Improved prefabricated section method for calculating multi-group scattering matrix

Abstract

The invention discloses an improved prefabricated section method for calculating a multi-group scattering matrix, which relates to the field of nuclear reactor physical calculation, and comprises the following steps: reading a continuous energy nuclear database, and establishing an incident energy grid of nuclide; according to different scattering mechanisms, integration is carried out on neutrons of different nuclide incident energy grids based on an energy group structure, and a nuclide prefabricated cross section is obtained; based on the nuclide prefabricated section, establishing and compressing a material joint energy grid, and calculating a material prefabricated section; and calculating a multi-group scattering matrix of the material according to the compressed material joint energy grid. The method comprises the following steps: establishing a nuclide incident energy grid; complex scattering calculation is simplified based on the integration process of the energy group structure, and the calculation complexity is reduced. Data integration from nuclides to a material level is achieved through establishment and compression of the material joint energy grid, the calculation efficiency is improved while scattering characteristics are reserved, the overall quality factor FOM value is increased by three times to the maximum degree, and the quality factor of scattering matrix calculation is effectively improved.

Description

Technical Field

[0001] The invention relates to the field of nuclear reactor physics calculation, in particular to an improved prefabricated cross-section method for calculating multi-group scattering matrices. Background Art

[0002] In the calculation of group constants for pebble bed high temperature gas-cooled reactors, the calculation of multi-group scattering matrices is a key link. Traditionally, this process relies on the Monte Carlo method to obtain statistical data by simulating a large number of neutron histories to calculate the multi-group scattering matrix. However, in this method, the elements in the multi-group scattering matrix are only updated when inter-group scattering occurs, so the statistical efficiency is low. Especially for pebble bed high temperature gas-cooled reactors, where the neutron energy spectrum is hard and the neutron leakage effect is significant, an energy group structure of 361 groups is required. This means that in order to obtain reliable results with lower statistical standard deviations, more neutron histories need to be simulated, resulting in an inefficient calculation process and more computational resources.

[0003] In the calculation of Monte Carlo method, there are two main ideas for reducing variance: one is to increase the number of samples, and the other is to use analytical or determined values ​​instead of sampled values. Based on these two ideas, the main variance reduction methods for scattering matrices are repeated collision method and prefabricated cross section method.

[0004] The prefabricated cross-section method is an idea that uses analytical or determined values ​​instead of sampled values. The basic idea is to determine the incident energy-exit energy group distribution of neutrons based on the neutron emission energy and angle information stored in the continuous energy nuclear database, and use this distribution in the calculation instead of the emission information sampling during simulation. The advantage of this is that, on the one hand, the use of a determined distribution can reduce the uncertainty of the results, and on the other hand, the number of samples can be increased by track length counting. However, the traditional prefabricated cross-section method requires accumulation for multiple nuclides in the statistical process of multi-group scattering matrices. Since the incident energy points selected by the prefabricated cross-section are different from the incident energy points in the continuous energy nuclear database, cross-section interpolation operations need to be performed on all nuclides during statistics, which greatly reduces the calculation efficiency.

[0005] In summary, although the prefabricated section method has the potential to reduce variance and improve statistical efficiency in theory, it still faces the problem of computational efficiency caused by section interpolation operations in practical applications. Improved methods for this problem still need further research and optimization in order to improve computational efficiency while maintaining computational accuracy. Summary of the invention

[0006] In view of the problems of low efficiency and high computing resource consumption in the existing multi-group scattering matrix calculation method, the present invention is proposed.

[0007] Therefore, the problem to be solved by the present invention is how to improve the efficiency of multi-group scattering matrix calculation and reduce the consumption of computing resources.

[0008] In order to solve the above technical problems, the present invention provides the following technical solutions:

[0009] In a first aspect, an embodiment of the present invention provides an improved prefabricated cross-section method for calculating a multi-group scattering matrix, which includes reading a continuous energy nuclear database to establish an incident energy grid for nuclides; integrating neutrons of different nuclides incident energy grids based on energy group structures according to different scattering mechanisms to obtain nuclide prefabricated cross-sections; based on the nuclide prefabricated cross-sections, establishing and compressing a material joint energy grid to calculate the material prefabricated cross-sections; and calculating the multi-group scattering matrix of the material based on the compressed material joint energy grid.

[0010] As a preferred scheme of the improved prefabricated cross-section method for calculating multi-group scattering matrices described in the present invention, the nuclide prefabricated cross-section includes the following steps: defining the upper and lower energy limits of the group structure, and traversing the nuclides and scattering mechanisms; extracting the neutron emission energy and angle information stored in the nuclear database for the scattering mechanism; integrating the emission energy and angle information, and converting the continuous energy into an energy group structure; calculating the nuclide prefabricated cross-section of each nuclide at each incident energy point based on the integration results, and storing the prefabricated cross-section data using a sparse matrix.

[0011] As a preferred solution of the improved prefabricated cross-section method for calculating multi-group scattering matrices of the present invention, wherein: the scattering mechanism includes elastic scattering and inelastic scattering; the specific formula of the integral is as follows:

[0012]

[0013] in, is the l-th order emission cross section of the g-th energy group of nuclide k at the incident energy E', MT k is the total number of scattering reactions of nuclide k, μ is the cosine of the neutron scattering angle, and f k,i (E'→E,μ) is the emission distribution of the nuclide k at the incident energy E', σ s,k,i (E') is the scattering cross section of the ith reaction channel of nuclide k at incident energy E', P l (μ) is the lth order Legendre polynomial.

[0014] As a preferred solution of the improved prefabricated section method for calculating the multi-group scattering matrix described in the present invention, wherein: the material joint energy grid contains all incident energy points of each nuclide; the construction process of the material joint energy grid includes the following steps: reading the nuclide and atomic number density of the material, and using set operations to merge the energy grids of the nuclides contained in this material to form a material joint energy grid; judging whether the energy points and the nuclide grids in the material joint energy grid match; when the energy points and the nuclide grids match, directly using the prefabricated values ​​of the material prefabricated section; when the energy points and the nuclide grids do not match, using the interpolation method to calculate the material prefabricated section; according to each incident energy point of the material joint energy grid, the material prefabricated section is calculated, and the specific formula is as follows:

[0015]

[0016] in, is the l-th order emission cross section of the material joint energy grid at the incident energy E', K is the number of nuclides in the material joint energy grid, N k is the nuclide k nuclear density in the material joint energy grid.

[0017] The sparse matrix algorithm is used to compress and store the material prefabricated sections to realize the cache mechanism.

[0018] As a preferred solution of the improved prefabricated cross-section method for calculating multi-group scattering matrices described in the present invention, it also includes: using the Monte Carlo method to calculate the neutron flux of the energy group and performing normalization processing; the energy group includes an incident energy group and an exit energy group; initializing the multi-group scattering matrix data structure, and using track length counting to calculate the scattering contribution of the incident energy group and the exit energy group using the nuclide prefabricated cross-section and neutron flux; normalizing the multi-group scattering matrix, calculating and counting the error, and performing matrix balance; if the error is less than the preset error, adjusting the Monte Carlo sampling parameters and recalculating the energy group neutron flux; if the error is greater than or equal to the preset error, outputting the result of the multi-group scattering matrix.

[0019] As a preferred solution of the improved prefabricated section method for calculating the multi-group scattering matrix of the present invention, the related formula of the multi-group scattering matrix is ​​as follows:

[0020]

[0021] Among them, Σ s,g'→g,l is the l-th order scattering cross section from the g′th energy group to the gth energy group, φ g' is the neutron flux density of the g′th energy group, w n,m is the weight of the nth neutron in the mth track, l n,m is the length of the nth neutron on the mth track.

[0022] As a preferred solution of the improved prefabricated cross-section method for calculating multi-group scattering matrices described in the present invention, the process of establishing the incident energy grid includes the following steps: continuing to use the Monte Carlo method to read the continuous energy nuclear database, integrating the emission information stored in the ACE nuclear database of different nuclides to form the incident energy points of the nuclides; extracting the incident energy points in the nuclear database, and merging the incident energy points of all nuclides to form an incident energy grid; and sorting and deduplicating the incident energy grid at the same time.

[0023] In the second aspect, an embodiment of the present invention provides an improved prefabricated section system for calculating a multi-group scattering matrix, which includes: a reading module for reading a continuous energy nuclear database and establishing an incident energy grid; an integration module for integrating the neutrons of the incident energy grid according to different scattering mechanisms and energy group structures to obtain a nuclide prefabricated section; an establishment module for establishing and compressing a material joint energy grid based on the nuclide prefabricated section to calculate the material prefabricated section; a calculation module for calculating the multi-group scattering matrix of the material based on the compressed material joint energy grid.

[0024] In a third aspect, an embodiment of the present invention provides a computer device comprising a memory and a processor, wherein the memory stores a computer program, wherein: when the computer program instructions are executed by the processor, the steps of the improved prefabricated cross-section method for calculating multi-group scattering matrices as described in the first aspect of the present invention are implemented.

[0025] In a fourth aspect, an embodiment of the present invention provides a computer-readable storage medium having a computer program stored thereon, wherein: when the computer program instructions are executed by a processor, the steps of the improved prefabricated cross-section method for calculating multi-group scattering matrices as described in the first aspect of the present invention are implemented.

[0026] The beneficial effects of the present invention are as follows: the present invention realizes a standardized energy reference frame by establishing a nuclide incident energy grid; the integration process based on the energy group structure simplifies complex scattering calculations and reduces the calculation complexity; the establishment and compression of the material joint energy grid realizes data integration from the nuclide to the material level, and improves the calculation efficiency while retaining the scattering characteristics. The present invention increases the overall quality factor FOM value by a maximum of 3 times, effectively improving the quality factor of the scattering matrix calculation. BRIEF DESCRIPTION OF THE DRAWINGS

[0027] In order to more clearly illustrate the technical solutions of the embodiments of the present invention, the following briefly introduces the drawings required for describing the embodiments. Obviously, the drawings described below are only some embodiments of the present invention. For ordinary technicians in this field, other drawings can be obtained based on these drawings without creative work. Among them:

[0028] Figure 1 This is a flow chart of the improved prefabricated cross-section method for calculating multi-group scattering matrices in Example 1.

[0029] Figure 2 This is a diagram of the computer equipment for the improved prefabricated cross-section method for calculating multi-group scattering matrices in Example 2. DETAILED DESCRIPTION

[0030] In order to make the above-mentioned objects, features and advantages of the present invention more obvious and easy to understand, the specific implementation methods of the present invention are described in detail below in conjunction with the accompanying drawings.

[0031] In the following description, many specific details are set forth to facilitate a full understanding of the present invention, but the present invention may also be implemented in other ways different from those described herein, and those skilled in the art may make similar generalizations without violating the connotation of the present invention. Therefore, the present invention is not limited to the specific embodiments disclosed below.

[0032] Secondly, the term "one embodiment" or "embodiment" as used herein refers to a specific feature, structure, or characteristic that may be included in at least one implementation of the present invention. The term "in one embodiment" that appears in different places in this specification does not necessarily refer to the same embodiment, nor does it refer to a separate or selective embodiment that is mutually exclusive with other embodiments.

[0033] Example 1

[0034] Reference Figure 1 , which is the first embodiment of the present invention, provides an improved prefabricated cross-section method for calculating a multi-group scattering matrix, comprising:

[0035] S1: Read the continuous energy nuclear database and establish the incident energy grid of the nuclide.

[0036] Specifically, the process of establishing the incident energy grid includes the following steps: continue to use the Monte Carlo method to read the continuous energy nuclear database, integrate the emission information stored in the ACE nuclear database of different nuclides, and form the incident energy points of the nuclides; extract the incident energy points in the nuclear database, and merge the incident energy points of all nuclides to form the incident energy grid; at the same time, sort and deduplicate the incident energy grid.

[0037] S2: According to different scattering mechanisms, the neutrons of different nuclides incident energy grids are integrated based on the energy group structure to obtain the nuclide prefabricated cross section.

[0038] Specifically, the nuclide prefabricated cross section includes the following steps: defining the upper and lower energy limits of the group structure, and traversing the nuclides and scattering mechanisms; extracting the neutron emission energy and angle information stored in the nuclear database for the scattering mechanism; integrating the emission energy and angle information, and converting the continuous energy into an energy group structure; calculating the nuclide prefabricated cross section of each nuclide at each incident energy point based on the integration results, and storing the prefabricated cross section data using a sparse matrix.

[0039] Furthermore, the scattering mechanism includes elastic scattering and inelastic scattering; the relevant formula for integrating the outgoing energy and angle information is as follows:

[0040]

[0041] in, is the l-th order emission cross section of the g-th energy group of nuclide k at the incident energy E', MT k is the total number of scattering reactions of nuclide k, μ is the cosine of the neutron scattering angle, and f k,i (E'→E,μ) is the emission distribution of the nuclide k at the incident energy E', σ s,k,i (E') is the scattering cross section of the ith reaction channel of nuclide k at incident energy E', P l (μ) is the lth order Legendre polynomial.

[0042] S3: Based on the nuclide prefabrication section, establish and compress the material joint energy grid and calculate the material prefabrication section.

[0043] Specifically, the material joint energy grid contains all incident energy points of each nuclide; the construction process of the material joint energy grid includes the following steps: reading the nuclide and atomic number density of the material, and using set operations to merge the energy grids of the nuclides contained in this material to form the material joint energy grid; judging whether the energy points in the material joint energy grid match the nuclide grid.

[0044] Furthermore, when the energy point and the nuclide grid match, the prefabricated value of the material prefabricated section is directly used; when the energy point and the nuclide grid do not match, the interpolation method is used to calculate the material prefabricated section.

[0045] Furthermore, the material prefabricated cross section is calculated based on each incident energy point of the material joint energy grid. The specific formula is as follows:

[0046]

[0047] in, is the l-th order emission cross section of the material joint energy grid at the incident energy E', K is the number of nuclides in the material joint energy grid, N kis the nuclide k nucleus density in the material joint energy grid;

[0048] Specifically, a sparse matrix algorithm is used to compress and store material prefabricated sections, and a cache mechanism is implemented to avoid repeated calculations.

[0049] S4: Calculate the multi-group scattering matrix of the material based on the compressed material joint energy grid.

[0050] Specifically, the neutron flux of the energy group is calculated using the Monte Carlo method and normalized; the energy group includes the incident energy group and the outgoing energy group; the multi-group scattering matrix data structure is initialized, and the scattering contribution of the incident energy group and the outgoing energy group is calculated using the nuclide prefabricated cross section and the neutron flux using the track length counting; the multi-group scattering matrix is ​​normalized, the error is calculated and statistically analyzed, and the matrix is ​​balanced;

[0051] Furthermore, if the error is less than the preset error, the Monte Carlo sampling parameters are adjusted and the energy group neutron flux is recalculated; if the error is greater than or equal to the preset error, the result of the multi-group scattering matrix is ​​output.

[0052] Furthermore, the relevant formula for the multi-group scattering matrix is ​​as follows:

[0053]

[0054] Among them, Σ s,g'→g,l is the l-th order scattering cross section from the g′th energy group to the gth energy group, φ g' is the neutron flux density of the g′th energy group, w n,m is the weight of the nth neutron in the mth track, l n,m is the length of the nth neutron on the mth track.

[0055] Furthermore, this embodiment also provides an improved prefabricated section system for calculating multi-group scattering matrices, including: a reading module, used to read a continuous energy nuclear database and establish an incident energy grid; an integration module, used to integrate the neutrons of the incident energy grid according to different scattering mechanisms and energy group structures to obtain the nuclide prefabricated section; an establishment module, based on the nuclide prefabricated section, establishes and compresses the material joint energy grid to calculate the material prefabricated section; a calculation module, used to calculate the multi-group scattering matrix of the material based on the compressed material joint energy grid.

[0056] Example 2

[0057] Reference Figure 2 , is an embodiment of the present invention, which is different from the previous embodiment in that:

[0058] If the functions are implemented in the form of software functional units and sold or used as independent products, they can be stored in a computer-readable storage medium. Based on this understanding, the technical solution of the present invention, or the part that contributes to the prior art or the part of the technical solution, can be embodied in the form of a software product. The computer software product is stored in a storage medium, including several instructions for a computer device (which can be a personal computer, a server, or a network device, etc.) to perform all or part of the steps of the methods described in each embodiment of the present invention. The aforementioned storage medium includes: U disk, mobile hard disk, read-only memory (ROM, Read-Only Memory), random access memory (RAM, Random Access Memory), disk or optical disk, etc., which can store program codes.

[0059] The logic and / or steps represented in the flowchart or otherwise described herein, for example, can be considered as an ordered list of executable instructions for implementing logical functions, and can be embodied in any computer-readable medium for use by an instruction execution system, device or apparatus (such as a computer-based system, a system including a processor, or other system that can fetch instructions from an instruction execution system, device or apparatus and execute instructions), or in conjunction with such instruction execution systems, devices or apparatuses. For the purposes of this specification, "computer-readable medium" can be any device that can contain, store, communicate, propagate or transmit a program for use by an instruction execution system, device or apparatus, or in conjunction with such instruction execution systems, devices or apparatuses.

[0060] More specific examples of computer-readable media (a non-exhaustive list) include the following: an electrical connection with one or more wires (electronic device), a portable computer disk case (magnetic device), a random access memory (RAM), a read-only memory (ROM), an erasable and programmable read-only memory (EPROM or flash memory), an optical fiber device, and a portable compact disk read-only memory (CDROM). In addition, the computer-readable medium may even be a paper or other suitable medium on which the program is printed, since the program may be obtained electronically, for example, by optically scanning the paper or other medium, followed by editing, deciphering or, if necessary, processing in another suitable manner, and then stored in a computer memory.

[0061] It should be understood that the various parts of the present invention can be implemented by hardware, software, firmware or a combination thereof. In the above-mentioned embodiments, a plurality of steps or methods can be implemented by software or firmware stored in a memory and executed by a suitable instruction execution system. For example, if implemented by hardware, as in another embodiment, it can be implemented by any one of the following technologies known in the art or their combination: a discrete logic circuit having a logic gate circuit for implementing a logic function for a data signal, a dedicated integrated circuit having a suitable combination of logic gate circuits, a programmable gate array (PGA), a field programmable gate array (FPGA), etc.

[0062] Example 3

[0063] Referring to Tables 1 to 3, which are the second embodiment of the present invention, this embodiment provides an improved prefabricated section method for calculating multi-group scattering matrices. In order to verify the beneficial effects of the present invention, scientific demonstration is carried out through economic benefit calculation and simulation experiments.

[0064] Specifically, a method for improving the statistical efficiency of multi-group scattering matrices was numerically tested using a pressurized water reactor fuel grid cell problem. The grid cell problem includes two examples. In the first example, the fuel area is filled with new materials, and in the second example, the fuel area is filled with burnup materials. A total of 48 major nuclides were selected. During the test, 70 groups of scattering matrices of the grid cell problem were statistically analyzed using the event estimator, repeated collision method, and improved prefabricated cross-section method. The multi-group scattering matrix statistically analyzed by the event estimator was used as the reference solution, and the number of repeated collisions in the repeated collision method was set to 10. In addition, for the event estimator, NECP-MCX invested a total of 1050 generations of particles, including 50 inactive generations, with 100,000 particles per generation; for the repeated collision method and the improved prefabricated cross-section method, NECP-MCX invested a total of 550 generations of particles, including 50 inactive generations, with 100,000 particles per generation.

[0065] Furthermore, the statistical efficiency of different methods is compared, and the comparison standard is the quality factor FOM value. The larger the FOM value, the higher the calculation efficiency. Based on the FOM value of the event estimator, as shown in Tables 1 and 2, the calculation results of the overall FOM value of the scattering matrix are given. The overall FOM value is the sum of the FOM values ​​of each element of the scattering matrix.

[0066] Table 1 Comparison of overall FOM values ​​of zero-burnup gate elements

[0067] feature Zero-order scattering matrix First-order scattering matrix <![CDATA[FOM 重复碰撞 / FOM 事件估计 ]]> 1.274 1.517 <![CDATA[FOM 预制截面 / FOM 事件估计 ]]> 8.520 13.006

[0068] Table 2 Comparison of overall FOM values ​​of gate elements with burnup

[0069] feature Zero-order scattering matrix First-order scattering matrix <![CDATA[FOM 重复碰撞 / FOM 事件估计 ]]> 1.518 1.816 <![CDATA[FOM 预制截面 / FOM 事件估计 ]]> 11.802 17.960

[0070] Furthermore, the FOM values ​​of the improved prefabricated section method and the original prefabricated section method are compared. Since the relative statistical deviations calculated by the two methods are basically the same, only the overall FOM values ​​are compared. As shown in Table 3, the comparison results of the two examples are given. It can be seen that due to the shortened calculation time of the section lookup table, the FOM value of the improved prefabricated section method has been improved to a certain extent. For the problem of belt fuel consumption, due to the large number of nuclides, the FOM value has been increased by nearly 3 times.

[0071] Table 3 Comparison of FOM values ​​of prefabricated section method

[0072] feature Improved prefabricated section method FOM value increased multiples Zero-order scattering matrix of zero-burnup grid element 1.233 Zero-burnup grid element first-order scattering matrix 1.233 Zero-order scattering matrix with burnup elements 2.917 First-order scattering matrix with burnup elements 2.918

[0073] It should be noted that the above embodiments are only used to illustrate the technical solutions of the present invention rather than to limit it. Although the present invention has been described in detail with reference to the preferred embodiments, those skilled in the art should understand that the technical solutions of the present invention may be modified or replaced by equivalents without departing from the spirit and scope of the technical solutions of the present invention, which should all be included in the scope of the claims of the present invention.

Claims

1. An improved prefabricated cross-section method for calculating multi-group scattering matrices, characterized in that: include, Read the continuous energy nuclear database and establish the incident energy grid of the nuclide; According to different scattering mechanisms, the neutrons of different nuclides incident energy grids are integrated based on the energy group structure to obtain the nuclide prefabricated cross section; Based on the nuclide prefabrication section, the material joint energy grid is established and compressed, and the material prefabrication section is calculated; Based on the compressed material joint energy grid, the multi-group scattering matrix of the material is calculated.

2. The improved prefabricated cross-section method for calculating multi-group scattering matrices according to claim 1, characterized in that: The nuclide prefabrication section comprises the following steps: Define the upper and lower energy limits of the group structure and iterate over the nuclides and scattering mechanisms; Extract the neutron emission energy and angle information stored in the nuclear database for the scattering mechanism; Integrate the outgoing energy and angle information and convert the continuous energy into energy group structure; According to the integration results, the nuclide prefabrication cross section of each nuclide at each incident energy point is calculated, and the prefabrication cross section data is stored using a sparse matrix.

3. The improved prefabricated cross-section method for calculating multi-group scattering matrices according to claim 2, characterized in that: The scattering mechanism includes elastic scattering and inelastic scattering; the specific formula of the integral is as follows: in, is the l-th order emission cross section of the g-th energy group of nuclide k at the incident energy E', MT k is the total number of scattering reactions of nuclide k, μ is the cosine of the neutron scattering angle, and f k,i (E'→E,μ) is the emission distribution of the nuclide k at the incident energy E', σ s,k,i (E') is the scattering cross section of the ith reaction channel of nuclide k at incident energy E', P l (μ) is the lth order Legendre polynomial.

4. The improved prefabricated cross-section method for calculating multi-group scattering matrices according to claim 1, characterized in that: The material joint energy grid includes all incident energy points of each nuclide; the construction process of the material joint energy grid includes the following steps: Read the nuclide and atomic number density of the material, and use the collection operation to merge the energy grids of the nuclides contained in this material to form the material joint energy grid; Determine whether the energy points in the material joint energy grid and the nuclide grid match; When the energy point and nuclide grid match, the prefabricated value of the material prefabricated section is used directly; When the energy points and nuclide grids do not match, the material precast cross-section is calculated using an interpolation method; According to each incident energy point of the material joint energy grid, the material prefabricated cross section is calculated. The specific formula is as follows: in, is the l-th order emission cross section of the material joint energy grid at the incident energy E', K is the number of nuclides in the material joint energy grid, N k is the nuclide k nucleus density in the material joint energy grid; The sparse matrix algorithm is used to compress and store the material prefabricated sections to realize the cache mechanism.

5. The improved prefabricated section method for calculating multi-group scattering matrices according to claim 4, characterized in that: Also includes, The neutron flux of energy groups is calculated using the Monte Carlo method and normalized; The energy group includes an incident energy group and an outgoing energy group; Initialize the multi-group scattering matrix data structure and calculate the scattering contributions using nuclide preformed cross sections and neutron flux for both the incident and outgoing energy groups using track length counts; Normalize the multi-group scattering matrix, calculate and statistically analyze the errors, and perform matrix balance; If the error is less than the preset error, the Monte Carlo sampling parameters are adjusted and the energy group neutron flux is recalculated; If the error is greater than or equal to the preset error, the result of the multi-group scattering matrix is ​​output.

6. The improved prefabricated cross-section method for calculating multi-group scattering matrices according to claim 5, characterized in that: The specific formula of the multi-group scattering matrix is ​​as follows: Among them, Σ s,g'→g,l is the l-th order scattering cross section from the g′th energy group to the gth energy group, φ g' is the neutron flux density of the g′th energy group, w n,m is the weight of the nth neutron in the mth track, l n,m is the length of the nth neutron on the mth track.

7. The improved prefabricated cross-section method for calculating multi-group scattering matrices according to claim 1, characterized in that: The process of establishing the incident energy grid includes the following steps: Continue to use the Monte Carlo method to read the continuous energy nuclear database, integrate the emission information stored in the ACE nuclear database of different nuclides, and form the incident energy point of the nuclide; Extract the incident energy points in the nuclear database and merge the incident energy points of all nuclides to form an incident energy grid; The incident energy grids are sorted and deduplicated at the same time.

8. An improved prefabricated section system for calculating multi-group scattering matrices, based on the improved prefabricated section method for calculating multi-group scattering matrices according to any one of claims 1 to 7, characterized in that: include, A reading module is used to read the continuous energy nuclear database and establish the incident energy grid of the nuclide; The integration module is used to integrate the neutrons of different nuclides incident energy grids based on the energy group structure according to different scattering mechanisms to obtain the nuclide prefabricated cross section; Establish a module, establish and compress the material joint energy grid based on the nuclide prefabrication section, and calculate the material prefabrication section; The calculation module is used to calculate the multi-group scattering matrix of the material according to the compressed material joint energy grid.

9. A computer device comprising a memory and a processor, wherein the memory stores a computer program, wherein: When the processor executes the computer program, the steps of the improved prefabricated cross-section method for calculating multi-group scattering matrices according to any one of claims 1 to 7 are implemented.

10. A computer-readable storage medium having a computer program stored thereon, characterized in that: When the computer program is executed by a processor, the steps of the improved prefabricated cross-section method for calculating multi-group scattering matrices according to any one of claims 1 to 7 are implemented.

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