A method for making multi-group cross sections of micro-pile with high precision and complex energy spectrum

Abstract

The present invention relates to a method for making a high-precision complex energy spectrum micro-reactor multi-group cross section, comprising the following steps: step 1: modeling the components or cores to be calculated, and performing continuous energy Monte Carlo transport calculation based on ACE format cross section data; step 2: tracking neutrons in different incident angle intervals within a specified homogenization area; step 3: according to step 2, counting the number of scattering events in different scattering exit angle cosine intervals, calculating the probability of the scattering exit angle cosine in each interval, and determining the scattering angle distribution; step 4: according to step 2 and step 3, calculating and obtaining multi-group data; step 5: according to step 2 and step 4, making a multi-group database. The present invention solves the calculation problem of all cross sections by tracking and counting the neutron incident angle, reduces the influence of the approximation and error used in the existing method on the micro-reactor multi-group cross section data, and improves the accuracy of the micro-reactor multi-group cross section data.

Description

Technical Field

[0001] The invention belongs to the field of particle transport calculation, and in particular relates to a method for making a complex energy spectrum micro-pile multi-group cross section based on a Monte Carlo method. Background Art

[0002] In order to obtain reactor parameters quickly and efficiently, the "two-step method" calculation has become the main method in reactor engineering. The first step of the "two-step method" is to perform neutron transport calculations on various structural materials in the core to generate homogenized multi-group cross sections; the second step is to perform core multi-group calculations based on the multi-group cross sections generated in the first step to obtain core physical quantities.

[0003] The main methods for reactor physics calculations are the Monte Carlo method and the deterministic method. Microreactors have complex geometry and strong inhomogeneity. Deterministic methods use a lot of approximations and corrections when processing them, and their applicability and reliability for microreactor calculations remain to be verified. The Monte Carlo method's powerful processing capabilities for complex geometry are suitable for the fine characterization of microreactor models, and have become an effective means of generating multi-group cross sections for complex energy spectra of microreactors.

[0004] At present, the flux separability approximation is usually used in the process of generating multi-group cross sections by Monte Carlo method. The flux separability approximation means that the energy correlation and angle correlation of neutron flux are independent of each other. Based on the above approximation, the multi-group cross sections are calculated using the scalar flux in the specified homogenization region instead of the weighted average of the angular flux. However, the above approximation will have a significant impact on the accuracy of the group constant when the homogenization region is highly non-uniform or the homogenization space material is discontinuous.

[0005] At present, the multi-group cross section generation method mainly uses the Pn method to deal with scattering anisotropy. The Pn method is to expand the scattering cross section with respect to the cosine of the neutron emission angle, and then fit and calculate the scattering matrix of each order to form the multi-group scattering cross section data. However, the above method will produce polynomial truncation errors, which may cause the fitted scattering cross section to have negative values ​​with respect to the scattering angle cosine distribution function, which is physically impossible. For points where the distribution function value of the scattering cross section with respect to the angle cosine is negative, the current multi-group cross section generation method specifies the function value of the above point as 0, which will affect the accuracy of the multi-group constant.

[0006] In the preparation of micro-reactor multi-group cross-section, due to the asymmetry of the micro-reactor structure and the non-uniformity or discontinuity caused by the parallel area, the truncation error of the flux separable approximation and the Pn method to characterize the anisotropic scattering cannot be ignored. Therefore, it is necessary to develop a method to obtain high-precision micro-reactor multi-group cross-section data. Summary of the invention

[0007] In order to solve the technical problems existing in the above-mentioned prior art, the present invention provides a method for making a high-precision micro-pile multi-group database based on the Monte Carlo method. The method for making the database is to count the multi-group parameters of different neutron incident angle intervals in the specified multi-group space region in the continuous energy Monte Carlo calculation result processing, and characterize the anisotropic scattering by counting the scattering events in different neutron scattering angle intervals, calculate the multi-group data, and make the multi-group database. Counting the different neutron incident angle intervals reduces the influence of the flux separable approximation on the group constant, and the statistical method of characterizing anisotropic scattering eliminates the influence of the negative distribution function value caused by the truncation error of the traditional Pn method on the multi-group constant. The intervals divided by different neutron incident angles in the multi-group space region are applicable to all cross sections.

[0008] The present invention adopts the following technical scheme:

[0009] Specifically, a method for making a high-precision complex energy spectrum micro-pile multi-group cross section is characterized by comprising the following steps:

[0010] Step 1: Model the components or cores to be calculated and perform continuous energy Monte Carlo transport calculations based on ACE format cross-sectional data;

[0011] Step 2: Tracking neutrons at different incident angles within a specified homogenization region;

[0012] Step 3: According to step 2, the number of scattering events in different scattering angle cosine intervals is counted, the probability of the scattering angle cosine in each interval is calculated, and the scattering angle distribution is determined;

[0013] Step 4: According to step 2 and step 3, multiple groups of data are calculated;

[0014] Step 5: Create a multi-group database based on steps 2 and 4.

[0015] Furthermore,

[0016] The step 2 includes: tracking the neutrons in different incident angle intervals in the designated homogenization area according to the calculation in step 1, and converting z, n, Ω n , g, r, E, ΔE, V are used as the screening data of the neutron flux in the specified incident angle range. After Monte Carlo transport calculation, the W, T, i, li parameters are output. The neutron flux in the specified incident angle range is calculated by formula (1); z, x, Ω n , g, ΔE, V, r, E are used as screening data for various reaction rates of neutrons in a specified homogenized region. After Monte Carlo transport calculation, W, i, wi, Ag parameters are output, and the various reaction rates of the neutrons in the specified homogenized region are calculated using formula (2).

[0017]

[0018] in:

[0019] ---In the range of energy group g, the incident angle is Ω n The flux of neutrons in the homogenized space z

[0020] ---Incident angle in Ω n The x-type reaction rate produced by neutrons in the interval in the energy group g and the homogenized space z

[0021] z---homogeneous spatial variation

[0022] n---incident angle discrete interval number

[0023] Ω n ---Discrete interval of the nth neutron incident angle

[0024] g---energy group number

[0025] r---spatial position

[0026] E---Neutron Energy

[0027] ΔE---the upper and lower bounds of the energy of energy group g

[0028] V---the volume of the homogenized space z

[0029] W---initial neutron total weight

[0030] T---the set of trajectories of all neutrons in the homogenized space volume

[0031] i---Event number

[0032] li---the length of the neutron's i-th trajectory

[0033] x---reaction type

[0034] wi---weight of neutron before reaction event i

[0035] Ag---the total set of all reaction events with neutron energy in energy group g

[0036] The step 3 comprises: in the multi-group parameter statistics of step 2, discretizing the scattered emission angle cosine, counting the number of scattering events in different scattering emission angle cosine intervals, calculating the probability of the scattering emission angle cosine in each interval by formula (3), determining the scattering angle distribution, and determining the scattering cross section by formula (4) after Monte Carlo transport calculation.

[0037]

[0038] in:

[0039] ---Incident angle in Ω n The probability that the neutron scattering reaction has an exit angle cosine of μ in the interval satisfies the relation

[0040] ---Incident angle in Ω n The macroscopic cross section of the scattering reaction with the cosine of the scattering reaction angle μ is

[0041] μ---scattering reaction cosine of the emission angle

[0042] Δμn---the interval length of the nth scattering angle cosine discrete interval

[0043] Wμ---neutron weight with cosine of scattered emission angle μ

[0044] wμ---the number of events with a scattering angle cosine of μ

[0045] ---Incident angle in Ω n The macroscopic cross section of the scattering reaction of a neutron with energy E at r is

[0046] The step 4 comprises: obtaining parameters by Monte Carlo transport calculation based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3 f g'→g (μ), use formula (5) to calculate the multi-group scattering matrix, and use formula (6) to calculate other multi-group cross sections.

[0047]

[0048] in:

[0049] s---scattering reaction

[0050] g′---energy group number

[0051] ---Incident angle in Ω n interval, the cosine of the exit angle is μ, and the multi-group scattering matrix between energy group g' and energy group g.

[0052] f g'→g (μ)---the probability that the scattering reaction with the cosine of the exit angle μ will transfer the neutron energy from g' to energy group g

[0053] ---In the energy group g and the homogenized space z, the incident angle is Ω n Macroscopic cross section of x-type reaction produced by neutrons in the interval

[0054] ---Incident angle in Ω n Interval, energy group g' scattering reaction cross section

[0055] The step 5 includes: according to step 2 and step 4, generating a high-precision complex energy spectrum micro-pile multi-group database.

[0056] Parameters described in step 4 f g'→g (μ) is determined by formula (7) and formula (8),

[0057]

[0058] Furthermore, the modeling in step 1 uses a voxel construction method; the voxel construction method is used to establish a three-dimensional fine model of the target core, and the real boundary conditions are used to perform full-core continuous energy Monte Carlo transport calculations.

[0059] Furthermore, the calculation includes fuel and structure.

[0060] Furthermore,

[0061] The step 2, based on the calculations described in step 1, includes the reactivity and neutron flux in all required homogenization spaces of the fuel and structure.

[0062] The step 3, based on the calculations in step 1, includes all scattering events in the required homogenized space of the fuel and structure.

[0063] The step 4 is to calculate multiple groups of cross-sectional data of all required homogenization spaces including fuel and structures based on the neutron flux, reaction rate and scattering angle distribution obtained in the steps 2 and 3.

[0064] Furthermore, in step 5, a high-precision complex energy spectrum micro-reactor multi-group database is generated according to step 2 and step 4, including the neutron incident angle discrete method in step 2 and the multi-group cross-section data calculated in step 4, to generate a high-precision complex energy spectrum micro-reactor multi-group database.

[0065] Furthermore, the tracking of neutrons in different incident angle intervals in the designated homogenized region in step 2 includes discretizing a number of neutron incident angle intervals for the designated homogenized space, tracking the neutrons incident into the homogenized region during the calculation process, screening them according to the incident angle, determining the neutrons in different incident angle intervals, and counting the behaviors of the neutrons in different incident angle intervals in the homogenized region, and calculating the neutron flux and reactivity of the neutrons in the designated homogenized space.

[0066] Furthermore, the step 3 includes discretizing a number of neutron incident angle intervals for the specified homogenized space, tracking neutrons in different incident angle intervals respectively, and counting neutron scattering events in the specified homogenized space.

[0067] Furthermore, in step 4, based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3, multiple groups of cross-section data are generated for all homogenized spaces and all incident angle intervals.

[0068] The advantages of the present invention are:

[0069] The multi-group space regions of the present invention are divided by neutron incident angles, and the neutron incident angles are tracked, and the incident angle intervals generate multi-group cross-sectional data applicable to all cross-sections.

[0070] The present invention adopts real core boundary conditions when generating multiple group cross sections, thereby eliminating the influence of the boundary conditions on the accuracy of the multiple group cross section data.

[0071] The present invention counts the energy group neutron fluxes related to the incident direction in the continuous energy Monte Carlo calculation, effectively reducing the influence of the flux separable approximation on the accuracy of the generated multi-group cross-section data.

[0072] The present invention adopts a statistical method to characterize anisotropic scattering, thereby eliminating the influence of the truncation error of the Pn method causing a negative value of the scattering cross-section distribution function on the accuracy of the group constant. BRIEF DESCRIPTION OF THE DRAWINGS

[0073] Figure 1 A flow chart for producing a high-precision complex energy spectrum micro-reactor multi-group cross section. DETAILED DESCRIPTION

[0074] The technical solution of the present invention is described in more detail below in conjunction with the accompanying drawings. The present invention includes but is not limited to the following embodiments.

[0075] As attached Figure 1 As shown, a method for making a high-precision complex energy spectrum micro-pile multi-group cross section of the present invention comprises the following steps:

[0076] Step 1: Model the components or cores to be calculated and perform continuous energy Monte Carlo transport calculations based on ACE format cross-sectional data;

[0077] Step 2: Tracking neutrons at different incident angles within a specified homogenization region;

[0078] Step 3: According to step 2, the number of scattering events in different scattering angle cosine intervals is counted, the probability of the scattering angle cosine in each interval is calculated, and the scattering angle distribution is determined;

[0079] Step 4: According to step 2 and step 3, multiple groups of data are calculated;

[0080] Step 5: Create a multi-group database based on steps 2 and 4.

[0081] Among them, step 1 is as follows: for the components or cores to be calculated, fine geometric modeling is performed. The precise modeling of the model geometry is mainly carried out using the voxel construction method (CSG). The voxel construction method regards complex entities as being composed of several basic entities, such as spheres, cylinders, planes, etc., and is generated through Boolean operations. This method can strictly and accurately characterize the reactor geometry. Based on the evaluation nuclear database, such as the ENDF database, the ACE format cross-sectional data file is processed and obtained. According to the processed ACE format cross-sectional data file, the real core boundary conditions are set, and the full-heap continuous energy Monte Carlo neutron transport calculation is performed.

[0082] Step 2 is as follows: According to the calculation described in step 1, neutrons in different incident angle intervals in the specified homogenized region are tracked. Specifically, for the specified homogenized space, the neutron incident polar angle and amplitude angle are divided into several intervals (i.e., several discrete neutron incident angle intervals), and the neutrons incident into the homogenized region are tracked during the calculation process, and the neutrons in different incident angle intervals are determined according to the incident angle, and the behaviors of the neutrons in different incident angle intervals in the homogenized region are counted respectively, and the parameters z, n, Ω are calculated. n , g, r, E, ΔE, V are used as the screening parameters for the neutron flux in the specified incident angle range. The W, T, i, and li parameters are screened out through Monte Carlo transport calculation. The neutron flux in the specified incident angle range is calculated using formula (1). The parameters z, x, Ω n , g, ΔE, V, r, E are used as screening parameters for various types of reaction rates of neutrons in a specified homogenization region. W, i, wi, and Ag parameters are screened out through Monte Carlo transport calculation. Various types of reaction rates of neutrons in a specified homogenization region are calculated through formula (2). The Monte Carlo transport calculation can use a variety of methods, such as path length counting using formula (1) to count neutron flux. Various types of reaction rates of neutrons in different incident angle intervals in a specified homogenization region are counted through formula (2), and the various types of reaction rates include total reaction rate, absorption rate, scattering reaction rate, fission reaction rate, fission fraction, etc. Step 2 is calculated according to step 1, including the reaction rate and neutron flux in all required homogenization spaces of fuel and structure.

[0083]

[0084] ---In the range of energy group g, the incident angle is Ω n The flux of neutrons in the homogenized space z

[0085] ---Incident angle in Ω n The x-type reaction rate produced by neutrons in the interval in the energy group g and the homogenized space z

[0086] z---homogeneous spatial variation

[0087] n---incident angle discrete interval number

[0088] Ω n ---Discrete interval of the nth neutron incident angle

[0089] g---energy group number

[0090] r---spatial position

[0091] E---Neutron Energy

[0092] ΔE---the upper and lower bounds of the energy of energy group g

[0093] V---the volume of the homogenized space z

[0094] W---initial neutron total weight

[0095] T---the set of trajectories of all neutrons in the homogenized space volume

[0096] i---Event number

[0097] li---the length of the neutron's i-th trajectory

[0098] x---reaction type

[0099] wi---weight of neutron before reaction event i

[0100] Ag---the total set of all reaction events with neutron energy in energy group g

[0101] Step 3 is as follows: In the multi-group parameter statistics of step 2, anisotropic scattering is characterized by a statistical method for the scattering cross section. The anisotropy of scattering (i.e., scattering cross sections of different values) is determined by formula (4). The statistical method discretizes the cosine of the neutron scattering reaction into a number of neutron incident angle intervals. In the calculation, the number of scattering events of neutrons in different scattering incident angles or cosine intervals of the exit angle after the scattering reaction is tracked and counted. The frequency of scattering events in each scattering angle cosine interval is regarded as the probability of the exit angle cosine of the scattering reaction in each interval. The probability density distribution function of the scattering cross section with respect to the cosine of the scattering exit angle is calculated by formula (3) to determine the scattering angle distribution. After Monte Carlo transport calculation, the scattering events of neutrons in the specified homogenized space are counted by formula (4). The anisotropic scattering is represented in the form of a histogram distribution to determine the scattering cross section. Step 3, according to the calculation described in step 1, includes all the scattering events in the required homogenized space of the fuel and the structure. Among them, the statistical method to characterize anisotropic scattering is to find the distribution of scattering emission angles, and the scattering cross-section value is different at each emission angle.

[0102]

[0103] in:

[0104] ---Incident angle in Ω n The probability that the neutron scattering reaction has an exit angle cosine of μ in the interval satisfies the relation

[0105] ---Incident angle in Ω n The macroscopic cross section of the scattering reaction with the cosine of the scattering reaction angle μ is

[0106] μ---scattering reaction cosine of the emission angle

[0107] Δμn---the interval length of the nth scattering angle cosine discrete interval

[0108] Wμ---neutron weight with cosine of scattered emission angle μ

[0109] wμ---the number of events with a scattering angle cosine of μ

[0110] ---Incident angle in Ω n The macroscopic cross section of the scattering reaction of a neutron with energy E at r is

[0111] Step 4 is as follows: Based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3, the parameters are obtained by Monte Carlo transport calculation. f g'→g(μ), the multi-group scattering matrix, i.e., the scattering reaction cross section in the multi-group cross section, is calculated by formula (5), and other multi-group cross sections are calculated by formula (6). In step 4, based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3, the multi-group cross section data of all required homogenized spaces including fuel and structures are calculated. In step 4, based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3, multi-group cross section data are generated for all homogenized spaces and all incident angle intervals.

[0112]

[0113] in:

[0114] s---scattering reaction

[0115] g′---energy group number

[0116] ---Incident angle in Ω n interval, the cosine of the exit angle is μ, and the multi-group scattering matrix between energy group g' and energy group g is the scattering reaction cross section.

[0117] f g'→g (μ)---the probability that the scattering reaction with the cosine of the exit angle μ will transfer the neutron energy from g' to energy group g

[0118] ---In the energy group g and the homogenized space z, the incident angle is Ω n Macroscopic cross section of x-type reaction produced by neutrons in the interval

[0119] ---Incident angle in Ω n Interval, energy group g' scattering reaction cross section

[0120] The parameters f g'→g (μ) is determined by formula (7) and formula (8),

[0121]

[0122] in, Same as other sections, f g'→g (μ) is the inter-group transfer probability. Monte Carlo method is used to calculate the inter-group transfer probability, which can be obtained by counting the ratio of the number of neutrons that change from g' energy group to g energy group to the total number of neutrons in the reaction. as follows:

[0123]

[0124] Step 5 is as follows: according to the neutron incident angle discretization method in step 2 and the multi-group cross-section data calculated in step 4, a high-precision complex energy spectrum micro-reactor multi-group database is generated. The multi-group cross-section data calculated in step 4 is the multi-group cross-section data for all incident neutron angle intervals discretized in step 2. When the data is processed and produced, data of the number of energy groups × the number of incident angle discrete intervals will be generated.

[0125] The present invention can achieve the same accuracy by using fewer energy groups, thereby improving the efficiency of core multi-group calculation.

[0126] The present invention can obtain the required spatial homogenization region cross section of all target cores including fuel and structures through one full-core calculation.

[0127] The present invention is not limited to the above-mentioned specific implementation modes. A person skilled in the art may implement the present invention in various other specific implementation modes according to the embodiments and the disclosure of the drawings. Therefore, any design that adopts the design structure and concept of the present invention and makes some simple transformations or changes shall fall within the scope of protection of the present invention.

Claims

1. A method for making a high-precision complex energy spectrum micro-pile multi-group cross section, characterized by: The steps include: Step 1: Model the components or cores to be calculated and perform continuous energy Monte Carlo transport calculations based on ACE format cross-sectional data; Step 2: Tracking neutrons at different incident angles within a specified homogenization region; Step 3: According to step 2, the number of scattering events in different scattering angle cosine intervals is counted, the probability of the scattering angle cosine in each interval is calculated, and the scattering angle distribution is determined; Step 4: According to step 2 and step 3, multiple groups of data are calculated; Step 5: Create a multi-group database based on steps 2 and 4; The step 2 includes: tracking the neutrons in different incident angle intervals in the designated homogenization area according to the calculation in step 1, and converting z, n, Ω n , g, r, E, ΔE, V are used as the screening data of the neutron flux in the specified incident angle range. After Monte Carlo transport calculation, the W, T, i, li parameters are output. The neutron flux in the specified incident angle range is calculated by formula (1); z, x, Ω n , g, ΔE, V, r, E are used as screening data for various reaction rates of neutrons in a specified homogenized region, and after Monte Carlo transport calculation, W, i, wi, Ag parameters are output, and various reaction rates of the neutrons in the specified homogenized region are calculated by formula (2); in: ---Within the energy group g range, the incident angle is Ω n The flux of neutrons in the homogenized space z ---Incident angle in Ω n The x-type reaction rate produced by neutrons in the interval in the energy group g and the homogenized space z z---homogeneous spatial variation n---incident angle discrete interval number Ω n ---Discrete interval of the nth neutron incident angle g---energy group number r---spatial position E---Neutron Energy ΔE---the upper and lower bounds of the energy of energy group g V---the volume of the homogenized space z W---initial neutron total weight T---the set of trajectories of all neutrons in the homogenized space volume i---Event number li---the length of the neutron's i-th trajectory x---reaction type wi---weight of neutron before reaction event i Ag---the total set of all reaction events with neutron energy in energy group g The step 3 comprises: in the multi-group parameter statistics of step 2, discretizing the scattering exit angle cosine, counting the number of scattering events in different scattering exit angle cosine intervals, calculating the probability of the scattering exit angle cosine in each interval by formula (3), determining the scattering angle distribution, and determining the scattering cross section by formula (4) after Monte Carlo transport calculation; in: ---Incident angle in Ω n The probability that the neutron scattering reaction has an exit angle cosine of μ in the interval satisfies the relation ---Incident angle in Ω n The macroscopic cross section of the scattering reaction with the cosine of the scattering reaction angle μ is μ---scattering reaction cosine of the emission angle Δμn---the interval length of the nth scattering angle cosine discrete interval Wμ---neutron weight with cosine of scattered emission angle μ wμ---the number of events with a scattering angle cosine of μ ---Incident angle in Ω n The macroscopic cross section of the scattering reaction of a neutron with energy E at r is The step 4 comprises: obtaining parameters by Monte Carlo transport calculation based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3 f g'→g (μ), use formula (5) to calculate the multi-group scattering matrix, and use formula (6) to calculate other multi-group cross sections; in: s---scattering reaction g′---energy group number ---Incident angle in Ω n interval, the cosine of the exit angle is μ, and the multi-group scattering matrix between energy group g' and energy group g; f g'→g (μ)---the probability that the scattering reaction with the cosine of the exit angle μ will transfer the neutron energy from g' to energy group g ---In the energy group g and the homogenized space z, the incident angle is Ω n Macroscopic cross section of x-type reaction produced by neutrons in the interval ---Incident angle in Ω n Interval, energy group g' scattering reaction cross section The step 5 comprises: according to step 2 and step 4, generating a high-precision complex energy spectrum micro-pile multi-group database; Parameters described in step 4 f g'→g (μ) is determined by formula (7) and formula (8), 2. The method for making a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: The modeling described in step 1 uses a voxel construction method; the voxel construction method is used to establish a three-dimensional fine model of the target core, and real boundary conditions are used to perform full-core continuous energy Monte Carlo transport calculations.

3. The method for making a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: The calculation includes fuel and structure.

4. The method for producing a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 3, characterized in that: Step 2, calculating the reaction rate and neutron flux in all required homogenized spaces including fuel and structures according to step 1; Step 3, calculating, according to step 1, all scattering events in the required homogenized space including fuel and structure; The step 4 is to calculate multiple groups of cross-sectional data of all required homogenization spaces including fuel and structures based on the neutron flux, reaction rate and scattering angle distribution obtained in the steps 2 and 3.

5. The method for producing a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: In step 5, according to step 2 and step 4, a high-precision complex energy spectrum micro-reactor multi-group database is generated, including the neutron incident angle discrete method in step 2 and the multi-group cross-section data calculated in step 4, to generate a high-precision complex energy spectrum micro-reactor multi-group database.

6. The method for producing a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: The tracking of neutrons in different incident angle intervals in the designated homogenized region in step 2 includes discretizing a plurality of neutron incident angle intervals for the designated homogenized space, tracking the neutrons incident into the homogenized region during the calculation process, screening them according to the incident angle, determining the neutrons in different incident angle intervals, and counting the behaviors of the neutrons in different incident angle intervals in the homogenized region, and calculating the neutron flux and reaction rate of the neutrons in the designated homogenized space.

7. The method for producing a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: The step 3 includes discretizing a number of neutron incident angle intervals for a designated homogenized space, tracking neutrons in different incident angle intervals respectively, and counting neutron scattering events in the designated homogenized space.

8. The method for producing a high-precision complex energy spectrum micro-pile multi-group cross section according to claim 1, characterized in that: In step 4, based on the neutron flux, reaction rate and scattering angle distribution obtained in steps 2 and 3, multiple groups of cross-section data are generated for all homogenized spaces and all incident angle intervals.

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