to adjust the total cross-section and the zero-order self-scattering cross-section of the material region before transport calculations or during the cross-section processing stage. This forms the transport cross-section for diffusion calculations and further yields the minority group diffusion coefficient. However, traditional transport correction methods assume the separation of the spatial and energy variables of neutron flux density, use numerical methods to solve for the first-order neutron flux moment, and then calculate the correction amount based on the assumption that the first-order scattering source is zero. This correction is then used to adjust the total cross-section and the zero-order self-scattering cross-section of the material region.
[0004] However, the aforementioned transport correction method introduces approximation errors due to the separation of the spatial and energy variables of neutron flux density, which limits the accuracy of the minority group diffusion coefficient output by component homogenization. This makes it difficult for subsequent core diffusion calculations to fully inherit the leakage and anisotropic effects of the transport solution. Therefore, it is necessary to provide a more direct and accurate method for solving the minority group diffusion coefficient, taking into account anisotropic neutron transport effects, to improve the calculation accuracy and engineering applicability of the minority group diffusion coefficient in the two-step method system. Summary of the Invention
[0005] To address the problems existing in current transport correction techniques, the present invention aims to provide a method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport. This method utilizes the accurate energy spectrum obtained from anisotropic P1 neutron transport calculations to solve for the minority group diffusion coefficient and performs transport corrections to obtain the transport cross section and the corrected zero-order self-scattering cross section of the material region. This leads to the calculation of isotropic neutron transport with transport corrections, followed by leakage correction and component homogenization to obtain an accurate minority group diffusion coefficient, thereby improving the accuracy of core diffusion calculations.
[0006] To achieve the above objectives, the present invention adopts the following technical solution: a method for solving the few-group diffusion coefficient of a nuclear reactor based on anisotropic neutron transport, comprising the following steps: Step 1: Based on the geometric dimensions and material information of the nuclear reactor fuel assembly, a physical model of the fuel assembly is constructed, and the material region and the flat-source region are meshed. A multi-group macroscopic cross section of the material region is generated from a multi-group cross section library; Step 2: Anisotropic P1 neutron transport is calculated using the multi-group macroscopic cross section of the material region to obtain the neutron flux density in the flat-source region; this step, by considering the anisotropic P1 neutron transport calculation of the anisotropic scattering source term, obtains a more accurate neutron flux density in the flat-source region, which is used in step... In step 3, the neutron flux density in the flat source region is normalized to the component scale according to the volume weight of the flat source region to obtain the component neutron spectrum, thereby reducing the influence of the spectrum deviation on the solution of the multi-group diffusion coefficient and improving the accuracy of the multi-group diffusion coefficient calculation results. In addition, the solution relationship determined by the formula derivation can reduce the calculation steps, thereby improving the computational efficiency of anisotropic P1 neutron transport calculation. Step 3: The neutron flux density in the flat source region is normalized to the component scale according to the volume weight of the flat source region to obtain the component neutron spectrum; then, a linear equation system for solving the multi-group diffusion coefficient is constructed, and the multi-group diffusion coefficient is solved; based on the component neutron spectrum, the total cross section of the material region and the first-order scattering cross section, the multi-group diffusion coefficient is established according to the neutron flow equation in the P1 equation system. Solving the linear equations can reduce the approximation steps introduced by additional assumptions, thereby improving the accuracy of the multi-group diffusion coefficient solution. Step 4: Use the obtained multi-group diffusion coefficient to perform transport correction on the total cross section and the zero-order self-scattering cross section of the material region, correcting the total cross section of the material region to the transport cross section of the material region and obtaining the corrected zero-order self-scattering cross section of the material region. Based on the multi-group diffusion coefficient obtained in Step 3, solve for the transport cross section of the material region according to Fick's law to replace the total cross section of the material region, completing the transport correction of the total cross section of the material region, and simultaneously correcting the self-scattering cross section of the material region to ensure the physical consistency of each cross section of the material region and reduce numerical instability caused by inconsistencies in the cross sections of the material region. 5: Using the zero-order self-scattering cross section and transport cross section of the material region obtained from the transport correction in step 4, as well as the remaining uncorrected multi-group macroscopic cross sections of the material region, perform isotropic neutron transport calculations with transport correction to obtain the neutron flux density in the flat source region. This step obtains the neutron flux density in the flat source region that matches the transport cross section and the corrected zero-order self-scattering cross section of the material region, providing a basis for subsequent leakage correction and assembly homogenization. Step 6: Perform leakage correction on the neutron flux density in the flat source region obtained in step 5 to obtain the corrected neutron flux density in the flat source region. Use the corrected neutron flux density in the flat source region to homogenize the multi-group diffusion coefficient of the fuel assembly and determine the small-group diffusion coefficient of the fuel assembly.
[0007] Preferably, the zeroth moment of the average angular flux density moment corresponding to the neutron flux density in the flat source region obtained in step 2 is the average angular flux density moment with a spherical harmonic expansion order of zero. The average angular flux density moment is a physical quantity involved in the iterative calculation in the anisotropic P1 neutron transport calculation, and the calculation formula is shown in the following equation:
[0008] In the formula: i represents the index of the flat source region; g represents the index of the energy group after scattering; l represents the index of the order of the spherical harmonic function expansion; m represents the index of the order of the spherical harmonic function expansion; j represents the index of the angle; k represents the index of the characteristic line. —The average angular flux density moment of the spherical harmonic function expansion order l and expansion degree m in the g group of the i-th flat source region; —The product of the angle weight of the k-th feature line with angle number j and the feature line width; —The total source term of the g-th group in the i-th flat source region, with angle number j; —The length of the k-th characteristic line with angle number j in the i-th flat source region; —The g-group incident angular flux of the i-th flat source region with angle number j and the k-th characteristic line; —The total cross section of the g-th group in the i-th flat source region; —The expanded moments of a spherical harmonic function with order l, degree m, and angle number j; —Volume of the i-th flat source region.
[0009] Preferably, the linear equation system for solving the multi-group diffusion coefficient in step 3 is constructed as follows: The neutron flow equation in the P1 equation system is shown in the following equation:
[0010] In the formula: r represents the spatial location; g' represents the index of the energy group before scattering; —The spatial gradient of the neutron flux density of the g-th group at position r; —The total cross section of the g-th group at position r; —The neutron flux of the g-th group at position r; —The neutron flux of the g'th group at position r; —The first-order scattering cross section at position r from the g' group to the g group; based on the neutron flux equation in the P1 equation set, the following equation is derived using the assumptions of Fick's law:
[0011] In the formula: —The spatial gradient of the neutron flux density of the g' group at position r; —The multigroup diffusion coefficient of the g-th group at position r; —The multigroup diffusion coefficient of the g'th group at position r; then, by integrating in the spatial domain, the multigroup diffusion coefficient is obtained by solving a system of linear equations. This system of equations uses the neutron spectrum of the component, the total cross section of the material region, and the first-order scattering cross section as parameters. The solution to the linear equations for the multigroup diffusion coefficient is shown in the following equation:
[0012] In the formula: — Neutron flux density of the g-th group; — Neutron flux density of the g' group.
[0013] Preferably, the transport correction in step 4 is specifically implemented as follows: Based on the assumptions of Fick's law, the transport cross-section of the material region is calculated using the multigroup diffusion coefficient obtained in step 3. The transport cross-section of the material region is then used to replace the total cross-section of the material region, thus completing the transport correction of the total cross-section of the material region. The formula for calculating the transport cross-section of the material region is shown below:
[0014] —The transport cross section of the g-th group at position r; When correcting the zero-order self-scattering cross section of the material region, let the correction amount be the difference between the total cross section of the material region and the transport cross section of the material region. Subtract this correction amount from the zero-order self-scattering cross section of the material region to obtain the corrected zero-order self-scattering cross section of the material region; The transport correction formula is shown in the following equation:
[0015] In the formula: —The zero-order self-scattering cross section at position r.
[0016] Compared with existing technologies, this invention has the following advantages: The present invention, based on anisotropic neutron transport, provides a method for solving the few-group diffusion coefficient of nuclear reactors. First, it calculates the neutron spectrum of the assembly using anisotropic P1 neutron transport to solve for the many-group diffusion coefficient. This method considers anisotropic scattering source terms, reducing the deviation in the assembly neutron spectrum caused by the approximation of scattering source terms. Furthermore, the present invention derives and directly solves a linear equation system for solving the many-group diffusion coefficient based on the neutron flux equation in the P1 equation set, reducing the approximation steps introduced by additional assumptions. Therefore, it improves the accuracy of the results compared to traditional methods for solving the many-group diffusion coefficient. Second, it uses the solved many-group diffusion coefficient for transport correction, improving the accuracy of the scattering and transport cross sections after the transport correction and maintaining the consistency of the cross section in the material region. Further, it uses the isotropic neutron transport with transport correction to calculate the neutron flux density in the flat-source region, and obtains the few-group diffusion coefficient through leakage correction and assembly homogenization. Thus, the present invention improves the overall accuracy of solving the few-group diffusion coefficient and has engineering application value. Attached Figure Description
[0017] Figure 1 is a flowchart of the present invention; Figure 2 shows the relative deviation between the core power distribution obtained by the method of the present invention and the core power distribution obtained by the Monte Carlo program reference solution. Detailed Implementation
[0018] The present invention will be further described in detail below with reference to the accompanying drawings and specific embodiments: As shown in Figure 1, the present invention provides a method for solving the few-group diffusion coefficient of a nuclear reactor based on anisotropic neutron transport, comprising the following steps: Step 1: Based on the geometric dimensions and material information of the nuclear reactor fuel assembly, a physical model of the fuel assembly is constructed, and the material region and the source region are meshed. The material information includes at least the nuclide composition and density parameters of each material region. Under a preset energy group structure, a multi-group macroscopic cross-section of the material region is generated from a multi-group cross-section library. Preferably, the component calculation program LOCUST in the advanced pressurized water reactor core physics analysis software Bamboo-C is used to complete the construction of the fuel assembly physical model, mesh generation, and multi-group macroscopic cross-section generation, and outputs the material region and source region indices, volume information, and corresponding multi-group macroscopic cross-section data for use in subsequent steps.
[0019] Step 2: Perform anisotropic P1 neutron transport calculations using the multi-group macroscopic cross-section of the material region. Preferably, the anisotropic P1 neutron transport calculations are performed by the component calculation program LOCUST, and a first-order anisotropic scattering source is introduced when calculating the total source term. The total source term is the sum of the fission source term and the anisotropic scattering source term.
[0020] In this embodiment, the component computation program LOCUST performs a feature line scan solution based on the total source term, and normalizes the angular flux during the scan to obtain the average angular flux density moment; wherein, the neutron flux density in the flat source region is the zero-order moment of the average angular flux density moment (corresponding to the average angular flux density moment with a spherical harmonic function expansion order of zero). Subsequently, it is determined whether the transport computation has converged according to a preset convergence criterion; if it has not converged, the total source term is updated using the current average angular flux density moment and the feature line scan iteration is repeated until the convergence criterion is met, and the finally converged neutron flux density in the flat source region is output.
[0021] The zeroth moment of the mean angular flux density moment corresponding to the neutron flux density in the flat source region obtained in step 2 is the mean angular flux density moment with a spherical harmonic expansion order of zero. The mean angular flux density moment is a physical quantity involved in the iteration of the anisotropic P1 neutron transport calculation, and the calculation formula is shown in the following equation:
[0022] In the formula: i represents the index of the flat source region; g represents the index of the energy group after scattering; l represents the index of the order of the spherical harmonic function expansion; m represents the index of the order of the spherical harmonic function expansion; j represents the index of the angle; k represents the index of the characteristic line. —The average angular flux density moment of the spherical harmonic function expansion order l and expansion degree m in the g group of the i-th flat source region; —The product of the angle weight of the k-th feature line with angle number j and the feature line width; —The total source term of the g-th group in the i-th flat source region, with angle number j; —The length of the k-th characteristic line with angle number j in the i-th flat source region; —The g-group incident angular flux of the i-th flat source region with angle number j and the k-th characteristic line; —The total cross section of the g-th group in the i-th flat source region; —The expanded moments of a spherical harmonic function with order l, degree m, and angle number j; —Volume of the i-th flat source region.
[0023] The formula for calculating anisotropic scattering sources is shown below:
[0024] In the formula: G represents the total number of transport energy groups; g' represents the index of the energy group number before scattering; —Anisotropic scattering source term with angle number j in the g group of the i-th flat source region; —The first-order scattering cross section of the i-th flat source region from the g'-th group to the g-th group; —The expanded moments of a spherical harmonic function with order l, degree m, and angle number j; —The average angular flux density moment of the spherical harmonic function expansion order l and expansion degree m in the i-th flat source region and g'-th group.
[0025] Step 3: Normalize the neutron flux density in the flat source region to the component scale according to the volume weight of the flat source region to obtain the component neutron energy spectrum; then construct the multi-group diffusion coefficient to solve the linear equation system, and solve the multi-group diffusion coefficient; the construction method of the multi-group diffusion coefficient solution linear equation system in Step 3 is as follows: The neutron flux equation in the P1 equation system is shown in the following equation:
[0026] In the formula: r represents the spatial position; g' represents the energy group number index before scattering; g represents the energy group number index after scattering; —The spatial gradient of the neutron flux density of the g-th group at position r; —The total cross section of the g-th group at position r; —The neutron flux of the g-th group at position r; —The neutron flux of the g'th group at position r; —The first-order scattering cross section at position r from the g' group to the g group.
[0027] Based on the neutron flow equation in the P1 equation system, the following equation is derived using the assumptions of Fick's law:
[0028] In the formula: —The spatial gradient of the neutron flux density of the g' group at position r; —The multigroup diffusion coefficient of the g-th group at position r; — The multigroup diffusion coefficient of the g'th group at position r.
[0029] Then, by integrating in the spatial domain, the multigroup diffusion coefficient is obtained, and a system of linear equations is solved. This system of equations uses the neutron energy spectrum of the component, the total cross section of the material region, and the first-order scattering cross section as parameters. The solution to the multigroup diffusion coefficient linear equations is shown in the following equation:
[0030] In the formula: — Neutron flux density of the g-th group; — Neutron flux density of the g' group.
[0031] Step 4: Use the multigroup diffusion coefficient obtained by the solution to perform transport correction on the total cross section and the zero-order self-scattering cross section of the material region. Correct the total cross section of the material region to the transport cross section of the material region and obtain the corrected zero-order self-scattering cross section of the material region. Do not process the cross sections of the remaining material regions.
[0032] The transport correction described in step 4 is specifically as follows: Based on the assumptions of Fick's law, the transport cross-section of the material region is calculated using the multigroup diffusion coefficient obtained in step 3. This transport cross-section is then used to replace the total cross-section of the material region, thus completing the transport correction for the total cross-section of the material region. The formula for calculating the transport cross-section of the material region is shown below:
[0033] —The transport section of the g-th group at location r.
[0034] When correcting the zero-order self-scattering cross section of the material region, the correction amount is set to the difference between the total cross section and the transport cross section of the material region. This correction amount is then subtracted from the zero-order self-scattering cross section of the material region to obtain the corrected zero-order self-scattering cross section. The transport correction formula is shown below:
[0035] In the formula: —The zero-order self-scattering cross section at position r.
[0036] Step 5: Calculate isotropic neutron transport with transport corrections using the zero-order self-scattering cross section and transport cross section of the material region completed in Step 4, as well as the remaining uncorrected macroscopic cross sections of the multiple groups of the material region.
[0037] Preferably, the isotropic neutron transport calculation with transport correction is performed by the component calculation program LOCUST, and anisotropic scattering sources are ignored when calculating the total source term. The total source term is the sum of the fission source term and the isotropic scattering source term.
[0038] In this embodiment, the component calculation program LOCUST performs a feature line scan solution based on the total source term, and normalizes the angular flux during the scan to obtain the neutron flux density in the flat source region. Then, it determines whether the transport calculation has converged according to a preset convergence criterion. If it has not converged, it updates the total source term using the current neutron flux density in the flat source region and repeats the feature line scan iteration until the convergence criterion is met, and outputs the finally converged neutron flux density in the flat source region.
[0039] Step 6: Perform leakage correction on the neutron flux density in the flat source region obtained in Step 5. During the leakage correction process, a one-dimensional neutron transport model is constructed, and the energy spectrum under the critical state is solved to correct the neutron flux density in the flat source region. The correction formula is as follows:
[0040] In the formula: —The corrected neutron flux density of the g-th group in the i-th flat source region; — Neutron flux density of the g-th group in the i-th flat source region before correction; —Critical energy spectrum for leakage correction calculation; —Single-region multi-group neutron flux density; using the corrected flat-source neutron flux density to homogenize the multi-group diffusion coefficient, the fuel assembly's few-group diffusion coefficient is determined. The formula for energy group merging during assembly homogenization is as follows:
[0041] In the formula: r represents the spatial variable; h represents the minority group diffusion coefficient and energy group number index; V represents the homogenization volume of the component; — Minority group diffusion coefficient of group h; —The multi-group diffusion coefficient of the g-th group; — Neutron flux density of the g-th group at position r.
[0042] In step 2, anisotropic P1 neutron transport calculations are performed based on the multi-group macroscopic cross-section of the material region to obtain the neutron flux density in the flat-source region for the first transport calculation. This flat-source neutron flux density is normalized according to the flat-source volume weight to obtain the component neutron energy spectrum, which is used to solve for the multi-group diffusion coefficient. In step 5, isotropic neutron transport calculations with transport correction are performed based on the transport-corrected zero-order self-scattering cross-section of the material region, the transport cross-section of the material region, and the remaining uncorrected multi-group macroscopic cross-sections of the material region to obtain the flat-source neutron flux density for the second transport calculation. This flat-source neutron flux density, after leakage correction, is used to homogenize the multi-group diffusion coefficient in the component, obtaining the few-group diffusion coefficient.
[0043] After implementing the above steps using the pressurized water reactor assembly calculation program LOCUST, physical models of various fuel assemblies were constructed. After performing assembly calculations in the pressurized water reactor assembly calculation program LOCUST using both conventional methods and the method of this invention, minority group parameters were generated. The changes in the minority group diffusion coefficient were analyzed, and the accuracy was verified by comparing the diffusion coefficient calculated using the cumulative migration method in the Monte Carlo program with the reference solution. Table 1 lists the comparison results of the minority group diffusion coefficients of various fuel assemblies when using different methods.
[0044] Table 1
[0045] In addition, the core diffusion program SPARK was used to perform three-dimensional core calculations with minority group parameters, and the accuracy was verified by comparing the three-dimensional core calculations of the Monte Carlo program with the reference solution. The results are shown in Figure 2.
[0046] Numerical results show that, after adopting the method of this invention, the accuracy of the minority group diffusion coefficient calculated by the pressurized water reactor assembly calculation program LOCUST is significantly improved compared with the results before adopting the method of this invention, and the calculation error of the core power distribution meets the requirements of the industrial limit (the industrial limit for core power distribution deviation is ±10%). In addition, after adopting the method of this invention, the efficiency of the pressurized water reactor assembly calculation program LOCUST in generating the core diffusion calculation assembly library meets the needs of industrial applications.
Claims
1. A method for solving the few-group diffusion coefficient of nuclear reactors based on anisotropic neutron transport, characterized in that: The process includes the following steps: Step 1: Based on the geometric dimensions and material information of the nuclear reactor fuel assembly, construct a physical model of the fuel assembly and perform mesh generation for the material region and the source region. Generate a multi-group macroscopic cross section for the material region from a multi-group cross section library. Step 2: Calculate anisotropic P1 neutron transport using the multi-group macroscopic cross section of the material region to obtain the neutron flux density in the source region. Step 3: Normalize the neutron flux density in the source region to the assembly scale according to the volume weight of the source region to obtain the neutron spectrum of the assembly. Then, construct a multi-group diffusion coefficient system to solve the linear equations and solve for the multi-group diffusion coefficient. Step 4: Use the solved multi-group diffusion coefficient to calculate the total cross section of the material region and the zero-order self-electron diffusion coefficient of the material region. The scattering cross section is corrected for transport, and the total cross section of the material region is corrected to the transport cross section of the material region to obtain the corrected zero-order self-scattering cross section of the material region; Step 5: Using the zero-order self-scattering cross section of the material region obtained from the transport correction in Step 4, the transport cross section of the material region, and the remaining uncorrected multi-group macroscopic cross sections of the material region, isotropic neutron transport calculations are performed to obtain the neutron flux density in the flat source region; Step 6: The neutron flux density in the flat source region obtained in Step 5 is corrected for leakage to obtain the corrected neutron flux density in the flat source region. The multi-group diffusion coefficient is homogenized using the corrected flat source region neutron flux density to determine the small-group diffusion coefficient of the fuel assembly.
2. The method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport according to claim 1, characterized in that, The zeroth moment of the mean angular flux density moment corresponding to the neutron flux density in the flat source region obtained in step 2 is the mean angular flux density moment with a spherical harmonic expansion order of zero. The mean angular flux density moment is a physical quantity involved in the iterative calculation of anisotropic P1 neutron transport, and the calculation formula is shown in the following equation: In the formula: i represents the index of the flat source region; g represents the index of the energy group after scattering; l represents the index of the order of the spherical harmonic function expansion; m represents the index of the order of the spherical harmonic function expansion; j represents the index of the angle; k represents the index of the characteristic line. —The average angular flux density moment of the spherical harmonic function expansion order l and expansion degree m in the g group of the i-th flat source region; —The product of the angle weight of the k-th feature line with angle number j and the feature line width; —The total source term of the g-th group in the i-th flat source region, with angle number j; —The length of the k-th characteristic line with angle number j in the i-th flat source region; —The g-group incident angular flux of the i-th flat source region with angle number j and the k-th characteristic line; —The total cross section of the g-th group in the i-th flat source region; —The expanded moments of a spherical harmonic function with order l, degree m, and angle number j; —Volume of the i-th flat source region.
3. The method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport according to claim 1, characterized in that, The method for constructing the linear equation system for solving the multi-group diffusion coefficient in step 3 is as follows: The neutron flow equation in the P1 equation system is shown in the following equation: In the formula: r represents the spatial location; g' represents the index of the energy group before scattering; —The spatial gradient of the neutron flux density of the g-th group at position r; —The total cross section of the g-th group at position r; —The neutron flux of the g-th group at position r; —The neutron flux of the g'th group at position r; —The first-order scattering cross section at position r from the g' group to the g group; based on the neutron flux equation in the P1 equation set, the following equation is derived using the assumptions of Fick's law: In the formula: —The spatial gradient of the neutron flux density of the g' group at position r; —The multigroup diffusion coefficient of the g-th group at position r; —The multigroup diffusion coefficient of the g'th group at position r; Then, by integrating in the spatial domain, the multigroup diffusion coefficient is obtained, and a system of linear equations is solved. This system of equations uses the neutron energy spectrum of the component, the total cross section of the material region, and the first-order scattering cross section as parameters. The solution to the multigroup diffusion coefficient linear equations is shown in the following equation: In the formula: — Neutron flux density of the g-th group; — Neutron flux density of the g' group.
4. The method for solving the minority group diffusion coefficient of nuclear reactors based on anisotropic neutron transport according to claim 1, characterized in that, The transport correction described in step 4 is specifically as follows: Based on the assumptions of Fick's law, the transport cross-section of the material region is calculated using the multigroup diffusion coefficient obtained in step 3. This transport cross-section is then used to replace the total cross-section of the material region, thus completing the transport correction for the total cross-section of the material region. The formula for calculating the transport cross-section of the material region is shown below: —The transport cross section of the g-th group at position r; When correcting the zero-order self-scattering cross section of the material region, let the correction amount be the difference between the total cross section of the material region and the transport cross section of the material region. Subtract this correction amount from the zero-order self-scattering cross section of the material region to obtain the corrected zero-order self-scattering cross section of the material region; The transport correction formula is shown in the following equation: In the formula: —The zero-order self-scattering cross section of the material region at position r.