Reactor core physical calculation method based on component response relationship conservation

Abstract

The invention provides a reactor core physical calculation method based on component response relationship conservation. The method comprises the following steps: dividing a reactor core into a plurality of segments; for each segment, determining a homogenization group constant; for each segment, determining a boundary flux-net flow response relationship, and defining the boundary flux-net flow response relationship as a mapping relationship between neutron net flow information and neutron flux information at the boundary of the segment; and based on the boundary flux-net flow response relationship of each segment, performing calculation related to reactor core physics.

Description

Technical Field

[0001] The present invention relates to a core physics calculation method. Background Art

[0002] Core physics calculation is the basis of reactor physical design. The current mainstream calculation process is a two-step method. That is, according to the characteristics that the components in the core have a repetitive structure (or similar structure), first, a two-dimensional fine transport calculation of the components is carried out under reflective boundary conditions to obtain the equivalent few-group homogenized cross section of the components, and then a three-dimensional diffusion calculation of the whole core is carried out using the few-group equivalent cross section.

[0003] In addition to the mainstream method, some people also propose the response matrix method to perform component equivalence. The response matrix is a tool used to determine the influence of one or more variables on another or more variables in the fields of statistics and data analysis. From a mathematical perspective, the response matrix is an extension of the function mapping between single variables to the matrix mapping between vectors. For example, a known boundary flow response matrix method (see "Research on Component Equivalent Calculation Method Using Boundary Flow Response Matrix", DOI: 10.7538 / yzk.2013.47.zk.0005) does not require homogenization of the components. Instead, by directly performing transport calculations on the components under various boundary incident flow conditions, the coupling relationship between the incident flow and the outgoing flow on the surface of the components is obtained, and then the boundary flow coupling relationship is used to solve the whole core.

[0004] However, in actual application, the existing response matrix method needs to calculate a large number of response parameters under different incident flow conditions, including different neutron energies, angles, distributions, etc. Moreover, when the state parameters (such as boron concentration, fuel temperature, moderator temperature, burnup depth, etc.) in the components change, it is also necessary to recalculate the response parameters after the change. This makes the parameter calculation in the matrix complex and time-consuming, and it is difficult to become the mainstream core physics calculation method. Summary of the Invention

[0005] The present invention aims to provide a new core physics calculation method.

[0006] To this end, the present invention provides a core physics calculation method, which includes the following steps: dividing the core into several nodal blocks; for each nodal block, determining the homogenized group constants; for each nodal block, determining the boundary flux-net current response relationship, where the boundary flux-net current response relationship is defined as the mapping relationship between the neutron net current information and the neutron flux information at the nodal block boundary; based on the boundary flux-net current response relationship of each nodal block, performing calculations related to core physics.

[0007] In the sense of the terms of the present invention, the "core" is the object to be calculated or analyzed, which may be an actual core or a virtual core design. As is known in the art, a "node" is a sub-region obtained by coarsely meshing the whole or part of the core to be calculated for calculation purposes, especially for numerically solving the neutron diffusion equation. A "node" may correspond to an actual core fuel assembly or may be another meshing made for calculation purposes. A "node" can be divided into a reasonable pattern known or unknown in the art. As is known in the art, a "homogenized group constant" is an equivalent group constant calculated for a selected region by treating the region as a homogeneous medium region, which can be achieved by neutron transport calculation methods and programs known in the art. In particular, homogenization equivalence and calculation of group constants are performed as in the first step of the traditional "two-step method".

[0008] The present invention pioneerly proposes that, on the basis of obtaining the homogenized group constants, the coupling relationship between the neutron flux and the neutron net current at the node boundary (surface), that is, the "boundary flux-net current response relationship" as called in the present invention, is further obtained, and calculations related to core physics are carried out with this as a tool.

[0009] Thus, the present invention provides a new core physics calculation method.

[0010] Optionally, the above steps for determining the boundary flux-net current response relationship include determining the boundary flux-net current response matrix through calculation, wherein the homogenized group constants of the node are used in this calculation, and the boundary flux-net current response matrix is defined such that wherein, is the basis expansion of the neutron flux distribution function at the node boundary represented in vector form, s is the basis expansion of the neutron net current distribution function at the node boundary represented in vector form, and R is the boundary flux-net current response matrix of the node. Further optionally, the neutron flux distribution function and the neutron net current distribution function are expanded in terms of polynomial basis functions.

[0011] In this way, the steps for determining the boundary flux-net current response relationship can be conveniently implemented in a numerical calculation program.

[0012] In particular, the above steps for determining the boundary flux-net current response relationship mathematically involve analytically solving the neutron flux distribution function according to the steady-state neutron diffusion equation under Neumann boundary conditions.

[0013] In mathematical physics equations, boundary conditions are divided into three categories. The first type of boundary condition is also known as the Dirichlet boundary condition, which requires giving the function value of the function to be solved at the boundary; the second type of boundary condition is also known as the Neumann boundary condition, which requires giving the derivative value of the function to be solved at the boundary; the third type of boundary condition is also known as the mixed boundary condition, which is a linear combination of the function to be solved and its derivative.

[0014] In the field of reactor physics, the net current at the boundary is related to the second type (Neumann) boundary condition, and the incident current and outgoing current information belong to the third type of boundary condition. Therefore, analytically solving the neutron flux distribution function under the Neumann boundary condition will establish a mathematical relationship between the neutron flux and the neutron net current, which can be used as the basis for specifically determining the boundary flux-net current response relationship of the node.

[0015] Optionally, the above steps for determining the boundary flux-net current response relationship include determining the diffusion response relationship, which is based on neutron diffusion theory. Among them, each node is equivalent to a node with uniform group constants, and further includes correcting the diffusion response relationship to obtain a boundary flux-net current response relationship closer to neutron transport theory. Further optionally, the present invention also includes determining the response relationship correction factor for each node, and the response relationship correction factor is used to correct the diffusion response relationship. Among them, the response relationship correction factor for each node and the homogenized group constants are calculated together by a calculation program based on neutron transport theory.

[0016] In particular, the above diffusion response relationship can be obtained by analytically solving the neutron flux distribution function according to the neutron diffusion equation under the Neumann boundary condition, and can be further transformed into a vector-matrix form.

[0017] The diffusion response relationship regards the node as a uniform node, so the accuracy is low when the actual inhomogeneity in the node is strong. The present invention proposes to correct the diffusion response relationship to obtain a response relationship closer to neutron transport theory, which can be realized in a certain way by means of a calculation program based on neutron transport theory.

[0018] Optionally, each node is divided into rectangles. Therefore, each node has four boundaries. The computational complexity in this case is small.

[0019] Optionally, the above calculations related to core physics include calculating the effective neutron multiplication factor of the core and the information related to neutron flow at the boundaries of each node, and then calculating the core power distribution. Brief Description of the Drawings

[0020] Figure 1 is at least a partial flow schematic diagram of a core physics calculation method according to an embodiment of the present invention.

[0021] Figure 2 is a schematic diagram showing at least part of the process of another core physics calculation method according to an embodiment of the present invention.

[0022] Figure 3a is a schematic diagram of the core layout of an IAEA two-dimensional benchmark problem, Figure 3b showing the few-group constants of the components specified in the IAEA two-dimensional benchmark problem, Figure 3c showing the verification results of applying the present invention to the IAEA two-dimensional benchmark problem, where the calculated values, reference values, and relative errors are shown.

[0023] Figure 4 is a schematic diagram of the core layout of a simplified pressurized water reactor (PWR) example, used to verify the core physics calculation method according to an embodiment of the present invention.

[0024] Figure 5 is a schematic diagram of the core layout of a simplified boiling water reactor (BWR) example, used to verify the core physics calculation method according to an embodiment of the present invention. Detailed implementation manners

[0025] The following introduces the implementation manners and advantages of the present invention through some embodiments.

[0026] Figure 1 is a schematic diagram showing at least part of the process of a core physics calculation method according to an embodiment of the present invention. Basically, as Figure 1 shown, the core physics calculation method according to the present invention includes the following steps: Step S1, dividing the core into a plurality of nodal blocks; Step S2, determining the homogenized group constants (for each nodal block); Step S3, determining the boundary flux-net current response relationship (for each nodal block); Step S4, performing calculations related to core physics.

[0027] Steps S1 and S2 can refer to the first step in the traditional "two-step method", that is, performing fine transport calculations for components / nodal blocks (specifically, it may involve, for example, the method of characteristics, SN method, etc.) to obtain the equivalent homogenized multi-group or few-group constants (especially cross-section constants) of components / nodal blocks. However, the subsequent steps to be executed are different from the traditional method.

[0028] Based on obtaining the homogenized group constants of each nodal block, the present invention proposes that the next step is to examine the coupling relationship between the neutron net current and neutron flux of each nodal block. Specifically, perform Step S3. Based on the nodal division and obtaining the homogenized group constants, determine the boundary flux-net current response relationship of the nodal block. The so-called boundary flux-net current response relationship of the nodal block is defined herein as the mapping relationship between the neutron net current information and neutron flux information at the nodal block boundary.

[0029] As an example, the present invention proposes to start from the Neumann boundary condition (the second type of boundary condition). In reactor physics, this boundary condition involves the net neutron current. As an example, by solving the neutron diffusion equation under the Neumann boundary condition, the coupling relationship between the neutron flux and the net neutron current is investigated.

[0030] Taking the two-group case (i.e., the so-called thermal group and fast group) as an example, the relationship between the neutron flux information and the net neutron current information at the nodal boundary is analyzed below. The node is regarded as a homogenized node (i.e., within the same node, the group constants are independent of position, that is, the homogenized group constants generally applicable to this node).

[0031] First, consider the fixed-source two-group steady-state neutron diffusion equation as follows:

[0032]

[0033] In the formula, the subscripts 1 and 2 represent the fast group and the thermal group respectively, D is the diffusion coefficient, Σ r represents the fast-group removal cross section, Σ a2 is the thermal-group absorption cross section, υ is the average number of neutrons emitted per fission, Σ f1 is the fast-group fission cross section, Σ f2 is the thermal-group fission cross section. The above coefficients, that is, the homogenized group constants of the node, are obtained through step S2. and are the neutron flux distribution functions of the fast group and the thermal group respectively.

[0034] The boundary conditions are set as: both the fast group and the thermal group are unit point sources, that is, the source strength is 1.

[0035] Analytically solve equation (1), that is, solve the neutron flux expression, and the solution is in the form of:

[0036]

[0037] In the formula, φ1(r) and φ2(r) are the fundamental solutions of the Helmholtz equation for an infinite homogeneous medium, and a, b, c, and d are calculated from the boundary conditions and the cross-section constants.

[0038] For example, define the following coefficient matrix:

[0039]

[0040] The relevant parameters are the same as those in equation (1).

[0041] Denote the eigenvalue vector of matrix F as Λ = [Λ1, Λ2], and the eigenvector as

[0042] Then the coefficients A1 and A2 related to the boundary source strength can be calculated, and the coefficients A1 and A2 should satisfy equation (4).

[0043]

[0044] Then, calculate a, b, c, and d in Equation (1) through Equation (5).

[0045]

[0046] Then, considering that the source is a certain distribution, still taking the two-group case as an example, integrate the source, and the solution is in the form of:

[0047]

[0048] In the formula, f1(u) and f2(u) respectively represent the distribution functions of the fast-group and thermal-group neutron sources on the boundary, that is, the boundary neutron net current distribution function, and a, b, c, and d are the same as the coefficients in Equation (1).

[0049] Then, for a single node, under the second-type boundary condition, use the mirror image method to expand the geometry to infinity, and then use the result of Equation (6) and the superposition principle to obtain the analytical solution in the form of:

[0050]

[0051] In the formula, f i1 , f i2 respectively represent the i-th fast-group and thermal-group distribution sources of the infinite homogeneous medium, and the coefficients a, b, c, and d are the same as above.

[0052] Therefore, taking the two-group case as an example, for the homogenized node, a mapping relationship between the boundary neutron flux and the neutron net current can be deduced. Based on this, the boundary flux-net current response relationship of the present invention can be defined or further calculated.

[0053] Furthermore, for N (N>2) groups, the dimensions of the matrices and vectors in the above derivation process will all be N-dimensional, which can be proved by induction, that is, the N-group neutron flux is a linear combination of the fundamental solutions of the Helmholtz equation corresponding to different eigenvalues of the N-group neutron diffusion equation coefficient matrix of the infinite homogeneous medium. Here, an example of the g-th group (2 < g ≤ N) is given:

[0054]

[0055] In the formula, f ik represents the i-th k-th group distribution source of the infinite homogeneous medium.

[0056] Therefore, for any multi-group, theoretically, the coupling or mapping relationship between the neutron flux and the neutron net current can be deduced by solving the neutron diffusion equation under the Neumann boundary condition.

[0057] Thus, in some embodiments, step S3 mathematically involves analytically solving the neutron flux distribution function according to the steady-state neutron diffusion equation under Neumann boundary conditions.

[0058] As an example, especially for numerical calculations, it is advantageous to represent or record the linear coupling relationship in vector-matrix form.

[0059] For this purpose, the neutron flux distribution function and the neutron net current distribution function are expanded using basis functions (such as polynomial-type basis functions) at each boundary of the node. Based on the analytical solution of the above-mentioned neutron diffusion equation, such as Equation (8), the vector-matrix expression can be obtained:

[0060]

[0061] Where, is the basis expansion of the neutron flux distribution function at the node boundary represented in vector form, s is the basis expansion of the neutron net current distribution function at the node boundary represented in vector form, and R is the boundary flux-net current response matrix of the node.

[0062] Taking the two-dimensional scenario as an example, exemplarily, in step S1, each node is divided into rectangles, so that each node has four boundaries in step S3.

[0063] Exemplarily, the flux and net current are expanded using polynomials at each boundary, the distribution of the flux and net current is replaced by the expansion coefficients of the polynomials, and then according to Equation (8), for each boundary d (two-dimensional, d = 1, 2, 3, 4, representing the left, upper, right, and lower boundaries of the node respectively), each energy group g (g = 1, 2,..., G, G is the total number of energy groups), and the net current unit order n (n = 0, 1, 2..., N, N is the highest expansion order), the expansion coefficients of the flux at each boundary of the node are combined into a column vector Thus, the response matrix R can be obtained. Taking the two-dimensional, single-group, zero-order expansion as an example, there is

[0064]

[0065] In the formula, taking as an example, when there is a single-group zero-order unit net current at the left boundary of the node, the column vector composed of the expansion coefficients of the flux at the four boundaries of the node at this time is The other column vectors of matrix (10) have similar meanings.

[0066] Therefore, by replacing the distribution functions of the flux and net current with the expansion coefficients of the polynomials, calculating the response matrix according to the analytical solution of the neutron diffusion equation, and then arranging it in matrix-vector form, the boundary flux-net current response matrix expression of each node can be obtained.

[0067] In the above example, the response matrix R is defined as the response matrix of the flux to the net flow. However, obviously, in some other embodiments, the response matrix R can also be defined as the response matrix of the net flow to the flux.

[0068] Therefore, by performing step S3, the boundary flux-net flow response relationship of each section block is obtained.

[0069] Then, step S4 is performed to conduct calculations related to core physics. For example, calculate the effective neutron multiplication factor of the core and information related to neutron flow at the boundaries of each section block, and then calculate the core power distribution.

[0070] Specifically, step S4, for example, includes solving the steady-state neutron diffusion equation for the entire reactor, which can be divided into a preprocessing stage and an official solution stage. The preprocessing stage is the process of constructing a parameterized interpolation table, that is, the process of making an interpolation table of the expansion coefficients of the basis functions for different parameters under a specific section block geometry. The preprocessing stage can be completed offline. Once completed, this table can be used as a general interpolation table for solving the neutron diffusion equation of the entire reactor with this section block shape worldwide. In the official solution stage, according to the relationship between neutron flux, interface source and incident flow, outgoing flow, the response matrix of the net flow to the flux can be transformed into the response matrix of the incident flow to the outgoing flow, that is, J + =RJ - , and then existing solution methods for all traditional boundary flow response matrices can be used. For example, the effective neutron multiplication factor of the entire reactor and the expansion coefficients of the boundary flow can be obtained. Further, by separately solving the fixed source problem of the response for each section block / component, the detailed flux distribution of each section block / component can be obtained. Or, if only the average power distribution of the section blocks / components of the entire reactor needs to be obtained, the neutron flux equation can be integrated over volume, and by solving a gⅹg linear equation system, the total flux of each group in the g groups can be obtained, and then its power distribution can be obtained (usually referred to as component power reconstruction).

[0071] Figure 2 is at least a partial flow schematic diagram of another core physics calculation method according to an embodiment of the present invention. Different from Figure 1 , this method further includes determining a response relationship correction factor, that is, step S2.2, and step S3 now includes determining a diffusion response relationship, that is, step S3.1, and then correcting the diffusion response relationship, that is, step S3.2.

[0072] In the previous example, the boundary flux-net flow response relationship or response matrix of the section block directly originates from the analytical solution of solving the multi-group steady-state neutron diffusion equation under Neumann boundary conditions. Now, the response relationship obtained in this way is defined as the diffusion response relationship, and the corresponding response matrix is defined as the diffusion response matrix R d . Obviously, the diffusion response relationship or diffusion response matrix regards the section block as a section block with uniform group constants and is based on neutron diffusion theory.

[0073] However, as the design of the reactor core becomes increasingly complex, the non-uniformity within the core components is also enhanced. Although the method of component / assembly homogenization equivalence has a fast calculation speed, its calculation accuracy deteriorates in a core with strong non-uniform effects. This is mainly because when the same components in the core are in different positions, their boundary neutron fluxes are different, and component homogenization may not be able to accurately consider the impact of changes in component boundary neutron fluxes on the equivalent homogenized cross-section.

[0074] This defect can be solved by correcting the diffusion response relationship or the diffusion response matrix. The specific operation can still draw on the discontinuous factor in the traditional "two-step method" or the SPH factor in the super homogenization method. In particular, the homogenization constant of an assembly can generally be easily obtained through a calculation program based on neutron transport theory. On this basis, for the present invention, the response relationship correction factor can be calculated together.

[0075] For example, first use a neutron transport program to calculate each assembly to obtain neutron flux and boundary current information. Then, on the one hand, perform assembly homogenization according to the flux information, that is, obtain the homogenized group constants. On the other hand, calculate the response relationship correction factor for the flux-net current response relationship according to the boundary current information. Here, a calculation method for the response correction factor is given. For a target assembly, construct a boundary condition where there is a net current on only one boundary surface and the other boundaries are fully reflective. Perform a transport calculation before homogenization to obtain the polynomial expansion coefficients of the transport boundary flux of the assembly This process can be completed by existing transport programs. After homogenization, perform a diffusion calculation on this assembly under the same boundary conditions as the transport to obtain the polynomial expansion coefficients of the diffusion boundary flux of the assembly Due to the differences in the calculation models, the two results will inevitably be different, and the result of the transport is more accurate Then use the method of undetermined coefficients to calculate and The correction factor M d , M d satisfies This is the correction factor for the current non-fully reflective boundary. The calculation methods for the correction factors of the other 3 boundaries are similar. If the assembly has rotational invariance, only the correction factor for one boundary needs to be calculated.

[0076] According to some embodiments of the present invention, the diffusion response relationship from the homogenized assembly net current to the flux can be analytically derived. However, a more accurate description of an assembly or a core should be the transport model, that is, the transport response relationship. Therefore, the role of the response relationship correction factor is, for example, to correct the diffusion response relationship to obtain a boundary flux-net current response relationship closer to the neutron transport theory.

[0077] Described in mathematical language, according to the relationship between neutron flux and net flow response under the diffusion model Consider the transport response relationship where is the boundary outgoing flow of the heterogeneous node based on the transport theory model; in the case of strong heterogeneity, Therefore, consider correcting R d For example, obtain the correction matrix M to get the corrected response matrix R m = MR d , so that And M that satisfies such conditions can be called the response relationship correction factor. The present invention believes that further design based on the existing neutron transport calculation program can achieve the above purpose.

[0078] Figure 3a is an IAEA two-dimensional two-group benchmark problem. The node shape is square, the node width is 20 cm, and there are 177 fuel assemblies. This benchmark problem has been homogenized and does not require a response relationship correction factor, or equivalently, the response relationship correction matrix is the identity matrix. The few-group constants are shown in Figure 3b . This general benchmark problem is widely used to verify the effectiveness of new methods and programs.

[0079] Here, the neutron flow and neutron flux are expanded by a second-order polynomial. First, calculate the response matrix of each node corresponding to the k value according to the existing group constants, then solve the discontinuity degree α of the whole reactor, re-interpolate the k corresponding to α = 1 according to α, and repeat this process until α - 1 satisfies the convergence residual. At this time, the k is the k-effective (effective neutron multiplication factor) of the whole reactor, and the boundary flow at each interface of the whole reactor can be obtained. The k-effective calculated by the new method is 1.029573, which is 1.2 pcm different from the reference solution of 1.029585. The calculation time using MATLAB is about 0.4 s. Then, the total flux of each group is obtained by solving the 2x2 linear equations, and then the average power distribution of the whole reactor components is obtained. The results are as shown in Figure 3c . The results show that the maximum relative error of the average power distribution of the nodes is 0.412%. According to the calculation results of the IAEA two-dimensional two-group benchmark problem, the feasibility of the core physics calculation method of the present invention is verified.

[0080] Figure 4It is a simplified PWR example (quarter geometry) with two enrichment levels, 1.8% and 3% respectively. The assembly shape is square, the assembly width is 16.256 cm, there are 16 fuel assemblies, and each fuel assembly consists of 10×10 fuel rods. An existing homogenization program is used to homogenize the assemblies, and the response relationship correction factor is calculated. This example is used to test the effectiveness of the core physics calculation method according to the embodiments of the present invention in solving typical PWR problems.

[0081] The specific process of applying the core physics calculation method according to the embodiments of the present invention to solve this example can be the same as or similar to the aforementioned IAEA benchmark problem, and the calculation results are directly given here. The calculation results of the effective multiplication factor (k-effective) of the whole core are shown in Table 1, where the reference solution is obtained by performing a full-core calculation using the Monte Carlo method. As can be seen from Table 1, the difference between the k-effective obtained by the core physics calculation method of the present invention and the reference solution is 93 pcm, and the calculation deviation of the two-step method is 316 pcm. Compared with the traditional two-step method, the calculation error is reduced by about 70%, and the calculation accuracy is improved by about 3 times. The calculation results of the core power distribution are shown in Table 2. It can be seen that the accuracy of the power distribution obtained by the core physics calculation method of the present invention is slightly better than that of the two-step method and is closer to the Monte Carlo reference solution. Therefore, the superiority and effectiveness of the core physics calculation method of the present invention are verified.

[0082] Table 1

[0083]

[0084] Table 2

[0085]

[0086] Figure 5 It is a simplified BWR (without control rods) example with two enrichment levels, 1.8% and 3% respectively. The assembly shape is square, the assembly width is 16.256 cm, there are 16 fuel assemblies, and each fuel assembly consists of 9×9 fuel rods. An existing homogenization program is used to homogenize the assemblies, and the response relationship correction factor is calculated. This example is used to test the effectiveness and superiority of the core physics calculation method according to the embodiments of the present invention on a core with a wide water gap, i.e., a core with strong non-uniformity.

[0087] The solution process is the same as or similar to the aforementioned example, and the calculation results are directly given here. The k-effective results are shown in Table 3, and the power distribution is shown in Table 4. The reference solution is obtained by performing a full-core calculation using the Monte Carlo method. It can be seen that both the k-effective and the power distribution obtained by the method of the present invention are better than those of the two-step method. Therefore, the superiority and effectiveness of the core physics calculation method of the present invention are further verified.

[0088] Table 3

[0089]

[0090] Table 4

[0091]

[0092] Although some embodiments of the present invention have been shown and described, those of ordinary skill in the art will understand that various combinations, changes, modifications, substitutions, and variations can be made to these embodiments without departing from the principles and spirit of the present invention. The scope of the present invention is defined by the claims and their equivalents.

Claims

1. A core physics calculation method based on component response relationship conservation, comprising the following steps: Divide the core into several sections; For each node, determine the homogenization group constant; For each node, a boundary flux-net flow response relationship is determined, wherein the boundary flux-net flow response relationship is defined as a mapping relationship between neutron net flow information and neutron flux information at a node boundary; Based on the boundary flux-net flow response relationship of each node, calculations related to core physics are performed.

2. The core physics calculation method according to claim 1, characterized in that: The step of determining the boundary flux-net flow response relationship comprises determining the boundary flux-net flow response matrix by calculation, wherein the homogenization group constant of the node is used in the calculation, Wherein, the boundary flux-net flow response matrix is ​​defined as in, is the basis expansion of the neutron flux distribution function at the node boundary expressed in vector form, s is the basis expansion of the neutron net flow distribution function at the node boundary expressed in vector form, and R is the boundary flux-net flow response matrix of the node.

3. The core physics calculation method according to claim 2, characterized in that: The neutron flux distribution function and the neutron net flow distribution function are expanded using polynomial basis functions.

4. The core physics calculation method according to claim 1, characterized in that: The step of determining the boundary flux-net flow response relationship mathematically involves analytically solving the neutron flux distribution function according to the steady-state neutron diffusion equation under Neumann boundary conditions.

5. The core physics calculation method according to claim 1, characterized in that: The step of determining the boundary flux-net flow response relationship includes determining a diffusion response relationship, wherein the diffusion response relationship is based on neutron diffusion theory, wherein each node is equivalent to a node with uniform group constant, and further includes correcting the diffusion response relationship to obtain a boundary flux-net flow response relationship that is closer to the neutron transport theory.

6. The core physics calculation method according to claim 5, characterized in that: The core physics calculation method includes determining a response relationship correction factor for each node, wherein the response relationship correction factor is used to correct the diffusion response relationship, wherein the response relationship correction factor for each node and a homogenization group constant are calculated together by a calculation program based on neutron transport theory.

7. The core physics calculation method according to claim 1, characterized in that: Each section is divided into rectangles.

8. The core physics calculation method according to claim 1, characterized in that: The calculations related to core physics include calculating the effective neutron multiplication coefficient of the core and information related to neutron flow at the boundaries of each node, and then calculating the core power distribution.

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