Method for making neutron-photon nuclear database suitable for reactor multi-group monte carlo calculation

Abstract

The application discloses a method for manufacturing a neutron-photon nuclear database suitable for reactor multi-group Monte Carlo calculation, and the method comprises the following steps: generating first format data of a single nuclear element by using an evaluation nuclear database; merging first format data of a plurality of nuclear elements obtained based on the first format data of the single nuclear element to obtain a database of the first format data; performing data processing on the database of the first format data to obtain a database of second format data related to a problem; and converting the database of the second format data into a database of third format data by using a preset format conversion program. The application improves the efficiency of the accompanying calculation of the Monte Carlo program, further improves the shielding calculation precision of the Monte Carlo program, checks the multi-group determinism, verifies the multi-group parameter processing function of the determinism, inherits the advantages of the Monte Carlo program, and has a powerful geometric processing function.

Description

Technical Field

[0001] This invention relates to the field of database creation technology, and in particular to a method for creating a neutron photon nucleus database suitable for multi-group Monte Carlo computation in reactors. Background Technology

[0002] Currently, the multi-group nuclear database used by the Monka program is mainly the multi-group ACE format database mgxsnp provided by Los Alamos National Laboratory in the United States. This database is created by mounting the evaluation nuclear database ENDF / BV released in 1994 onto the NJOY program. The NJOY parameters are selected as follows: (1) Number of energy groups: 30 groups of photons and 12 groups of neutrons (2) Legendre expansion order: P4 (3) Weight spectrum: CLAW (also called TD-Division Weight Function) (4) Nuclide types: 95 (5) Temperature 300K; Disadvantages of existing technology:

[0003] The outdated data sources of the multi-group nuclear database mgxsnp primarily rely on the 1994 version ENDF / BV, and the cross-section processing program is the 1999 NJOY99 version. However, recent years have seen significant updates and improvements to evaluation databases and cross-section processing programs worldwide. Therefore, these outdated versions limit the accuracy of the multi-group nuclear database. Furthermore, the incomplete theoretical model stems from the fact that early versions of mgxsnp were primarily designed for fast neutron spectral systems, neglecting the crucial upscattering effect at hot groups. Additionally, mgxsnp also ignored the resonance self-shielding effect, which significantly impacts the neutron flux density of both resonance and hot groups. Therefore, the neglected upscattering and resonance self-shielding effects limit the application of mgxsnp in low- and medium-energy neutron spectral systems. The lack of tools for developing the multi-group kernel database mgxsnp has led to the official reliance on the CRSRD program. However, the source code for this program has not been released publicly, resulting in other research institutions' over-reliance on the mgxsnp multi-group kernel database and limiting further improvements in the computational accuracy of multi-group Monte Carlo programs. Summary of the Invention

[0004] The present invention aims to at least partially solve one of the technical problems in the related art.

[0005] Therefore, the purpose of this invention is to propose a method for creating a neutron-photon nuclear database suitable for reactor multi-group Monte Carlo calculations, which can significantly improve the calculation accuracy and efficiency of neutron-photon coupling transport in reactor multi-groups.

[0006] Another objective of this invention is to propose a neutron photon nucleus database creation system suitable for reactor multi-group Monte Carlo computation.

[0007] To achieve the above objectives, this invention proposes a method for creating a neutron photon nucleus database suitable for multi-group Monte Carlo computation in reactors, comprising:

[0008] First-format data for individual nuclides are generated using an evaluation nuclear database;

[0009] A database of first-format data of multiple nuclides obtained based on the first-format data of the single nuclide is obtained by merging the first-format data of the first nuclide.

[0010] Data processing is performed on the database of the first format data to obtain a database of the second format data related to the problem;

[0011] A database of data in the second format is converted into a database of data in the third format using a preset format conversion program.

[0012] The method for creating a neutron photon nucleus database for reactor multi-group Monte Carlo computation according to embodiments of the present invention may also have the following additional technical features:

[0013] Furthermore, in one embodiment of the present invention, after obtaining the database of the third-format data, the method further includes:

[0014] The test results were obtained by testing the database of the third-format data based on the Monte Carlo program.

[0015] The accuracy test results of the database of the third-format data are obtained based on the neutron and photon calculation results in the test results.

[0016] Furthermore, in one embodiment of the present invention, generating first-format data for a single nuclide using an evaluation nuclear database includes:

[0017] The resonance cross section is obtained by converting the resonance parameters in the evaluation kernel database;

[0018] The cross-section is subjected to Doppler broadening, and the scattering matrix and scattering cross-section of the nuclide are thermally processed. The cross-section of the indistinguishable resonance region is then approximated with a narrow resonance to obtain the cross-section processing result.

[0019] Based on the cross-section processing results, the continuous energy point cross-sections are grouped to obtain multi-group neutron data and photon cross-section data in multi-group form, so as to convert the multi-group neutron data and photon cross-section data into the first format data.

[0020] Furthermore, in one embodiment of the present invention, the step of processing the database of the first format data to obtain a database of problem-related second format data includes:

[0021] Using the BONDARENKO baseline cross-section iteration method, the database of the first format data generated by the NJOY program, which is a MATXS format multi-group kernel database, is transformed into a second format data database, which is an ANISN format multi-group kernel database.

[0022] Furthermore, in one embodiment of the present invention, the step of processing the database of the first format data to obtain a database of problem-related second format data includes:

[0023] The corresponding multi-group cross sections are calculated based on the preset dilution cross section parameters in the database of the first format data.

[0024] Calculate the dilution cross section parameters related to the actual problem based on the nuclide nucleon density ratio of the actual problem, and interpolate to obtain the multi-group cross section parameters for the corresponding actual problem;

[0025] A database of the second format data is obtained by calculating based on the multi-group cross-sectional parameters.

[0026] To achieve the above objectives, another aspect of the present invention proposes a neutron photon nucleus database creation system suitable for reactor multi-group Monte Carlo computation, comprising:

[0027] The initial data generation module is used to generate first-format data for a single nuclide using the evaluation nucleus database;

[0028] The first database generation module is used to merge the first format data of multiple nuclides obtained based on the first format data of the single nuclide to obtain a database of first format data.

[0029] The second database generation module is used to process the database of the first format data to obtain a database of the second format data related to the problem.

[0030] The third database generation module is used to convert the database of the second format data into a database of the third format data using a preset format conversion program.

[0031] The present invention provides a method and system for creating a neutron photon nucleus database for multi-group Monte Carlo calculations in reactors. This method improves the efficiency of the Monte Carlo program's adjoint calculations and further enhances the accuracy of the Monte Carlo program's shielding calculations. It also verifies the multi-group parameter processing capabilities of deterministic theory for multi-group calculations and inherits the advantages of the Monte Carlo program, possessing powerful geometric processing capabilities.

[0032] Additional aspects and advantages of the invention will be set forth in part in the description which follows, and in part will be obvious from the description, or may be learned by practice of the invention. Attached Figure Description

[0033] The above and / or additional aspects and advantages of the present invention will become apparent and readily understood from the following description of the embodiments taken in conjunction with the accompanying drawings, wherein:

[0034] Figure 1 This is a flowchart illustrating a method for creating a neutron photon nucleus database suitable for reactor multi-group Monte Carlo computation according to an embodiment of the present invention;

[0035] Figure 2 A logic diagram of a method for creating a neutron photon nucleus database for reactor multi-group Monte Carlo computation according to an embodiment of the present invention;

[0036] Figure 3 This is a flowchart of the NJOY process according to an embodiment of the present invention;

[0037] Figure 4 This is a flowchart of the ANISN to ACE conversion procedure according to an embodiment of the present invention;

[0038] Figure 5 According to the embodiments of the present invention 238 Schematic diagram of the input card algorithm for the ANISN to ACE tool in U-pure neutron mode;

[0039] Figure 6 This is a schematic diagram of the geometric structure of an infinitely large dielectric sphere according to an embodiment of the present invention;

[0040] Figure 7 According to the embodiments of the present invention 56 Schematic diagram of neutron flux density distribution in Fe media MCNP6.1 and RMC;

[0041] Figure 8 According to the embodiments of the present invention 56 Schematic diagram of the relative deviation between neutron flux density and MCNP6.1 for Fe medium RMC-mounted mgxsnp and MACE multigroup libraries;

[0042] Figure 9 According to the embodiments of the present invention 238 Schematic diagram of neutron flux density distribution in U-dielectrics MCNP6.1 and RMC;

[0043] Figure 10 According to the embodiments of the present invention 238 Schematic diagram of the relative deviation between neutron flux density of U-medium RMC-mounted mgxsnp and MACE multigroup libraries and MCNP6.1;

[0044] Figure 11 According to the embodiments of the present invention nat Schematic diagram of photon flux density distribution in Fe dielectric MCNP6.1 and RMC;

[0045] Figure 12 According to the embodiments of the present invention nat Schematic diagram of the relative deviation between the photon flux density of Fe dielectric RMC-mounted mgxsnp and MACE multigroup library and MCNP6.1;

[0046] Figure 13 According to the embodiments of the present invention nat Schematic diagram of photon flux density distribution in Fe dielectric MCNP6.1 and RMC;

[0047] Figure 14 According to the embodiments of the present invention nat Schematic diagram of the relative deviation between the photon flux density of Fe dielectric RMC-mounted mgxsnp and MACE multigroup library and MCNP6.1;

[0048] Figure 15 This is a schematic diagram of the neutron flux density before and after considering the upscattering effect according to an embodiment of the present invention;

[0049] Figure 16 This is a schematic diagram showing the relative deviation of neutron flux density before and after considering the upscattering effect according to an embodiment of the present invention;

[0050] Figure 17 This is a schematic diagram of the geometric structure of the SINBAD series OKTAVIAN-A1 and Iron shielding examples according to embodiments of the present invention;

[0051] Figure 18 This is a schematic diagram of subsurface flow in the OKTAVIAN-Al example according to an embodiment of the present invention;

[0052] Figure 19 This is a schematic diagram of photonic surface flow in the OKTAVIAN-Al example according to an embodiment of the present invention;

[0053] Figure 20 This is a schematic diagram of the relative deviation of sub-currents in the OKTAVIAN-Al example according to an embodiment of the present invention;

[0054] Figure 21 This is a schematic diagram of the relative deviation of sub-currents in the OKTAVIAN-Al example according to an embodiment of the present invention;

[0055] Figure 22 This is a schematic diagram of subsurface flow in the OKTAVIAN-Iron example according to an embodiment of the present invention;

[0056] Figure 23This is a schematic diagram of the relative deviation of the subsurface flow in the OKTAVIAN-Iron example according to an embodiment of the present invention;

[0057] Figure 24 This is a schematic diagram of the system structure for creating a neutron photon nucleus database for reactor multi-group Monte Carlo computation according to an embodiment of the present invention. Detailed Implementation

[0058] It should be noted that, unless otherwise specified, the embodiments and features described in this application can be combined with each other. The present invention will now be described in detail with reference to the accompanying drawings and embodiments.

[0059] To enable those skilled in the art to better understand the present invention, the technical solutions of the present invention will be clearly and completely described below with reference to the accompanying drawings of the embodiments of the present invention. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort should fall within the scope of protection of the present invention.

[0060] The following describes, with reference to the accompanying drawings, a method and system for creating a neutron photon nucleus database suitable for reactor multi-group Monte Carlo computation according to an embodiment of the present invention.

[0061] Figure 2 This is a logic diagram of the method for creating a neutron photon nucleus database applicable to multi-group Monte Carlo computation in reactors, as described in this invention. Figure 2 As shown, firstly, the NJOY program is used to mount the corresponding evaluation nuclear database to generate a MATXS data file for a single nuclide. Then, the BBC program is used to merge the MATXS files of multiple nuclides into a single MATXS database. Next, the TRANSX input card is filled in according to the actual calculation model, and the corresponding MATXS database is mounted for processing to obtain the problem-related ANISN database. In addition, a self-made multi-group ANISN to ACE conversion tool is used to realize the conversion of the multi-group ACE format nuclear database. Finally, the Monte Carlo program MCNP6.1 and RMC are used to mount the corresponding multi-group ACE database for testing. The neutron and photon calculation results of the infinite dielectric sphere model and the SINBAD series of shielded calculation examples are compared to test the calculation accuracy and applicability of the self-made multi-group Monte Carlo database.

[0062] Figure 1 This is a flowchart of a method for creating a neutron photon nucleus database for reactor multi-group Monte Carlo computation according to an embodiment of the present invention.

[0063] like Figure 1 As shown, the method includes, but is not limited to, the following steps:

[0064] S1, using the evaluation nuclear database to generate first-format data for a single nuclide;

[0065] S2, a database of first-format data obtained by merging first-format data of multiple nuclides based on first-format data of a single nuclide;

[0066] S3, perform data processing on the database of first-format data to obtain a database of second-format data related to the problem;

[0067] S4, using a preset format conversion program, converts the database of data in the second format to a database of data in the third format.

[0068] It is understood that the data sources for this invention have been updated: This invention will use the updated and more stable evaluation nuclear database version ENDF / B-VII.1 and the NJOY2016 cross-section processing program to develop and update the multi-group neutron and photon nuclear database; theoretical model optimization: In the process of database development, this invention will consider the influence of upscattering effects and use the multi-group cross-section post-processing program TRANSX to process the resonant self-screen cross-section; conversion tool development: This invention will develop a neutron and photon multi-group nuclear database conversion tool to convert the multi-group ANISN format nuclear database generated by the TRANSX program into the multi-group ACE format nuclear database used by the Monte Carlo program.

[0069] Specifically, this invention includes the following: Multi-group neutron and photon nuclear database updates: Based on the ENDF / B-VII.1 evaluation nuclear database and the NJOY2016 cross-section processing program, a multi-group MATXS database of 30 neutron groups and 12 photon groups is developed, considering the upscattering effect on 1H; Resonance self-shielding effect processing: The MATXS database is mounted using the multi-group cross-section post-processing program TRANSX to generate a problem-related multi-group ANISN database; Multi-group ACE database generation: A conversion tool is developed to convert the problem-related multi-group ANISN database into a multi-group ACE database usable by the Monte Carlo program; Multi-group ACE database testing: The self-made and official multi-group ACE databases are mounted using the Monte Carlo program MCNP and RMC for testing and verification.

[0070] The following describes in detail, with reference to the accompanying drawings, a method for creating a neutron photon nucleus database suitable for reactor multi-group Monte Carlo computation according to an embodiment of the present invention.

[0071] (1) Update of multi-group neutron and photon nucleus database: Currently, the evaluation nucleus database versions released by various countries are shown in Table 1. This invention will select the evaluation nucleus database ENDF / B-VII.1, which has undergone extensive testing and verification, to develop a multi-group neutron and photon nucleus database.

[0072] Table 1

[0073]

[0074] Understandably, the specific parameter information for NJOY input parameters is as follows: 1. Energy group structure: 30 neutron groups and 12 photons; 2. Weight spectrum: iwt = 9, TD-Division Weight Function (also known as "CLAW"); 3. Legendre expansion order: P4; 4. Dilution cross section: 1.0E+10, 1.0E+05, 1.0E+04, 1.0E+03, 3.0E+02, 1.0E+02, 30, 10, 3, 1; 5. Temperature: 300K; 6. For 1H, thermal evaluation library data needs to be read, and the maximum thermal treatment energy limit is 4eV.

[0075] Specifically, the NJOY flowchart is as follows: Figure 3 As shown, the modules are as follows: MODER is the binary conversion module, whose main function is to convert decimal files in the evaluation kernel database into binary; RECONR is the resonance reconstruction module, whose main function is to convert resonance parameters in the evaluation kernel database into resonance cross sections; BROADR is the Doppler broadening module, whose main function is to perform Doppler broadening on the cross sections; HEATR is the calorific value processing module, whose main function is to generate calorific value and KERMA factor data and add them to the PENDF file; GASPR is the gas processing module, whose main function is to add gas generation reaction channels to the PENDF file; and THERMR is the thermal processing module, whose main function is to process the scattering matrix of nuclides. The module performs thermal processing on the scattering cross section; UNRESR is the indistinguishable resonance processing module, whose main function is to perform narrow resonance approximation processing on the cross section of the indistinguishable resonance region; GROUPR is the neutron multi-group cross section generation module, whose main function is to merge the continuous energy point cross sections processed by the previous modules to obtain multi-group cross section data; GAMINR is the photon multi-group cross section generation module, whose main function is similar to GROUPR, generating photon cross section data in multi-group form; MATXSR is the MATXS format data generation module, whose main function is to convert the multi-group neutron and photon cross section data generated by the GROUPR and GAMINR modules into multi-group MATXS format data.

[0076] (2) Resonance Self-Shielding Effect Processing: This invention employs the multi-group cross-section post-processing program TRANSX to process the resonance self-shielding effect, thereby converting the MATXS format multi-group kernel database generated by the NJOY program into an ANISN format multi-group kernel database. Currently, the TRANSX program mainly uses the BONDARENKO background cross-section iteration method to process the resonance self-shielding effect. This method is mainly based on the Boltzmann transport equation in BN form, obtaining higher-order fluxes through narrow resonance approximation:

[0077]

[0078] At this point, the expression for the multi-group cross section is as follows:

[0079]

[0080] Where i is a nuclide, N i Let C(E) be the nucleon density of the corresponding nuclide, and let C(E) be a continuous function.

[0081] In addition, the dilution section is defined as Substituting this into formula (2), we can obtain the expression for the multi-group cross-section as follows:

[0082]

[0083] Furthermore, the NJOY program pre-calculates ten different dilution cross-section parameters and calculates the corresponding multi-group cross-sections, which are then stored in the MATXS database. The TRANSX program, based on the nuclide nucleon density ratio of the actual problem, calculates the dilution cross-section parameters relevant to the actual problem, and then uses these dilution cross-sections for interpolation to obtain the multi-group cross-section parameters for the corresponding actual problem. Finally, the multi-group cross-sections relevant to the actual problem are stored in the ANISN database for subsequent transport programs to perform calculations.

[0084] Understandably, when using the TRANSX program, the following should be noted: for fissile materials, the fission spectrum and average fission neutron number need to be output separately; for the processing of scattering effects on the MATXS database of 30 neutron groups and 12 photons related to this invention, the number of scattering energy groups needs to be set to 5; the neutron absorption cross section needs to be the cross section output by the TRANSX pure neutron mode, otherwise it will lead to excessive statistical error.

[0085] (3) Generation of Multi-Group ACE Database: To achieve the generation of a multi-group ACE database, this invention developed a conversion program specifically for converting a multi-group ANISN format database generated by the TRANSX program into a multi-group ACE format. The specific flowchart is as follows: Figure 4As shown. `input.yaml` is the input card file in YAML format. Because the ANSIN database file output by TRANSX only contains one-dimensional cross-section and two-dimensional scattering matrix data, additional parameters such as energy group structure, nuclide ID, and relative atomic mass are required. Therefore, these parameters are stored in the `input.yaml` file. `main` is the main program of the conversion program, which reads the parameters from the input card file. `process_neutron_data`, `process_photon_data`, and `process_neutron_photon_data` are subroutines that process the neutron, photon, and neutron-photon coupling data in the ANSIN file, selecting the corresponding subroutine based on the parameters in the input card. The `format_anisn` subroutine formats the ANISN database. Since the ANISN data format is an 80-column punched structure suitable for early computers, it needs to be converted to an exponential decimal structure readable by modern computers. The corresponding cross-sectional parameters are printed in the `format_anisn` file for initial comparison. The `read_anisn` subroutine reads the ANISN database, performing preliminary processing and reading of the data. The `anisn_convert_ace` subroutine converts the ANISN database to an ACE database. `multigroup_ace` and `xsdir` are the multigroup ACE format file and corresponding index file output by the program.

[0086] Furthermore, Figure 5 The example input card of the input.yaml file is shown, which mainly contains three categories of information: general, mt, and material. Among them, general mainly stores the general information of the ANISN file to be converted, mt mainly stores the information of each reaction channel, and material mainly stores the material information of the output.

[0087] The general section is described in detail as follows: Type: indicates the ANISN data type, neutron represents pure neutron, photon represents pure photon, and neutron-photon represents neutron-photon coupling mode; Neutron_group: indicates the total number of neutron energy groups; Photon_group: indicates the total number of photon energy groups; Table_length: indicates the total length of each energy group in the ANISN database under pure neutron and neutron-photon coupling modes; Table_length_photon: indicates the total length of each energy group in the ANISN database under pure photon mode; Upscattering_num: number of upscattering energy groups; Downscattering_num: number of downscattering energy groups; Lord_num: maximum Legendre expansion order; average_nubar: average fission neutron count indicator, =0 indicates no fission, =1 indicates total fission neutron count, =2 indicates providing both instantaneous fission neutron count and total fission neutron count; Neutron_energy: neutron energy group structure, note: needs to be identified by square brackets "[]"; Photon_energy: photon energy group structure.

[0088] The specific description of the mt part is as follows: Mt_num: indicates the number of reaction channels contained. It is 3 for non-fissile materials and 5 for fissile materials; Mt_name: identifies the specific reaction channel information. For non-fissile materials, only the absorption cross section "abs", neutron production cross section "nufis", and total cross section "total" need to be provided. However, for fissile materials, the fission spectrum "chi" and the average number of fission neutrons "nubar" also need to be provided.

[0089] The material section is described in detail below: Material_num: Represents the total number of materials in the ANISN database; ID: Represents the material ID in the ANISN database, which will be output as the first line of the multi-group ACE file; Atomic_weight: Represents the relative atomic weight of the material in the ANISN database. This parameter is very important for mixed materials. If the ANISN database contains mixed materials, it needs to be calculated according to the corresponding atomic ratio. For example: M mix =M1w1+M2w2+...+M n w n Among them: M i w represents the relative atomic mass of nuclide i. i Indicates the mass fraction of nuclide i; Suffix: indicates the nuclide suffix of the output, m indicates neutron data, g indicates photon data.

[0090] (4) Multi-group ACE Database Testing To test the accuracy of the self-made multi-group ACE database, this invention selected two main categories of benchmark problems for testing: the infinite dielectric sphere model and the SINBAD series of shielded benchmark problems. The Monte Carlo program MCNP6.1 developed by Los Alamos National Laboratory in the United States and the Monte Carlo program RMC with independent intellectual property rights were used for calculations. The official MCNP multi-group database mgxsnp and the self-made multi-group database MACE were also used for calculations to test the stability and applicability of the self-made database.

[0091] Specifically, the infinitely large dielectric sphere model example: This example consists of a sphere with a radius of 10... 9 The geometric composition of a sphere in cm (geometric structure such as...) Figure 6 As shown), the sphere is uniformly filled with 1g / cm³ 3 of 56 Fe and 238 U (To test the performance of RMC's multi-group neutron-photon transport calculation function under different mass nuclei), an isotropic neutron source with an energy of 16 MeV was placed at the center of the sphere with 5,000,000 particles. Both RMC and MCNP6.1 were enabled in pure neutron and pure photon transport calculation modes. Finally, the neutron and photon flux of the entire region was calculated.

[0092] Preferably, neutron transport testing: Figures 7-10 This demonstrates how RMC and MCNP6.1 can mount the official multi-group database mgxsnp and the self-made multi-group database MACE. 56 Fe and 238 Neutron flux density and relative deviation in U-medium media. The calculation results show that the neutron flux density trends obtained by RMC and MCNP6.1 with different databases for different media are generally consistent, and the relative deviations obtained by RMC and MCNP with the same database are all within the range of 3σ. Therefore, the multi-group neutron transport calculation function of the RMC program and the self-made multi-group neutron database are considered reliable. Furthermore, for... 238 In the U-medium example, no results were found for neutron flux density measurements below 10⁻⁴ MeV on RMC and MCNP6.1 arrays equipped with the mgxsnp multigroup database, and the 3σ statistical error exceeded ±60% in some energy ranges. There are two main reasons for this deviation: 1) The evaluation nuclear database used by mgxsnp is sourced from ENDF / BV. Early experimental setups may have had limitations in measurement accuracy and coverage, resulting in no results below 10⁻⁴ MeV; 2) 238 In the low to medium energy range, U exhibits a strong resonant self-shell effect. The treatment of this effect has a significant impact on the neutron energy spectrum, leading to a larger statistical error.

[0093] Preferably, photon transport testing Figures 11-14This demonstrates how RMC and MCNP6.1 in pure photon mode can mount the official multi-group library mgxsnp and the self-made database MACE. nat Fe and nat The photon flux density and relative deviation of U-medium media were calculated, considering only the influence of natural nuclides since isotopic cross-sectional data are unavailable in pure photon mode. The results show that the photon flux density trends obtained from different databases using RMC and MCNP6.1 are generally consistent, and the relative deviations obtained from the same databases using RMC and MCNP are all within the 3σ range, indicating that RMC and the self-made multi-group database are reliable. Furthermore, the deviation between mgxsnp and MACE also stems from the evaluation kernel database version; mgxsnp uses ENDF / BV, while MACE uses ENDF / B-VII.1.

[0094] Preferably, upscattering testing: Compared to the multi-group ACE library mgxsnp provided by MCNP, the multi-group ACE library MACE developed in this project further considers the influence of upscattering effects. For the scattering matrix of 1H, additional upscattering data is provided. To test the impact of upscattering effects on neutron multi-group transport calculations, the geometric structure of the examples in this work is as follows... Figure 6 As shown, it is uniformly filled with a mass density of 1.0 g / cm³. 3 1H. Figure 15 and Figure 16 This demonstrates the neutron flux density and relative bias in 1H medium calculations using the RMC and MCNP6.1 programs before and after considering upscattering. The calculation results show that, after considering upscattering data, the overall neutron flux density in the low-energy region is lower, while the neutron flux density at the upscattering energy group increases sharply. Furthermore, the relative bias and statistical error are both within the 1% deviation range, consistent with actual physical laws. Therefore, it is concluded that the self-made multi-group library and the RMC program possess stable and reliable upscattering simulation capabilities.

[0095] Furthermore, SINBAD series OKTAVIAN-AI and Iron shielding examples: Figure 17 This paper primarily showcases the geometry of the SINBAD series OKTAVIAN-Al and Iron shielded examples, with specific material and radius information shown in Table 2. Furthermore, the OKTAVIAN-Al example employs a neutron-photon coupled transport calculation mode, while the OKTAVIAN-Iron example uses a pure neutron transport calculation mode with a particle number of 5,000,000. Calculations were performed using RMC and MCNP6.1 with the official multi-group library mgxsnp and the self-made multi-group library MACE. The outermost layer flow data were statistically analyzed, and the calculated results were compared with experimental values.

[0096] Table 2

[0097]

[0098] Figures 18-21 The results of neutron and photon surface flows in the OKTAVIAN-Al shielding example and the relative deviations between RMC and MCNP6.1 are presented. The calculation results show that the neutron and photon surface flows obtained by RMC and MCNP6.1 programs with the official multi-group library mgxsnp and the self-made multi-group library MACE are in good agreement with the experimental values, and the relative errors are all within the deviation range of 2% and 3σ, respectively, which meets the 5% error range required for engineering design.

[0099] Figure 22 and Figure 23 This section presents the neutron surface flux calculation results for the OKTAVIAN-Iron shielding example and the relative deviations between RMC and MCNP6.1. The calculation results also show that the neutron surface fluxes obtained by RMC and MCNP6.1 programs with the official multi-group library mgxsnp and the self-made multi-group library MACE agree well with experimental values, and the relative errors are all within the deviation range of 2% and 3σ, meeting the maximum error range of 5% for engineering design. However, in the low-energy and high-energy regions, the statistical errors of RMC and MCNP are relatively large. This deviation mainly comes from two aspects: 1) In the low-energy region, the resonance effect of iron has a significant impact on neutrons, leading to statistical error deviation; 2) In the high-energy region, the absorption cross section obtained from the ANISN database after TRANSX processing is reduced by (n, 2n), resulting in a negative absorption cross section in the high-energy region, thus increasing the statistical error.

[0100] The method for creating a neutron photon nucleus database for reactor multi-group Monte Carlo calculations according to embodiments of the present invention improves the adjoint calculation efficiency of the Monte Carlo program and further enhances the shielding calculation accuracy of the Monte Carlo program; it verifies the multi-group parameter processing function of the deterministic theory for multi-group calculations; and it inherits the advantages of the Monte Carlo program, possessing powerful geometric processing capabilities.

[0101] To achieve the above embodiments, such as Figure 24 As shown, this embodiment also provides a neutron photon nucleus database creation system 10 suitable for reactor multi-group Monte Carlo calculations. The system 10 includes: an initial data generation module 100, a first database generation module 200, a second database generation module 300, and a third database generation module 400.

[0102] The initial data generation module 100 is used to generate first-format data for a single nuclide using the evaluation nuclide database;

[0103] The first database generation module 200 is used to merge the first format data of multiple nuclides obtained based on the first format data of a single nuclide to obtain a database of first format data.

[0104] The second database generation module 300 is used to process the database of first format data to obtain a database of second format data related to the problem.

[0105] The third database generation module 400 is used to convert the database of second format data into a database of third format data using a preset format conversion program.

[0106] Furthermore, following the aforementioned third database generation module 400, a model testing module is also included, used for:

[0107] Test results were obtained by testing a database of third-format data based on the Monte Carlo program;

[0108] The accuracy test results of the database of third-format data are obtained based on the neutron and photon calculation results in the test results.

[0109] Furthermore, the aforementioned initial data generation module 100 is also used for:

[0110] The resonance cross section is obtained by transforming the resonance parameters in the evaluation kernel database;

[0111] The cross-section is processed by Doppler broadening, thermal processing of the scattering matrix and scattering cross-section of the nuclide, and narrow resonance approximation processing of the cross-section of the indistinguishable resonance region.

[0112] Based on the cross-section processing results, the continuous energy point cross-sections are grouped to obtain multi-group neutron data and photon cross-section data in multi-group form, so as to convert the multi-group neutron data and photon cross-section data into first format data.

[0113] Furthermore, the aforementioned second database generation module 300 is also used for:

[0114] Using the BONDARENKO baseline cross-section iteration method, the database of the first format data generated by the NJOY program, which is a MATXS format multi-group kernel database, is transformed into a second format data database, which is an ANISN format multi-group kernel database.

[0115] Furthermore, the aforementioned second database generation module 300 is also used for:

[0116] The corresponding multi-group cross sections are calculated based on the preset dilution cross section parameters in the database of the first format data.

[0117] Calculate the dilution cross section parameters related to the actual problem based on the nuclide nucleon density ratio of the actual problem, and interpolate to obtain the multi-group cross section parameters for the corresponding actual problem;

[0118] A database of second-format data is obtained by calculating based on multi-group cross-sectional parameters.

[0119] The neutron photon nucleus database creation system for reactor multi-group Monte Carlo calculations according to embodiments of the present invention improves the adjoint calculation efficiency of the Monte Carlo program and further enhances the shielding calculation accuracy of the Monte Carlo program; it verifies the multi-group parameter processing function of the deterministic theory for multi-group calculations; and it inherits the advantages of the Monte Carlo program, possessing powerful geometric processing capabilities.

[0120] Furthermore, the terms "first" and "second" are used for descriptive purposes only and should not be construed as indicating or implying relative importance or implicitly specifying the number of technical features indicated. Thus, a feature defined as "first" or "second" may explicitly or implicitly include at least one of that feature. In the description of this invention, "a plurality of" means at least two, such as two, three, etc., unless otherwise explicitly specified.

[0121] In the description of this specification, the references to terms such as "one embodiment," "some embodiments," "example," "specific example," or "some examples," etc., indicate that a specific feature, structure, material, or characteristic described in connection with that embodiment or example is included in at least one embodiment or example of the present invention. In this specification, the illustrative expressions of the above terms do not necessarily refer to the same embodiment or example. Furthermore, the specific features, structures, materials, or characteristics described may be combined in any suitable manner in one or more embodiments or examples. Moreover, without contradiction, those skilled in the art can combine and integrate the different embodiments or examples described in this specification, as well as the features of different embodiments or examples.

[0122] Although embodiments of the present invention have been shown and described above, it is understood that the above embodiments are exemplary and should not be construed as limiting the present invention. Those skilled in the art can make changes, modifications, substitutions and variations to the above embodiments within the scope of the present invention.

Claims

1. A method for constructing a neutron photon nucleus database suitable for multi-group Monte Carlo computation in reactors, characterized in that, Includes the following steps: First-format data for individual nuclides are generated using an evaluation nuclear database; A database of first-format data of multiple nuclides obtained based on the first-format data of the single nuclide is obtained by merging the first-format data of the first nuclide. Data processing is performed on the database of the first format data to obtain a database of the second format data related to the problem; A database of data in the second format is converted into a database of data in the third format using a preset format conversion program; The step of generating first-format data for a single nuclide using an evaluation nuclear database includes: converting the resonance parameters in the evaluation nuclear database to obtain a resonance cross section; performing Doppler broadening on the resonance cross section and thermalizing the scattering matrix and scattering cross section of the nuclide; and performing narrow resonance approximation on the cross section of the indistinguishable resonance region to obtain a cross section processing result. Based on the cross-section processing results, the continuous energy point cross-sections are grouped to obtain multi-group neutron data and photon cross-section data in multi-group form, so as to convert the multi-group neutron data and photon cross-section data into the first format data; The step of processing the database of the first format data to obtain a database of the second format data related to the problem includes: calculating the corresponding multi-group cross section based on the preset dilution cross section parameters in the database of the first format data; calculating the dilution cross section parameters related to the actual problem based on the nuclide nucleon density ratio of the actual problem, and interpolating to obtain the multi-group cross section parameters under the corresponding actual problem. A database of the second format data is obtained by calculating based on the multi-group cross-sectional parameters.

2. The method according to claim 1, characterized in that, After obtaining the database of data in the third format, the method further includes: The test results were obtained by testing the database of the third-format data based on the Monte Carlo program. The accuracy test results of the database of the third-format data are obtained based on the neutron and photon calculation results in the test results.

3. The method according to claim 1, characterized in that, The process of processing the database of the first format data to obtain a database of the second format data related to the problem includes: Using the BONDARENKO baseline cross-section iteration method, the database of the first format data generated by the NJOY program, which is a MATXS format multi-group kernel database, is transformed into a second format data database, which is an ANISN format multi-group kernel database.

4. A system for creating a neutron photon nucleus database suitable for multi-group Monte Carlo simulations in reactors, characterized in that, include: The initial data generation module is used to generate first-format data for a single nuclide using the evaluation nucleus database; The first database generation module is used to merge the first format data of multiple nuclides obtained based on the first format data of the single nuclide to obtain a database of first format data. The second database generation module is used to process the database of the first format data to obtain a database of the second format data related to the problem. The third database generation module is used to convert the database of the second format data into a database of the third format data using a preset format conversion program; The initial data generation module is further configured to: convert the resonance parameters in the evaluation kernel database to obtain the resonance cross section; perform Doppler broadening on the resonance cross section and thermal processing on the scattering matrix and scattering cross section of the nuclide; and perform narrow resonance approximation processing on the cross section of the indistinguishable resonance region to obtain the cross section processing result. Based on the cross-section processing results, the continuous energy point cross-sections are grouped to obtain multi-group neutron data and photon cross-section data in multi-group form, so as to convert the multi-group neutron data and photon cross-section data into the first format data; The second database generation module is further configured to: calculate the corresponding multi-group cross section based on the preset dilution cross section parameters in the database of the first format data; calculate the dilution cross section parameters related to the actual problem based on the nuclide nucleon density ratio of the actual problem, and interpolate to obtain the multi-group cross section parameters under the corresponding actual problem; A database of the second format data is obtained by calculating based on the multi-group cross-sectional parameters.

5. The system according to claim 4, characterized in that, Following the third database generation module, a model testing module is also included, used for: The test results were obtained by testing the database of the third-format data based on the Monte Carlo program. The accuracy test results of the database of the third-format data are obtained based on the neutron and photon calculation results in the test results.

6. The system according to claim 4, characterized in that, The second database generation module is also used for: Using the BONDARENKO baseline cross-section iteration method, the database of the first format data generated by the NJOY program, which is a MATXS format multi-group kernel database, is transformed into a second format data database, which is an ANISN format multi-group kernel database.

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