Retention volume acquisition method and device based on Monte Carlo simulation

Abstract

The invention belongs to the technical field of nuclear material retention volume analysis, and particularly relates to a retention volume acquisition method and device based on Monte Carlo simulation. The method comprises the following steps: according to Monte Carlo calculation requirements, obtaining physical geometric models of a to-be-detected process system and a detector, and completing material modeling; obtaining radiation parameters of retentate in the to-be-tested process system, and establishing a virtual source on the physical geometric model; on the basis of the physical geometric model, Monte Carlo simulation is adopted, and the response coefficient of the detector of the unit mass retentate at the target measurement position is obtained; and obtaining the retention volume of the target measurement position according to the neutron fluence measured by a detector and the response coefficient. The neutron detection efficiency calibration test and the correction analysis process are simplified, the reliability of the retention volume analysis result is improved, and important data are provided for nuclear material closing balance.

Description

Technical Field

[0001] The present application belongs to the technical field of nuclear material retention analysis, and specifically relates to a retention acquisition method and device based on Monte Carlo simulation. Background Art

[0002] The retention of nuclear materials in the process system of bulk nuclear facilities is a key data for achieving closed balance, and its accurate measurement and analysis are crucial. In nuclear material balance, non-destructive analysis mainly includes gamma measurement and analysis technology and neutron measurement and analysis technology. Gamma measurement and analysis is often used to measure the isotopic composition of nuclear materials and analyze low-density samples. Compared with gamma rays, neutrons emitted by nuclear materials have strong penetrating ability and less self-absorption, and are more suitable for retention measurement in closed systems with shielding layers. Neutron measurement and analysis technology mainly includes total neutron analysis method, coincident neutron analysis method, and multiplicity analysis method. These methods have been widely used in quantitative measurements of radioactive waste barrels, nuclear fuel assemblies, decommissioned nuclear facility pipelines, filters, etc.

[0003] For the analysis of holdup in process systems within nuclear facilities, accurate neutron measurement and measurement data analysis are difficult due to the complex equipment layout, strong neutron scattering and limited measurement space in nuclear facilities. The efficiency calibration of neutron detectors is one of the key technologies for applying neutron measurement and analysis technology to holdup measurement. Because the calibration under actual application conditions is difficult to achieve, and the geometric structure, spatial conditions, shielding materials, etc. of equipment and containers in the process system will affect the measurement results, the detector efficiency calibrated under laboratory conditions cannot be directly introduced into the calculation. In addition, the structure of equipment and containers in the process system is special and there is obvious mutual influence, making it difficult to use conventional correction methods for data processing. Summary of the invention

[0004] The purpose of the present application is to provide a method and device for obtaining retention based on Monte Carlo simulation, so as to solve the problem in the prior art that it is difficult to accurately calibrate the neutron detector under actual application conditions.

[0005] Technical solution to achieve the purpose of this application:

[0006] A first aspect of an embodiment of the present application provides a method for obtaining retention based on Monte Carlo simulation, the method comprising:

[0007] According to the Monte Carlo calculation requirements, obtain the physical geometric model of the process system and detector to be tested, and complete the material modeling;

[0008] Acquire radiation parameters of the retentate in the process system to be measured, and establish a virtual source on the physical geometric model;

[0009] Based on the physical geometric model, a Monte Carlo simulation is used to obtain a detector response coefficient of a unit mass of retentate at a target measurement position;

[0010] The retention amount at the target measurement position is obtained according to the neutron flux measured by the detector and the response coefficient.

[0011] Optionally, obtaining the radiation parameters of the retentate in the process system to be measured specifically includes:

[0012] Obtaining the chemical composition, nuclide composition and content of the retentate, and determining the neutron source and the content of the radioactive nuclides that produce the neutrons;

[0013] According to the neutron source, the neutron yield of the radioactive nuclide is determined; the neutron yield includes spontaneous fission neutron yield and (α, n) reaction neutron yield.

[0014] Optionally, if the neutron source is a certain nuclide in the retentate, determining the neutron yield of the radionuclide according to the neutron source specifically includes:

[0015] The neutron yield of the corresponding nuclide is determined as the neutron yield of the radioactive nuclide, and a single-source system is set.

[0016] Optionally, if the neutron source is a plurality of nuclides in a compound in the retentate, determining the neutron yield of the radioactive nuclides according to the neutron source specifically includes:

[0017] Determining the neutron yield of the radioactive nuclide according to the neutron yield of each nuclide, and setting up a multi-source system according to the ratio of the intensity of the neutron emission source;

[0018] Alternatively, the neutron yield of each nuclide can be normalized to a single-source system.

[0019] Optionally, if the neutron source is multiple nuclides of multiple compounds in the retentate, determining the neutron yield of the radioactive nuclides according to the neutron source specifically includes:

[0020] According to the classification of the compounds, the neutron yield of each nuclide in each compound is calculated respectively; combined with the content proportion of the compounds, a multi-source system is set according to the source strength ratio of each nuclide;

[0021] Alternatively, according to the classification of the compounds, the total neutron yield of each type of compound is calculated by normalization, and a multi-source system is set according to the ratio of the neutron source intensity of the compounds;

[0022] Alternatively, the total neutron yield of the retentate is calculated by normalization and is set as a single-source system.

[0023] Optionally, obtaining the retention amount at the target measurement position according to the neutron fluence measured by the detector and the response coefficient specifically includes:

[0024] Correcting the measured neutron fluence according to a calibration factor of the detector under a reference radiation field to obtain a corrected value;

[0025] The retention amount at the target measurement position is obtained according to the corrected value and the response coefficient.

[0026] Optionally, obtaining the retention amount at the target measurement position according to the corrected value and the response coefficient specifically includes:

[0027] The retention amount at the target measurement position is obtained according to formula (1):

[0028] N=x*η (1)

[0029] Wherein, N is the corrected value, x is the retention amount, and η is the response coefficient.

[0030] A second aspect of an embodiment of the present application provides a retention amount acquisition device based on Monte Carlo simulation, the device comprising:

[0031] The model building module is used to obtain the physical geometric model of the process system and detector to be tested according to the Monte Carlo calculation requirements, and complete the material modeling;

[0032] A virtual source establishment module, used to obtain radiation parameters of the retentate in the process system to be measured and establish a virtual source on the physical geometric model;

[0033] A coefficient acquisition module, used to acquire the response coefficient of the detector of the unit mass of the retentate at the target measurement position by using Monte Carlo simulation based on the physical geometric model;

[0034] The retention amount acquisition module is used to acquire the retention amount at the target measurement position according to the neutron flux measured by the detector and the response coefficient.

[0035] A third aspect of an embodiment of the present application provides a computer-readable storage medium having a computer program stored thereon; when the computer program is executed, any one of the retention amount acquisition methods based on Monte Carlo simulation provided in the first aspect of the embodiment of the present application is implemented.

[0036] The fourth aspect of an embodiment of the present application provides an electronic device, including a memory and a processor; a computer program is stored on the memory; when the processor runs the computer program, any one of the retention amount acquisition methods based on Monte Carlo simulation provided in the first aspect of the embodiment of the present application is implemented.

[0037] The beneficial technical effects of this application are:

[0038] The embodiment of the present application provides a method and device for obtaining retention based on Monte Carlo simulation, the method comprising: obtaining the physical geometric model of the process system to be tested and the detector according to the Monte Carlo calculation requirements, and completing the material modeling; obtaining the radiation parameters of the retention in the process system to be tested, and establishing a virtual source on the physical geometric model; based on the physical geometric model, using Monte Carlo simulation to obtain the response coefficient of the detector per unit mass of retention at the target measurement position; obtaining the retention at the target measurement position according to the neutron injection measured by the detector and the response coefficient. A physical model is established based on the structural layout of the process system to be tested, a source model is set in combination with the distribution of radiation source terms and key parameters, particle transport calculations are performed through the Monte Carlo model, the neutron detector is calibrated, and the simulated calculation value of the detection efficiency in the real application scenario is obtained, the neutron detection efficiency calibration test and correction analysis process are simplified, the reliability of the retention analysis results is improved, and important data is provided for the closed balance of nuclear materials. BRIEF DESCRIPTION OF THE DRAWINGS

[0039] Figure 1 A schematic diagram of a flow chart of a method for obtaining retention based on Monte Carlo simulation provided in an embodiment of the present application;

[0040] Figure 2 A schematic diagram of the neutron energy spectrum of an Am-Be source in a retention amount acquisition method based on Monte Carlo simulation provided in an embodiment of the present application. DETAILED DESCRIPTION

[0041] In order to make those skilled in the art better understand the present application, the technical scheme in the embodiments of the present application will be clearly and completely described below in conjunction with the drawings in the embodiments of the present application. Obviously, the embodiments described below are only a part of the embodiments of the present application, not all. Based on the embodiments recorded in the present application, all other embodiments obtained by those skilled in the art without paying creative work are within the scope of protection of the present application.

[0042] See also Figure 1 , which is a flow chart of a method for obtaining retention volume based on Monte Carlo simulation provided in an embodiment of the present application.

[0043] The embodiment of the present application provides a method for obtaining retention based on Monte Carlo simulation, comprising:

[0044] Step S101: according to the Monte Carlo calculation requirements, the physical geometric model of the process system to be tested and the detector is obtained, and the material modeling is completed;

[0045] In one example, a physical geometric model of the process system and detector can be established. Comprehensively analyze the key equipment and containers that are prone to retention in the closed system, determine the type of neutron detector and measurement points for retention measurement, obtain the relative position, geometric structure, material composition, density and other information of the key equipment, containers and detectors in the process system, and establish the physical geometric model of the process system and detector to be tested according to the Monte Carlo calculation requirements. The detector structure and material composition should be as detailed as possible, such as 3 He tube size, moderator size and other cladding material size, 3 The accuracy of the He tube filling gas composition, purity, filling pressure, material density and other information will directly affect the detector efficiency calculation results. Analyze the material parameters involved in the established physical model to create a material card and assign the corresponding geometric body in the geometric model to complete the material modeling.

[0046] Step S102: obtaining radiation parameters of the retentate in the process system to be measured, and establishing a virtual source on the physical geometric model.

[0047] As an example, the chemical composition, nuclide composition and content of the retentate can be analyzed to determine the neutron source and the content of the radioactive nuclides that produce neutrons. According to experimental measurement or calculation, the neutron spectrum of the retentate is obtained, and the neutron spectrum of each nuclide or the normalized neutron spectrum of each compound or retentate is calculated accordingly. According to the distribution characteristics of the radiation source items in the process system, a source model is added to the physical geometric model or the source coordinates and geometric parameters are directly input to establish a virtual source, and the material parameters and radiation parameters of the source are set according to the parameters of the radiation source. In the specific implementation, it should be noted that the virtual source setting method and parameter setting must be consistent.

[0048] In one example, the detector count needs to be set. According to the detector measurement output, the counting card parameters are set to record the neutron injection rate / count rate / surrounding dose equivalent rate in the detector sensitive volume. When calculating the count rate, the multiplication factor, material number, and nuclear reaction number need to be set.

[0049] Step S103: Based on the physical geometric model, a Monte Carlo simulation is used to obtain a response coefficient of the detector per unit mass of the retained matter at the target measurement position.

[0050] In the specific implementation, boundary conditions and truncation methods are set and Monte Carlo simulation is run to obtain the detector response coefficient η of unit mass retention at the target measurement position, that is, the neutron injection rate / counting rate / ambient dose equivalent rate generated in the sensitive volume.

[0051] It is understandable that Monte Carlo simulation is based on computer simulation and constructs a physical model according to the geometric characteristics of the physical process to conduct digital simulation experiments. It is an effective method for solving particle transport problems and will not be elaborated here.

[0052] Step S104: acquiring the retention amount at the target measurement position according to the neutron flux measured by the detector and the response coefficient.

[0053] The embodiment of the present application establishes a physical model based on the structural layout of the process system to be tested, sets the source model in combination with the distribution of radiation source terms and key parameters, performs particle transport calculations through the Monte Carlo model, calibrates the neutron detector, and obtains the simulated calculated value of the detection efficiency in the actual application scenario, thereby simplifying the neutron detection efficiency calibration test and correction analysis process, improving the reliability of the retention analysis results, and providing important data for the closure balance of nuclear materials.

[0054] In some possible implementations of the embodiments of the present application, step S102 may specifically include:

[0055] Obtaining the chemical composition, nuclide composition and content of the retentate, and determining the neutron source and the content of the radioactive nuclides that produce the neutrons;

[0056] According to the neutron source, the neutron yield of the radioactive nuclide is determined; the neutron yield includes spontaneous fission neutron yield and (α, n) reaction neutron yield.

[0057] In one example, if the neutron source is a certain nuclide in the retentate, then determining the neutron yield of the radionuclide according to the neutron source specifically includes:

[0058] The neutron yield of the corresponding nuclide is determined as the neutron yield of the radioactive nuclide, and a single-source system is set.

[0059] In another example, if the neutron source is a plurality of nuclides in a compound in the retentate, determining the neutron yield of the radioactive nuclides according to the neutron source may specifically include:

[0060] Determining the neutron yield of the radioactive nuclide according to the neutron yield of each nuclide, and setting up a multi-source system according to the ratio of the intensity of the neutron emission source;

[0061] Alternatively, the neutron yield of each nuclide can be normalized to a single-source system.

[0062] In another example, if the neutron source is multiple nuclides of multiple compounds in the retentate, determining the neutron yield of the radioactive nuclides according to the neutron source may specifically include:

[0063] According to the classification of the compounds, the neutron yield of each nuclide in each compound is calculated respectively; combined with the content proportion of the compounds, a multi-source system is set according to the source strength ratio of each nuclide;

[0064] Alternatively, according to the classification of the compounds, the total neutron yield of each type of compound is calculated by normalization, and a multi-source system is set according to the ratio of the neutron source intensity of the compounds;

[0065] Alternatively, the total neutron yield of the retentate is calculated by normalization and is set as a single-source system.

[0066] In some possible implementations of the embodiments of the present application, obtaining the retention amount at the target measurement position according to the neutron fluence measured by the detector and the response coefficient may specifically include:

[0067] Correcting the measured neutron fluence according to a calibration factor of the detector under a reference radiation field to obtain a corrected value;

[0068] The retention amount at the target measurement position is obtained according to the corrected value and the response coefficient.

[0069] In the specific implementation, the detector calibrated by the response test can be used to measure each determined measurement point to obtain the measured value N of the neutron injection rate / count rate / ambient dose equivalent rate of each point. m The measured value is corrected by the calibration factor of the detector under the reference radiation field, N = N m *k, where k is the calibration factor of the detector.

[0070] In one example, obtaining the retention amount at the target measurement position according to the corrected value and the response coefficient may specifically include:

[0071] The retention amount at the target measurement position is obtained according to formula (1):

[0072] N=x*η (1)

[0073] Wherein, N is the corrected value, x is the retention amount, and η is the response coefficient.

[0074] In practical applications, if the mutual influence of neutron measurement data between various measurement points is considered, it is necessary to calculate the detector response coefficient matrix, establish a multivariate linear equation system, and solve the retention amount of each concerned part, which will not be repeated here.

[0075] A method for obtaining retention volume based on Monte Carlo simulation provided in an embodiment of the present application is described in detail below with reference to two specific examples.

[0076] Embodiment 1:

[0077] The embodiment of the present application provides a method for obtaining the retention based on Monte Carlo simulation, assuming that a certain process system mainly includes two containers and pipelines, the container structure and material are the same, and the interval between the two containers is 120 cm. Except for the bottom of the container, the container is covered with a 20 cm thick polyethylene shielding layer. The retention is simulated with an Am-Be source to illustrate the calculation process, and the retention mass is calculated as the Am metal mass.

[0078] The embodiment of the present application provides a method for obtaining retention based on Monte Carlo simulation, which is carried out according to the following steps:

[0079] Step (1), after analysis, the system is sealed in a glove box, the main prone to retention is the two containers, and the retention in the pipeline can be ignored. The neutron measurement points are located directly below the two containers, the distance between the center of the detector and the bottom of the container is 17.5 cm, and the neutron detector model is ARM-Z06.

[0080] Step (2), establishing a geometric model according to the relative positions and geometric structures of the container, the detector, the polyethylene shielding body, etc., is mainly achieved by performing Boolean operations on bodies and surfaces.

[0081] Step (3), the materials involved in the geometric model are mainly polyethylene, 3 He, stainless steel, concrete, and air, their elemental composition, density and other parameters are shown in Table 1.

[0082] Table 1 Material parameters

[0083]

[0084] Create a material card based on the material parameter information and assign it to the corresponding geometry.

[0085] Step (4), after analysis, the retentate in the system is solid, and the neutrons it emits include fission neutrons produced by spontaneous fission of some nuclides; α particles emitted by α decay of nuclides produce neutrons in (α, n) reactions. The neutron emission rate of the retentate is calculated according to the following formula based on the isotope abundance, the spontaneous fission neutron yield of the isotope and the (α, n) reaction neutron yield.

[0086]

[0087] Where δ is the neutron emission rate of the material; m is the mass of the material; f i is the abundance of the ith isotope; Y SF is the spontaneous fission neutron yield per unit mass of the isotope; Y (α,n) is the neutron yield of the (α, n) reaction per unit mass of isotope.

[0088] The neutron emission rate per unit mass of retained material is calculated to be 7.56E+06n / s.

[0089] The neutron spectrum of Am-Be source is as follows Figure 2 shown.

[0090] Step (5), according to the distribution characteristics of the retentate, it is mainly deposited at the bottom of the container, so a source model is established at the bottom of the container as a cylindrical source with a height of 3 mm and a diameter the same as the inner diameter of the container, and the source material, particle type, energy spectrum, etc. are set.

[0091] Step (6), ARM-206 can output neutron fluence rate and neutron dose rate, and set the detector counting type to fluence and dose statistics.

[0092] 1) 3 The He tube geometry is the sensitive volume, and the neutron fluence or neutron dose (rate) in the sensitive volume of the detector is calculated. When calculating the neutron dose, the neutron fluence-dose conversion coefficient refers to ICRP Publication No. 74.

[0093] 2) You can also set up a virtual detector for calculation. 3 The center coordinates, radius, and height distribution of the He tube geometry are input to calculate the neutron injection or neutron dose (rate) within the volume.

[0094] Step (7), set the particle number cutoff to track 1E+08 particles (the relative standard deviation of the calculated results is within 1%).

[0095] Step (8), output result data processing, calculate the neutron injection rate per unit mass of the retained material in the detector sensitive volume.

[0096] Neutron injection rate per unit mass of retentate [n / (cm -2 ·s)] = program output neutron injection rate [1 / (cm -2 · Neutron)]*7.56E+06[n / s]

[0097] The response coefficient η of the detector to the unit mass of the retentate is calculated to be 346.14 n·cm -2 ·s -1 ·g -1 .

[0098] Step (9): Use ARM-206 to measure each of the determined measurement points to obtain the neutron flux measurement value N of each point. m 16.52n·cm -2 ·s -1 、16.59n·cm -2 ·s -1 .

[0099] Step (10), detector measurement value correction. According to the detector calibration certificate, the calibration factor is 1.05.

[0100] Step (11), retention calculation. The calculated N is 17.35n·cm -2 ·s -1 、17.41n·cm -2 ·s -1 Further, according to N = x*η, the material retention in the key equipment and container in the closed system is 50.12 mg and 50.31 mg respectively.

[0101] Embodiment 2:

[0102] The measurement conditions are the same as those in the first embodiment, except that the detector is replaced with a neutron counting rate measurement. If the detector is a LB123+LB6411 neutron ambient dose equivalent rate meter (hereinafter referred to as a dose rate meter), the neutron counting rate is measured.

[0103] Steps (1) to (5) are the same, only the detector model is different. In step (6), the detector counting type is set to reaction rate statistics. 3 The He tube geometry is the sensitive volume, and the calculation of the material neutrons emitted by the moderated body after reaching the detector sensitive volume and the working gas 3 The reaction rate of He.

[0104] Step (7) According to step (7) of Example 1, the particle number cutoff is set to track 1E+08 particles (the relative standard deviation of the calculated results is within 1%).

[0105] Step (8) outputs the result data processing and calculates the neutron count rate per unit mass of the retained material in the sensitive volume of the detector.

[0106] Neutron counting rate per unit mass of retained material [cps] = program output reaction rate [1 / (neutron * nucleon)] * sensitive volume 3 He nucleon number*7.56E+06[n / s]

[0107] The response coefficient η of the detector to the unit mass of the retentate is calculated to be 214.61 cps·g -1 .

[0108] Step (9): Use a dose rate meter to measure each of the determined measurement points to obtain a neutron count rate measurement value N at each point. m They are 11.08cps and 11.12cps respectively.

[0109] Step (10), detector measurement value correction. According to the detector calibration certificate, the calibration factor is 0.97. The calculated N is 10.75 cps and 10.78 cps respectively.

[0110] Step (11), calculation of retention amount: Further, according to N=x*η, the retention amounts of materials in the key equipment and container in the closed system are 50.09 mg and 50.25 mg respectively.

[0111] If dose rate measurement is used, the detector output result is set as dose rate statistics in step (6). The dose rate per unit mass of retentate at the detector position is further calculated to obtain the detector response coefficient for unit mass of retentate: 487.22 μSv·h·g -1 Similarly, the retention amount can be calculated from the neutron dose rate at each point.

[0112] Note: 1) Since radioactive measurement has the characteristics of statistical fluctuations, when conducting neutron measurement, multiple measurements should be carried out at each point to reduce the statistical error of measurement; 2) When there is obvious mutual influence between equipment and containers, it is necessary to establish a set of equations and determine the coefficient matrix; 3) The ideas proposed in this method are also applicable to the calibration of nuclear material retention measurement detectors in the closed system of similar nuclear facilities, but there are many factors affecting neutron measurement, and the correction of the influence of interfering neutrons still needs to be further optimized; 4) The actual distribution of retention in the equipment\container in the system and the differences in the geometry, position, uniformity, etc. of the source during detector calibration will cause deviations in the retention calculation results, which need to be optimized by establishing different source models. 5) Due to factors such as the structure, energy response, data acquisition and processing of different types of detectors, the correction coefficients for the neutron measurement and calculation values ​​of materials of different quality levels are not fixed values, and need to be experimentally corrected according to the actual measurement situation.

[0113] (1) The embodiment of the present application models the actual measurement scene of the closed process system, and calibrates the detector according to the radiation characteristics and actual distribution of the retained material. The calculation model can be optimized and adjusted according to the actual distribution of the retained material, the actual measurement conditions of the detector, etc., to obtain a detection efficiency that is as close to the real scene as possible. The Monte Carlo simulation calculation method overcomes the problems of difficulty in calibrating the detector in the actual measurement scene and the complexity of the measurement data correction process;

[0114] (2) Compared with the traditional neutron detector efficiency calibration method, the calibration method described in the embodiment of the present application is not limited by the test conditions, and the obtained detection efficiency can be directly used for retention calculation and analysis without correction, and can realize the simultaneous calibration of multiple detection efficiencies in complex systems such as single source-multiple detectors, multiple sources-single detector, and multiple sources-multiple detectors;

[0115] (3) The embodiment of the present application measures the system in different regions to calculate the total retention volume in the system and its distribution in each region. Compared with the overall measurement of the system retention volume, the regional measurement can effectively improve the accuracy of the system retention volume measurement.

[0116] (4) The embodiment of the present application proposes a method for applying the measurement of neutron fluence rate / counting rate / ambient dose equivalent rate to the measurement of retention amount in a closed system and nuclear material balance. The method is a relative measurement method. The obtained neutron detector response coefficient can be used as an input parameter for retention amount calculation. The retention amount measurement system and the calculation program are integrated into one. By measuring the neutron data in combination with preset parameters (such as the neutron fluence rate / counting rate / ambient dose equivalent rate corresponding to unit mass), the retention amount at the measured position can be directly output, thereby simplifying the data processing process and improving the measurement and analysis efficiency.

[0117] The embodiments of the present application can be widely used in the in-situ measurement and analysis of nuclear material retention in the closed system of similar nuclear facilities. Through the Monte Carlo simulation calculation method, the detection efficiency of multiple detectors and multiple measurement positions in actual application scenarios can be calibrated simultaneously, and the measurement results are corrected by comparing with the laboratory calibration. The nuclear material retention measurement in the closed system can be effectively analyzed and obtained, providing important data for the closed balance of nuclear materials.

[0118] Based on the above-mentioned embodiment, a retention amount acquisition method based on Monte Carlo simulation is provided, the embodiment of the present application also provides a retention amount acquisition device based on Monte Carlo simulation.

[0119] The embodiment of the present application provides a retention amount acquisition device based on Monte Carlo simulation, comprising:

[0120] The model building module is used to obtain the physical geometric model of the process system and detector to be tested according to the Monte Carlo calculation requirements, and complete the material modeling;

[0121] A virtual source establishment module, used to obtain radiation parameters of the retentate in the process system to be measured and establish a virtual source on the physical geometric model;

[0122] A coefficient acquisition module, used to acquire the response coefficient of the detector of the unit mass of the retentate at the target measurement position by using Monte Carlo simulation based on the physical geometric model;

[0123] The retention amount acquisition module is used to acquire the retention amount at the target measurement position according to the neutron flux measured by the detector and the response coefficient.

[0124] Based on the Monte Carlo simulation-based retention amount acquisition method and device provided in the above embodiments, the embodiments of the present application also provide a computer-readable storage medium on which a computer program is stored; when the computer program is executed, any one of the Monte Carlo simulation-based retention amount acquisition methods provided in the above embodiments is implemented.

[0125] Based on the Monte Carlo simulation-based retention amount acquisition method and device provided in the above embodiments, the embodiments of the present application also provide an electronic device, including a memory and a processor; a computer program is stored in the memory; when the processor runs the computer program, any one of the Monte Carlo simulation-based retention amount acquisition methods provided in the above embodiments is implemented.

[0126] The present application is described in detail above in conjunction with the accompanying drawings and embodiments, but the present application is not limited to the above embodiments, and various changes can be made within the knowledge of ordinary technicians in the field without departing from the purpose of the present application. Any content not described in detail in the present application can adopt the existing technology.

Claims

1. A method for obtaining retention based on Monte Carlo simulation, characterized in that: The method comprises: According to the Monte Carlo calculation requirements, obtain the physical geometric model of the process system and detector to be tested, and complete the material modeling; Acquire radiation parameters of the retentate in the process system to be measured, and establish a virtual source on the physical geometric model; Based on the physical geometric model, a Monte Carlo simulation is used to obtain a detector response coefficient of a unit mass of retentate at a target measurement position; The retention amount at the target measurement position is obtained according to the neutron flux measured by the detector and the response coefficient.

2. The method for obtaining retention based on Monte Carlo simulation according to claim 1, characterized in that: The step of obtaining the radiation parameters of the retentate in the process system to be measured specifically includes: Obtaining the chemical composition, nuclide composition and content of the retentate, and determining the neutron source and the content of the radioactive nuclides that produce the neutrons; According to the neutron source, the neutron yield of the radioactive nuclide is determined; the neutron yield includes spontaneous fission neutron yield and (α, n) reaction neutron yield.

3. The method for obtaining retention based on Monte Carlo simulation according to claim 2, characterized in that: If the neutron source is a certain nuclide in the retentate, then determining the neutron yield of the radionuclide according to the neutron source specifically includes: The neutron yield of the corresponding nuclide is determined as the neutron yield of the radioactive nuclide, and a single-source system is set.

4. The method for obtaining retention based on Monte Carlo simulation according to claim 2, characterized in that: If the neutron source is a plurality of nuclides in a compound in the retentate, then determining the neutron yield of the radioactive nuclides according to the neutron source specifically includes: Determining the neutron yield of the radioactive nuclide according to the neutron yield of each nuclide, and setting up a multi-source system according to the ratio of the intensity of the neutron emission source; Alternatively, the neutron yield of each nuclide can be normalized to a single-source system.

5. The method for obtaining retention based on Monte Carlo simulation according to claim 2, characterized in that: If the neutron source is multiple nuclides of multiple compounds in the retentate, determining the neutron yield of the radioactive nuclides according to the neutron source specifically includes: According to the classification of the compounds, the neutron yield of each nuclide in each compound is calculated respectively; combined with the content proportion of the compounds, a multi-source system is set according to the source strength ratio of each nuclide; Alternatively, according to the classification of the compounds, the total neutron yield of each type of compound is calculated by normalization, and a multi-source system is set according to the ratio of the neutron source intensity of the compounds; Alternatively, the total neutron yield of the retentate is calculated by normalization and is set as a single-source system.

6. The method for obtaining retention based on Monte Carlo simulation according to any one of claims 1 to 5, characterized in that: The step of obtaining the retention amount at the target measurement position according to the neutron flux measured by the detector and the response coefficient specifically includes: Correcting the measured neutron fluence according to a calibration factor of the detector under a reference radiation field to obtain a corrected value; The retention amount at the target measurement position is obtained according to the corrected value and the response coefficient.

7. The method for obtaining retention based on Monte Carlo simulation according to claim 6, characterized in that: The step of obtaining the retention amount at the target measurement position according to the corrected value and the response coefficient specifically includes: The retention amount at the target measurement position is obtained according to formula (1): N=x*η (1) Wherein, N is the corrected value, x is the retention amount, and η is the response coefficient.

8. A retention amount acquisition device based on Monte Carlo simulation, characterized in that: The device comprises: The model building module is used to obtain the physical geometric model of the process system and detector to be tested according to the Monte Carlo calculation requirements, and complete the material modeling; A virtual source establishment module, used to obtain radiation parameters of the retentate in the process system to be measured and establish a virtual source on the physical geometric model; A coefficient acquisition module, used to acquire the response coefficient of the detector of the unit mass of the retentate at the target measurement position by using Monte Carlo simulation based on the physical geometric model; The retention amount acquisition module is used to acquire the retention amount at the target measurement position according to the neutron flux measured by the detector and the response coefficient.

9. A computer-readable storage medium, characterized in that: A computer program is stored thereon; when the computer program is executed, the retention amount acquisition method based on Monte Carlo simulation as described in any one of claims 1 to 7 is implemented.

10. An electronic device, characterized in that: It comprises a memory and a processor; a computer program is stored in the memory; when the processor runs the computer program, the retention amount acquisition method based on Monte Carlo simulation as described in any one of claims 1 to 7 is implemented.

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