An efficient method for obtaining neutron dose field of nuclear explosion based on MC concomitant transport

Abstract

The present invention relates to a numerical simulation calculation method for a nuclear explosion radiation environment, and specifically to a method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport. The method addresses the technical problem of efficiently simulating and calculating the neutron dose field of a nuclear explosion under conditions of varying blast heights and neutron source parameters. The method iteratively calculates a neutron co-transport model to obtain a neutron source's importance value for the neutron absorbed dose; iteratively calculates a secondary gamma co-transport model to obtain a neutron source's importance value for the secondary gamma absorbed dose; and based on these values, calculates the neutron absorbed dose value and the secondary gamma absorbed dose value of the nuclear explosion neutron radiation source, thereby completing the acquisition of the nuclear explosion neutron dose field. This method achieves efficient calculation of the nuclear explosion neutron dose field under any given blast height and source parameter conditions.

Description

Technical Field

[0001] The present invention relates to a numerical simulation calculation method for a nuclear explosion radiation environment, and in particular to a method for efficiently acquiring a nuclear explosion neutron dose field based on MC accompanying transport. Background Art

[0002] Neutron dose, a crucial component of early nuclear radiation, refers to the neutron and secondary gamma absorption dose generated by prompt neutrons from a nuclear explosion at different projected distances from the detonation center. Calculating and predicting the ground neutron dose field distribution under different burst heights and source parameters is crucial for assessing the human effects of early nuclear radiation exposure. The neutron dose field is typically calculated using the Monte Carlo method (MC) using a forward transport approach. Neutrons are extracted from the detonation center, transported through interactions with the atmosphere and the ground, and then statistically recorded using a series of detectors at different projected distances. This yields a dose field distribution. However, this method can only generate ground neutron or secondary gamma absorption dose distributions for one burst height and one neutron source in a single simulation. For neutron dose fields at different burst heights and source parameters, the calculation model must be rebuilt and multiple MC forward transport simulations performed, resulting in a complex and inefficient process.

[0003] MC co-transport is the reverse process of forward transport. In this co-transport method, co-transport neutrons extracted from the detection location are first "assigned" a certain weight. After interaction and transport with the atmosphere and the ground, the co-transport neutron flux is statistically recorded at source locations at different blast heights and relative distances. The most significant difference from forward transport is that the interaction process occurs in reverse and the energy of the co-transport neutrons gradually increases. The co-transport fluence given by co-transport has the physical meaning of reflecting the neutron importance parameter at the forward source location, that is, the contribution of a single source neutron to the detector response. Therefore, solving a series of "single source-multiple detector" forward transport problems can be transformed into solving a single "single detector-multiple source" co-transport problem. Through a single adjoint calculation, the importance parameter of each energy group radiation source to the detector response at different locations throughout space is obtained. For nuclear explosion neutron sources with arbitrary blast height and energy spectrum distribution, there is no need to perform multiple simulation calculations. It is only necessary to re-group the energy spectrum. By combining the neutron and secondary gamma absorbed dose importance value parameters of neutron sources in each energy group at different spatial positions obtained by simulation calculation, the distribution of corresponding ground absorbed dose with the projection distance of the explosion center can be obtained.

[0004] Among the existing related technologies, there has not been any report on the use of MC accompanying transport simulation calculations to efficiently obtain the neutron dose field of nuclear explosions with different blast heights and different neutron source parameters. If the existing MC forward transport simulation method is used, a single simulation can only obtain the neutron dose field distribution of nuclear explosions under a single blast height and one neutron source parameter, and the calculation efficiency is low. Summary of the Invention

[0005] The purpose of the present invention is to solve the technical problem of efficient simulation calculation of the neutron dose field of a nuclear explosion under different blast heights and different neutron source parameters, and to provide a method for efficiently obtaining the neutron dose field of a nuclear explosion based on MC accompanying transport, so as to realize efficient calculation of the neutron dose field of a nuclear explosion under any given blast height and source parameter conditions.

[0006] In order to solve the above technical problems, the technical solution adopted by the present invention is:

[0007] A method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport is special in that it includes the following steps:

[0008] S1. Establish a neutron transport model;

[0009] S2. Setting neutron co-transport model parameters; the neutron co-transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source neutron weight;

[0010] S3. Iteratively calculate the neutron concomitant transport model to obtain the concomitant fluence in each neutron energy group and the importance value of the neutron source of each neutron energy group in the forward model corresponding to the concomitant fluence to the neutron absorbed dose;

[0011] S4. Establish a secondary γ-emission transport model;

[0012] S5. Setting the secondary γ-adjoint transport model parameters; the secondary γ-adjoint transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source γ weight;

[0013] S6. Iteratively calculate the secondary gamma ray transport model to obtain the importance value of the neutron source of each neutron energy group to the secondary gamma ray absorbed dose;

[0014] S7. Based on the importance value of the neutron source to the neutron absorption dose obtained in S3 and the importance value of the neutron source to the secondary gamma absorption dose obtained in S6, the neutron absorption dose value and the secondary gamma absorption dose value of the nuclear explosion neutron radiation source are calculated to complete the acquisition of the nuclear explosion neutron dose field.

[0015] Furthermore, S1 is specifically:

[0016] 1.1) Set up a neutron point source as the companion neutron source;

[0017] 1.2) Multiple detectors are set up at different blast heights and projection distances from the neutron source to serve as neutron counting units;

[0018] 1.3) According to steps 1.1) and 1.2), the Monte Carlo geometric parameters of the neutron source's transport in the atmosphere are set to establish a neutron transport model.

[0019] Furthermore, S2 is specifically:

[0020] 2.1) The neutron transport model calculation mode is set to multi-group adjoint, and the multi-group neutron structure is subdivided using the logarithmic interpolation method to obtain neutron subdivided energy groups. At the same time, the average fluence-dose conversion coefficient of each neutron subdivided energy group is calculated;

[0021] 2.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the neutron transport model based on the neutron subdivision energy group and fluence-dose average conversion coefficient obtained in step 2.1);

[0022] 2.3) Set the source neutron weight wgt1 in the neutron transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 2.2).

[0023] Furthermore, step 2.1) is specifically as follows:

[0024] 2.1.1. Establish a point source-spherical MC calculation model;

[0025] 2.1.2. Set the parameters of the point source-sphere MC calculation model;

[0026] The source neutrons in the point source-sphere MC calculation model are sampled according to neutron subdivision energy groups, and the sampling probability of each neutron subdivision energy group is set to 1. The total number of neutron subdivision energy groups is multiplied by the area of ​​the sphere to serve as the weight of the source neutrons in the model. A sphere with a radius of 1 cm is set to record the fluence of each neutron subdivision energy group. A fluence-dose conversion factor with a point-valued response function is set to convert the recorded fluence into dose.

[0027] 2.1.3. Based on the parameters of the point source-spherical MC calculation model in step 2.1.2, obtain the counting results of each neutron energy group and use them as the corresponding average fluence-dose conversion coefficient;

[0028] Step 2.2) is specifically as follows:

[0029] 2.2.1. Set the neutron subdivision energy group to the neutron sampling energy group of the neutron transport model;

[0030] 2.2.2. Set the average fluence-dose conversion coefficient of each neutron energy group obtained in step 2.1.3 as the neutron sampling probability of the neutron transport model;

[0031] 2.2.3. Set the neutron subdivision energy group to the counting energy group of the neutron transport model;

[0032] 2.2.4. Set all counting multipliers to 1;

[0033] The source neutron weight wgt1 in step 2.3) is specifically:

[0034] According to steps 2.2.1 to 2.2.4, the source neutron weight wgt1 can be obtained

[0035]

[0036] Where n is the total number of neutron sampling energy groups, R g is the counting energy group of the neutron transport model, and g is the neutron sampling probability of the neutron transport model.

[0037] Furthermore, S3 is specifically:

[0038] 3.1) Divide the neutron transport model into grids and record the neutron fluence in each grid;

[0039] 3.2) Iterate the calculation formula using the neutron accompanying fluence and grid weight window minus variance parameter to generate the weight window lower limit parameter for each grid, and use the weight window lower limit parameter in step 2.3) Substitute into the grid weight window variance reduction parameter calculation formula:

[0040]

[0041] Where w th,i is the lower limit parameter of the weight window of the i-th grid;

[0042] λ is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window;

[0043] is the neutron companion fluence in the i-th grid;

[0044] is the maximum value among all neutron concomitant fluences.

[0045] 3.3) According to the grid weight window lower limit parameter w in step 3.2) th,i The grid weight window variance reduction method is used to solve the neutron accompanying transport model, and the neutron accompanying fluence at different explosion heights and explosion center projection distances from the accompanying neutron source is obtained, which is the importance value of the neutron source of each neutron energy group to the neutron absorption dose in the corresponding forward model.

[0046] Furthermore, S4 is specifically:

[0047] 4.1) Set up a point source as an accompanying secondary gamma source;

[0048] 4.2) Multiple detectors are set up at different explosion heights and projection distances from the secondary gamma ray source as neutron counting units;

[0049] 4.3) According to steps 4.1) and 4.2), the Monte Carlo geometric parameters of the secondary gamma source transport in the atmosphere are set to establish a secondary gamma source transport model.

[0050] Furthermore, S5 is specifically:

[0051] 5.1) Set the secondary gamma ray transport model calculation mode to multi-group adjoint, and use the logarithmic interpolation method to subdivide the multi-group secondary gamma ray structure to obtain secondary gamma subdivision energy groups. At the same time, calculate the average fluence-dose conversion coefficient for each secondary gamma subdivision energy group;

[0052] 5.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the secondary gamma ray transport model based on the secondary gamma ray subdivision energy group and the fluence-dose average conversion coefficient obtained in step 5.1);

[0053] 5.3) Set the source secondary γ weight wgt in the secondary γ adjoint transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 5.2).

[0054] Furthermore, step 5.1) is specifically as follows:

[0055] 5.1.1. Establish a point source-spherical MC calculation model;

[0056] 5.1.2. Set the parameters of the point source-sphere MC calculation model;

[0057] The source gamma of the point source-sphere MC calculation model is sampled according to the secondary gamma subdivision energy group, and the sampling probability of each secondary gamma subdivision energy group is set to 1. The total number of secondary gamma subdivision energy groups is multiplied by the area of ​​the sphere as the weight of the source gamma in the model. A sphere with a radius of 1 cm is set to record the fluence of each secondary gamma subdivision energy group. A fluence-dose conversion factor with a point-valued response function is set to convert the recorded fluence into dose.

[0058] 5.1.3. Based on the parameters of the point source-spherical MC calculation model in step 5.1.2, obtain the counting results of each secondary gamma subdivision energy group and use them as the corresponding average fluence-dose conversion coefficient;

[0059] Step 5.2) is specifically as follows:

[0060] 5.2.1. Set the secondary γ subdivision energy group to the secondary γ sampling energy group of the secondary γ adjoint transport model;

[0061] 5.2.2. Set the average fluence-dose conversion coefficient of each secondary gamma energy group obtained in step 5.1.3 as the secondary gamma sampling probability of the secondary gamma adjoint transport model;

[0062] 5.2.3. Set the neutron subdivision energy group to the neutron counting energy group of the secondary γ-linked transport model;

[0063] 5.2.4. Set all counting multipliers to 1;

[0064] The source secondary γ weight wgt2 in step 5.3) is specifically:

[0065] According to steps 5.2.1 to 5.2.4, the source neutron weight wgt2 can be obtained.

[0066]

[0067] Where n is the total number of neutron subdivision energy groups, m is the total number of secondary γ subdivision energy groups, R s is the neutron counting energy group of the secondary γ transport model, and s is the secondary γ sampling probability of the secondary γ transport model.

[0068] Furthermore, S6 is specifically:

[0069] 6.1) Divide the secondary γ-ray transport model into grids and record the neutron fluence within each grid;

[0070] 6.2) Perform iterative calculations using the neutron accompanying fluence and grid weight window minus variance parameter calculation formula to generate the weight window lower limit parameter for each grid, and use the weight window lower limit parameter in step 5.3) as the weight window lower limit parameter. Substitute into the grid weight window variance reduction parameter calculation formula:

[0071]

[0072] Where w th,j is the lower limit parameter of the weight window of the j-th grid;

[0073] λ' is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window;

[0074] is the secondary γ adjoint fluence in the jth grid;

[0075] is the maximum value among all secondary γ adjoint fluences.

[0076] 6.3) According to the grid weight window lower limit parameter w in step 6.2) th,jThe accompanying transport model of secondary γ is solved by using the grid weight window variance reduction method to obtain the accompanying neutron flux at different explosion heights and explosion center projection distances from the accompanying secondary γ source, which is the importance value of the neutron source of each neutron energy group in the corresponding forward model to the secondary γ absorbed dose.

[0077] Furthermore, S7 is specifically:

[0078] 7.1) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup on the neutron absorbed dose obtained from S3, and then add them together to obtain the absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions;

[0079] 7.2) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup to the secondary gamma absorbed dose obtained from S6, and then add them together to obtain the secondary gamma absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions.

[0080] Compared with the prior art, the beneficial effects of the technical solution of the present invention are:

[0081] 1. The present invention provides a method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport. The method converts the solution of a series of "single source-multiple detector" calculation models (Monte Carlo models of co-transport in the atmosphere) into the solution of a "single detector-multiple source" co-transport model (neutron co-transport model and secondary γ co-transport model). Through co-transport simulation calculation, the importance value parameters of neutron sources of various energy groups at different locations to the neutron and secondary γ absorbed doses are obtained. This method can be used as a universal data to quickly acquire the radiation dose field of neutrons from a nuclear explosion with any source energy spectrum.

[0082] 2. The present invention provides a method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport. The neutron dose field of a nuclear explosion under different explosion heights and source parameters can be obtained without repeated simulation calculations. One simulation can efficiently obtain the neutron or secondary γ absorption dose distribution under different explosion conditions. BRIEF DESCRIPTION OF THE DRAWINGS

[0083] Figure 1 This is a flow chart of an embodiment of the method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport of the present invention.

[0084] Figure 2 Schematic diagram of the concomitant transport model of the neutron dose field of a nuclear explosion under open conditions in an embodiment of the present invention, where 1#, 2#...n represent the number of detectors.

[0085] Figure 2 The accompanying drawings are marked as follows:

[0086] 1-Ground, 2-Nuclear explosion radiation source, 3-Detector.

[0087] Figure 3 Schematic diagram comparing the accompanying transport results of neutron absorption dose near the ground and the forward transport results corresponding to the neutron accompanying transport model under different blast heights for the same radiation source in an embodiment of the present invention.

[0088] Figure 4 Schematic diagram comparing the secondary gamma absorbed dose accompanying transport results near the ground and the forward transport results corresponding to the secondary gamma accompanying transport model under different blast heights for the same radiation source in an embodiment of the present invention.

[0089] Figure 5 Schematic diagram comparing the accompanying transport results of neutron absorption dose near the ground and the forward transport results corresponding to the neutron accompanying transport model under the same blast height and different radiation sources in an embodiment of the present invention.

[0090] Figure 6 Schematic diagram comparing the secondary gamma absorbed dose accompanying transport results near the ground and the forward transport results corresponding to the secondary gamma accompanying transport model under the same blast height and different radiation sources in an embodiment of the present invention. DETAILED DESCRIPTION

[0091] The technical solution of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the embodiments described are only some embodiments of the present invention, not all embodiments. All other embodiments obtained by ordinary technicians in this field based on the technical solution of the present invention without making any creative efforts shall fall within the scope of protection of the present invention.

[0092] A method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport is special in that it includes the following steps:

[0093] S1. Establish a neutron transport model;

[0094] 1.1) Set up a neutron point source as the companion neutron source;

[0095] 1.2) Multiple detectors 3 are set up at different explosion heights and projection distances from the explosion center from the accompanying neutron source, serving as neutron counting units;

[0096] 1.3) According to steps 1.1) and 1.2), the Monte Carlo geometric parameters of the neutron source's transport in the atmosphere are set to establish a neutron transport model.

[0097] S2. Setting neutron co-transport model parameters; the neutron co-transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source neutron weight;

[0098] 2.1) The neutron transport model calculation mode is set to multi-group adjoint, and the multi-group neutron structure is subdivided using the logarithmic interpolation method to obtain neutron subdivided energy groups. At the same time, the average fluence-dose conversion coefficient of each neutron subdivided energy group is calculated;

[0099] 2.1.1. Establish a point source-spherical MC calculation model;

[0100] 2.1.2. Set the parameters of the point source-sphere MC calculation model;

[0101] The source neutrons in the point source-sphere MC calculation model are sampled according to neutron subdivision energy groups, and the sampling probability of each neutron subdivision energy group is set to 1. The total number of neutron subdivision energy groups is multiplied by the area of ​​the sphere to serve as the weight of the source neutrons in the model. A sphere with a radius of 1 cm is set to record the fluence of each neutron subdivision energy group. A fluence-dose conversion factor with a point-valued response function is set to convert the recorded fluence into dose.

[0102] 2.1.3. Based on the parameters of the point source-spherical MC calculation model in step 2.1.2, obtain the counting results of each neutron energy group and use them as the corresponding average fluence-dose conversion coefficient;

[0103] 2.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the neutron transport model based on the neutron subdivision energy group and fluence-dose average conversion coefficient obtained in step 2.1);

[0104] 2.2.1. Set the neutron subdivision energy group to the neutron sampling energy group of the neutron transport model;

[0105] 2.2.2. Set the average fluence-dose conversion coefficient of each neutron energy group obtained in step 2.1.3 as the neutron sampling probability of the neutron transport model;

[0106] 2.2.3. Set the neutron subdivision energy group to the counting energy group of the neutron transport model;

[0107] 2.2.4. Set all counting multipliers to 1;

[0108] 2.3) Set the source neutron weight wgt1 in the neutron transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 2.2).

[0109] According to steps 2.2.1 to 2.2.4, the source neutron weight wgt1 can be obtained

[0110]

[0111] Where n is the total number of neutron sampling energy groups, R gis the counting energy group of the neutron transport model, and g is the neutron sampling probability of the neutron transport model.

[0112] S3. Iteratively calculate the neutron concomitant transport model to obtain the concomitant fluence in each neutron energy subgroup, that is, the importance value of the neutron source of each neutron energy subgroup to the neutron absorbed dose in the corresponding forward model;

[0113] 3.1) Divide the neutron transport model into grids and record the neutron fluence in each grid;

[0114] 3.2) Iterate the calculation formula of neutron accompanying fluence and grid weight window minus variance parameter to generate the weight window lower limit parameter of each grid (each grid has a corresponding weight window) and replace the weight window lower limit parameter in step 2.3) with the weight window lower limit parameter of each grid (each grid has a corresponding weight window). Substitute into the grid weight window variance reduction parameter calculation formula:

[0115]

[0116] Where w th,i is the lower limit parameter of the weight window of the i-th grid;

[0117] λ is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window;

[0118] is the neutron companion fluence in the i-th grid;

[0119] is the maximum value among all neutron concomitant fluences.

[0120] 3.3) According to the grid weight window lower limit parameter w in step 3.2) th,i The grid weight window variance reduction method is used to solve the neutron accompanying transport model, and the neutron accompanying fluence at different explosion heights and explosion center projection distances from the accompanying neutron source is obtained, which is the importance value of the neutron source of each neutron energy group to the neutron absorption dose in the corresponding forward model.

[0121] S4. Establish a secondary γ-emission transport model;

[0122] 4.1) Set up a point source as an accompanying secondary gamma source;

[0123] 4.2) Multiple detectors 3 are set up at different explosion heights and projection distances from the secondary gamma source to serve as neutron counting units;

[0124] 4.3) According to steps 4.1) and 4.2), the Monte Carlo geometric parameters of the secondary gamma source transport in the atmosphere are set to establish a secondary gamma source transport model.

[0125] S5. Setting the secondary γ-adjoint transport model parameters; the secondary γ-adjoint transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source γ weight;

[0126] 5.1) Set the secondary gamma ray transport model calculation mode to multi-group adjoint, and use the logarithmic interpolation method to subdivide the multi-group secondary gamma ray structure to obtain secondary gamma subdivision energy groups. At the same time, calculate the average fluence-dose conversion coefficient for each secondary gamma subdivision energy group;

[0127] 5.1.1. Establish a point source-spherical MC calculation model;

[0128] 5.1.2. Set the parameters of the point source-sphere MC calculation model;

[0129] The source gamma of the point source-sphere MC calculation model is sampled according to the secondary gamma subdivision energy group, and the sampling probability of each secondary gamma subdivision energy group is set to 1. The total number of secondary gamma subdivision energy groups is multiplied by the area of ​​the sphere as the weight of the source gamma in the model. A sphere with a radius of 1 cm is set to record the fluence of each secondary gamma subdivision energy group. A fluence-dose conversion factor with a point-valued response function is set to convert the recorded fluence into dose.

[0130] 5.1.3. Based on the parameters of the point source-spherical MC calculation model in step 5.1.2, obtain the counting results of each secondary gamma subdivision energy group and use them as the corresponding average fluence-dose conversion coefficient;

[0131] 5.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the secondary gamma ray transport model based on the secondary gamma ray subdivision energy group and the fluence-dose average conversion coefficient obtained in step 5.1);

[0132] 5.2.1. Set the secondary γ subdivision energy group to the secondary γ sampling energy group of the secondary γ adjoint transport model;

[0133] 5.2.2. Set the average fluence-dose conversion coefficient of each secondary gamma energy group obtained in step 5.1.3 as the secondary gamma sampling probability of the secondary gamma adjoint transport model;

[0134] 5.2.3. Set the neutron subdivision energy group to the neutron counting energy group of the secondary γ-linked transport model;

[0135] 5.2.4. Set all counting multipliers to 1;

[0136] 5.3) Set the source secondary γ weight wgt2 in the secondary γ adjoint transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 5.2).

[0137] According to steps 5.2.1 to 5.2.4, the source neutron weight wgt2 can be obtained.

[0138]

[0139] Where n is the total number of neutron subdivision energy groups, m is the total number of secondary γ subdivision energy groups, R s is the neutron counting energy group of the secondary γ transport model, and s is the secondary γ sampling probability of the secondary γ transport model.

[0140] S6. Iteratively calculate the secondary gamma ray transport model to obtain the importance value of the neutron source of each neutron energy group to the secondary gamma ray absorbed dose;

[0141] 6.1) Divide the secondary γ-ray transport model into grids and record the neutron fluence within each grid;

[0142] 6.2) Perform iterative calculations using the neutron accompanying fluence and grid weight window minus variance parameter calculation formula to generate the weight window lower limit parameter for each grid, and use the weight window lower limit parameter in step 5.3) as the weight window lower limit parameter. Substitute into the grid weight window variance reduction parameter calculation formula:

[0143]

[0144] Where w th,j is the lower limit parameter of the weight window of the j-th grid;

[0145] λ' is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window;

[0146] is the secondary γ adjoint fluence in the jth grid;

[0147] is the maximum value among all secondary γ adjoint fluences.

[0148] 6.3) According to the grid weight window lower limit parameter w in step 6.2) th,j The accompanying transport model of secondary γ is solved by using the grid weight window variance reduction method to obtain the accompanying neutron flux at different explosion heights and explosion center projection distances from the accompanying secondary γ source, which is the importance value of the neutron source of each neutron energy group in the corresponding forward model to the secondary γ absorbed dose.

[0149] S7. Based on the importance value of the neutron source to the neutron absorption dose obtained in S3 and the importance value of the neutron source to the secondary gamma absorption dose obtained in S6, the neutron absorption dose value and the secondary gamma absorption dose value of the nuclear explosion neutron radiation source are calculated to complete the acquisition of the nuclear explosion neutron dose field.

[0150] 7.1) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup on the neutron absorbed dose obtained from S3, and then add them together to obtain the absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions;

[0151] 7.2) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup to the secondary gamma absorbed dose obtained from S6, and then add them together to obtain the secondary gamma absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions.

[0152] The following is a specific embodiment of the present invention:

[0153] like Figure 1 As shown, the present invention provides a flow chart of a method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport, which includes the following steps:

[0154] Step 1: If Figure 2 As shown in the figure, a Monte Carlo geometric model of the transport of neutrons from a nuclear explosion in the atmosphere under open conditions is established; the specific approach is as follows:

[0155] The nuclear explosion radiation source 2 (companion source) is an isotropic point source set at an explosion height of 1 m above the ground 1. Within the explosion height range of 1 m to 2000 m and the explosion center projection distance range of 25 m to 4000 m, a series of counting units (detectors 3) with a radius of 1 m are set as neutron fluence counting units for the companion calculation to obtain the neutron companion transport model.

[0156] Among them, the ground 1 medium is concrete (density 2.31g / cm 3 ), the air medium is the standard atmospheric composition (pressure is 101.60kPa, humidity is 50.0%, density is 1.225mg / cm 3 );

[0157] Step 2: Set the parameters of the neutron transport model

[0158] Logarithmic interpolation is used to subdivide the neutron energy group structures of groups 1-30 shown in Table 1 (each neutron energy group structure is subdivided into 50 groups). After subdivision, new neutron subdivided energy groups are obtained, and the total number of energy bins is 1500 groups. The average fluence-dose conversion coefficient corresponding to the neutron subdivided energy groups is calculated; the neutron subdivided energy group is set as the neutron sampling energy group of the companion source, and the average fluence-dose conversion coefficient is set as the neutron sampling probability.

[0159] The total number of counting energy bins is 1500, and their energy multipliers are all set to 1. The source neutron weight in the neutron concomitant transport model is calculated to be 1.959E-09. The neutron fluence-dose conversion factor is shown in Table 2.

[0160] Table 1 Energy group structure of multi-group neutrons (neutrons: 30 groups)

[0161]

[0162]

[0163] Table 2 Neutron Fluence-Dose Conversion Factors

[0164]

[0165]

[0166] Step 3: Use cube grid to mesh the neutron transport model. Considering the mean free path of neutrons in air, set the cube grid size to 47×47×28. The calculation area is x: -4600~4600m, y: -4600~4600m, z: -2.2~5000m. Use cube grid to construct weight window parameters and corresponding grid counts. The parameter ratio λ of the upper limit of each grid weight window to the lower limit of the grid weight window is set to 5.

[0167] In this embodiment, the cubic meshing covers the entire solution area (i.e., the blast height range of 1m to 2000m and the projection distance from the detonation epicenter range of 25m to 4000m). Therefore, when calculating the lower limit parameters of the weight window for each grid, transport over long distances and within thick shielding will result in very small lower limit parameters for the weight windows in the peripheral regions and ground 1 of the neutron transport model. To avoid excessive splitting of simulated particles in the peripheral regions and ground 1 during the iterative calculation, the lower limit parameters of the weight windows for the grids in these regions are set to 0. Simulated particles do not split in these regions, and normal transport occurs.

[0168] The neutron accompanying transport model is iteratively calculated, requiring that the statistical error of the accompanying fluence of each neutron fluence counting unit is less than 10% (Monte Carlo programs have requirements for the statistical error of the counting results, and for body detector 3, the requirement is less than 10%). After two iterations, the neutron accompanying fluence at different explosion heights and explosion center projection distances from the point source is obtained, that is, the importance value of each subdivided energy group neutron source in the corresponding forward model to the secondary gamma absorption dose.

[0169] Step 4: If Figure 2 As shown in FIG, a Monte Carlo geometric model of the transport of secondary gamma rays from nuclear explosions in the atmosphere under open conditions is established. Under different explosion heights of the same radiation source, neutron detection is performed by detectors 1#, 2#...n 3. The specific method is as follows:

[0170] The companion source is set as an isotropic point source at the explosion height of 11 m above the ground. Within the explosion height range of 1 m to 2000 m and the explosion center projection distance range of 25 m to 4000 m, a series of counting units with a radius of 1 m are set as neutron fluence counting units for the companion calculation to obtain the secondary γ companion transport model.

[0171] Among them, the ground 1 medium is concrete (density 2.31g / cm 3 ), the air medium is the standard atmospheric composition (pressure is 101.60kPa, humidity is 50.0%, density is 1.225mg / cm 3 );

[0172] Step 5: Use logarithmic interpolation to subdivide the 1-12 photon energy group structure shown in Table 3 (each photon energy group is subdivided into 100 groups). After subdivision, the number of new photon subdivision energy groups obtained is 1200 groups. Calculate the average fluence-dose conversion coefficient corresponding to the photon subdivision energy group, set the photon subdivision energy group as the secondary γ sampling energy group of the companion source, and set the average fluence-dose conversion coefficient corresponding to the photon subdivision energy group as the secondary γ sampling probability. The counting energy bins of the photon subdivision energy group are also set according to the photon subdivision energy group structure, and their energy multipliers are all set to 1;

[0173] The source γ weight in the secondary γ-emission concomitant transport model is calculated to be 1.192E-09; the secondary γ-emission-dose conversion factor is shown in Table 4;

[0174] Table 3 Energy group structure of multi-group photons (photons: 12)

[0175]

[0176]

[0177] Table 4 Secondary γ-irradiation-dose conversion factors

[0178]

[0179] Step 6: Use cube grid to mesh the secondary γ transport model. Considering the mean free path of secondary γ in the air, set the cube grid size to 47×47×28, and the calculation area to x: -4600~4600m, y: -4600~4600m, z: -2.2~5000m. Use cube grid to construct weight window parameters and corresponding grid counts. The parameter ratio λ of the upper limit of each grid weight window to the lower limit of the grid weight window is set to 5.

[0180] In this example, the cubic meshing covers the entire solution area (i.e., the blast height range of 1m to 2000m and the projection distance of the detonation epicenter range of 25m to 4000m). Therefore, when calculating the lower limit parameters of the weight window for each grid, transport over long distances and within thick shielding bodies will result in very small lower limit parameters for the weight windows in the peripheral regions and ground 1 of the secondary gamma-ray transport model. To avoid excessive splitting of simulated particles in the peripheral regions and ground 1 during the iterative calculation, the lower limit parameters of the mesh weight windows in these regions are set to 0. This prevents simulated particles from splitting within these regions, allowing normal transport.

[0181] The secondary γ-ray accompanying transport model is iteratively calculated, requiring that the statistical error of the accompanying fluence of each secondary γ-ray fluence counting unit is less than 10%; after two iterations, the neutron accompanying fluence at different explosion heights and explosion center projection distances from the point source is obtained, that is, the importance value of the neutron source of each subdivided energy group in the corresponding forward model to the secondary γ absorbed dose.

[0182] Step 7: Calculations were performed for the same radiation source with different burst heights and for the same burst height with different source parameters. The specific process included:

[0183] 7.1. Multiply the energy spectrum of radiation source 1 and the neutron absorption dose importance values ​​of neutron sources of each energy group at different blast heights obtained in step 3 in the corresponding energy group and add them together to obtain the accompanying transport results of the neutron absorption dose field at different blast heights for the same radiation source, as follows: Figure 3 shown.

[0184] The energy spectrum of radiation source 1 and the importance values ​​of secondary gamma absorbed dose of neutron sources of each energy group under different blast heights obtained in step 6 are multiplied in the corresponding energy group and then added together to obtain the accompanying transport results of the secondary gamma absorbed dose field of the same radiation source under different blast heights, as shown in the following example: Figure 4 shown.

[0185] 7.2. Multiply the new energy spectra of radiation sources 1, 2, 3, and 4 with the neutron absorption dose importance values ​​of neutron sources in each energy group at a blast height of 400 m obtained in step 4, and then add them together to obtain the accompanying transport results of the neutron absorption dose field for different radiation sources at the same blast height, as shown in the following example: Figure 5 shown.

[0186] The new energy spectra of radiation sources 1, 2, 3, and 4 are multiplied by the importance values ​​of secondary γ absorbed doses of the secondary γ sources of each energy group at the 400 m blast height obtained in step 4, and then added together to obtain the accompanying transport results of the secondary γ absorbed dose field under different radiation sources at the same blast height, as shown in the following example: Figure 6 shown.

[0187] like Figures 3 to 6As shown in Figure 5, the transport results of the corresponding forward model of the neutron accompanying transport model or the secondary γ accompanying transport model are given. It can be seen from the figure that the accompanying transport results of the neutron accompanying transport model or the secondary γ accompanying transport model are in good agreement with the forward transport results. Table 5 gives the average relative deviation between the two.

[0188] Table 5 Average relative deviations between adjoint and forward transport results

[0189]

[0190] As can be seen from Table 5, the neutron accompanying transport model or the secondary γ accompanying transport model accompanying transport results are in good agreement with the forward transport results. The average relative deviation between the two is less than 15%, and the degree of agreement is high, indicating that the above technical solution has significant reference value for engineering applications.

Claims

1. A method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport, characterized in that: The following steps are involved: S1. Establish a neutron transport model; S2. Setting neutron transport model parameters; the neutron transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source neutron weight; the source neutron weight wgt1: Where n is the total number of neutron sampling energy groups, R g is the counting energy group of the neutron transport model, g is the neutron sampling probability of the neutron transport model; S3. Perform iterative calculation on the neutron concomitant transport model to obtain the concomitant fluence in each neutron energy subgroup, and the importance value of the neutron source of each neutron energy subgroup in the forward model to the neutron absorbed dose corresponding to the concomitant fluence: 3.1) Divide the neutron transport model into grids and record the neutron fluence in each grid; 3.2) Using the neutron accompanying fluence and grid weight window minus variance parameter calculation formula, iterative calculation is performed to generate the weight window lower limit parameter of each grid, and the weight window in S2 is Substitute into the grid weight window variance reduction parameter calculation formula: Where w th,i is the lower limit parameter of the weight window of the i-th grid; λ is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window; is the neutron companion fluence in the i-th grid; is the maximum value among all neutron accompanying fluences; 3.3) According to the grid weight window lower limit parameter w in step 3.2) th,i , the grid weighted window variance reduction method is used to solve the neutron concomitant transport model, and the neutron concomitant flux at different explosion heights and explosion center projection distances from the concomitant neutron source is obtained, which is the importance value of the neutron source of each neutron energy group to the neutron absorbed dose in the corresponding forward model; S4. Establish a secondary γ-emission transport model; S5. Setting secondary γ-adjoint transport model parameters; the secondary γ-adjoint transport model parameters include sampling energy group, sampling probability, counting energy group, counting multiplier and source secondary γ weight; The source secondary gamma weights wgt2: Where n is the total number of neutron subdivision energy groups, m is the total number of secondary γ subdivision energy groups, R s is the neutron counting energy group of the secondary γ-adjoint transport model, s is the secondary γ sampling probability of the secondary γ-adjoint transport model; S6. Perform iterative calculations on the secondary γ-ray transport model to obtain the importance value of the neutron source for each neutron energy group to the secondary γ-ray absorbed dose: 6.1) Divide the secondary γ-ray transport model into grids and record the neutron fluence within each grid; 6.2) Use the neutron accompanying fluence and grid weight window minus variance parameter calculation formula to iteratively calculate and generate the weight window lower limit parameter of each grid. Substitute into the grid weight window variance reduction parameter calculation formula: Where w th,j is the lower limit parameter of the weight window of the j-th grid; λ' is the ratio of the upper limit parameter of the weight window to the lower limit parameter of the weight window; is the secondary γ adjoint fluence in the jth grid; is the maximum value among all secondary γ adjoint fluences; 6.3) According to the grid weight window lower limit parameter w in step 6.2) th,j , the grid weighted window variance reduction method is used to solve the secondary γ accompanying transport model, and the neutron accompanying fluence at different explosion heights and explosion center projection distances from the accompanying secondary γ source is obtained, which is the importance value of the neutron source of each neutron energy group to the secondary γ absorbed dose in the corresponding forward model; S7. Based on the importance value of the neutron source to the neutron absorbed dose obtained in S3 and the importance value of the neutron source to the secondary gamma absorbed dose obtained in S6, the neutron absorbed dose value and the secondary gamma absorbed dose value of the nuclear explosion neutron radiation source are calculated to complete the acquisition of the nuclear explosion neutron dose field: 7.1) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup on the neutron absorbed dose obtained from S3, and then add them together to obtain the absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions; 7.2) Multiply the intensity values ​​of each energy subgroup of any nuclear explosion radiation neutron source by the importance values ​​of each energy subgroup to the secondary gamma absorbed dose obtained from S6, and then add them together to obtain the secondary gamma absorbed dose values ​​of the nuclear explosion radiation neutron source under different explosion height conditions.

2. The method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport according to claim 1, characterized in that: S1 is specifically: 1.1) Set up a neutron point source as the companion neutron source; 1.2) Multiple detectors are set up at different blast heights and projection distances from the neutron source to serve as neutron counting units; 1.3) According to steps 1.1) and 1.2), the Monte Carlo geometric parameters of the neutron source's transport in the atmosphere are set to establish a neutron transport model.

3. The method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport according to claim 2, characterized in that: S2 is specifically: 2.1) The neutron transport model calculation mode is set to multi-group adjoint, and the multi-group neutron structure is subdivided using the logarithmic interpolation method to obtain neutron subdivided energy groups. At the same time, the average fluence-dose conversion coefficient of each neutron subdivided energy group is calculated; 2.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the neutron transport model based on the neutron subdivision energy group and fluence-dose average conversion coefficient obtained in step 2.1); 2.3) Set the source neutron weight wgt1 in the neutron transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 2.2).

4. The method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport according to claim 3, characterized in that: Step 2.1) is specifically as follows: 2.1.

1. Establish a point source-spherical MC calculation model; 2.1.

2. Set the parameters of the point source-sphere MC calculation model; The source neutrons of the point source-spherical MC calculation model are sampled according to the neutron subdivision energy group, and the sampling probability of each neutron subdivision energy group is set to 1; The total number of neutron energy groups is multiplied by the area of ​​the sphere to serve as the weight of the source neutrons in the model. A sphere with a radius of 1 cm is set to record the fluence of each neutron energy group. A fluence-dose conversion factor with a point-value response function is set to convert the recorded fluence into dose. 2.1.

3. Based on the parameters of the point source-spherical MC calculation model in step 2.1.2, obtain the counting results of each neutron energy group and use them as the corresponding average fluence-dose conversion coefficient; Step 2.2) is specifically as follows: 2.2.

1. Set the neutron subdivision energy group to the neutron sampling energy group of the neutron transport model; 2.2.

2. Set the average fluence-dose conversion coefficient of each neutron energy group obtained in step 2.1.3 as the neutron sampling probability of the neutron transport model; 2.2.

3. Set the neutron subdivision energy group to the counting energy group of the neutron transport model; 2.2.

4. Set all counting multipliers to 1; The source neutron weight wgt1 in step 2.3) is specifically: According to steps 2.2.1 to 2.2.4, the source neutron weight wgt1 can be obtained 5. A method for efficiently acquiring nuclear explosion neutron dose field based on MC co-transport according to any one of claims 1 to 4, characterized in that: S4 is specifically: 4.1) Set up a point source as an accompanying secondary gamma source; 4.2) Multiple detectors are set up at different explosion heights and projection distances from the secondary gamma ray source as neutron counting units; 4.3) According to steps 4.1) and 4.2), the Monte Carlo geometric parameters of the secondary gamma source transport in the atmosphere are set to establish a secondary gamma source transport model.

6. The method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport according to claim 5, characterized in that: S5 is specifically: 5.1) Set the secondary gamma ray transport model calculation mode to multi-group adjoint, and use the logarithmic interpolation method to subdivide the multi-group secondary gamma ray structure to obtain secondary gamma subdivision energy groups. At the same time, calculate the average fluence-dose conversion coefficient for each secondary gamma subdivision energy group; 5.2) Setting the sampling energy group, sampling probability, counting energy group, and counting multiplier of the secondary gamma ray transport model based on the secondary gamma ray subdivision energy group and the fluence-dose average conversion coefficient obtained in step 5.1); 5.3) Set the source secondary γ weight wgt2 in the secondary γ adjoint transport model based on the sampling energy group, sampling probability, counting energy group and counting multiplier obtained in step 5.2).

7. The method for efficiently acquiring the neutron dose field of a nuclear explosion based on MC co-transport according to claim 6, characterized in that: Step 5.1) is specifically as follows: 5.1.

1. Establish a point source-spherical MC calculation model; 5.1.

2. Set the parameters of the point source-sphere MC calculation model; The source gamma of the point source-sphere MC calculation model is sampled according to the secondary gamma subdivision energy group, and the sampling probability of each secondary gamma subdivision energy group is set to 1. The total number of secondary gamma subdivision energy groups is multiplied by the area of ​​the sphere as the weight of the source gamma in the model. A sphere with a radius of 1 cm is set to record the fluence of each secondary gamma subdivision energy group. A fluence-dose conversion factor with a point-valued response function is set to convert the recorded fluence into dose. 5.1.

3. Based on the parameters of the point source-spherical MC calculation model in step 5.1.2, obtain the counting results of each secondary gamma subdivision energy group and use them as the corresponding average fluence-dose conversion coefficient; Step 5.2) is specifically as follows: 5.2.

1. Set the secondary γ subdivision energy group to the secondary γ sampling energy group of the secondary γ adjoint transport model; 5.2.

2. Set the average fluence-dose conversion coefficient of each secondary gamma energy group obtained in step 5.1.3 as the secondary gamma sampling probability of the secondary gamma adjoint transport model; 5.2.

3. Set the neutron subdivision energy group to the neutron counting energy group of the secondary γ-linked transport model; 5.2.

4. Set all counting multipliers to 1; The source secondary γ weight wgt2 in step 5.3) is specifically: According to steps 5.2.1 to 5.2.4, the source secondary gamma weight wgt2 can be obtained

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