Neutron source intensity determination method, device, equipment, medium and program product
By obtaining the target characteristic parameters and fitting function of the antimony-beryllium neutron source, the nuclear reaction cross section and activity of antimony-124 can be directly calculated, which solves the problems of large computational load, long time and easy error in the existing technology. It realizes the efficient and accurate determination of the neutron source intensity and ensures the safe and controllable start-up of the reactor.
Patent Information
- Authority / Receiving Office
- CN · China
- Patent Type
- Applications(China)
- Current Assignee / Owner
- HUANENG (FUJIAN) ENERGY DEVELOPMENT LIMITED COMPANY FUZHOU BRANCH
- Filing Date
- 2026-01-20
- Publication Date
- 2026-06-12
AI Technical Summary
Existing methods for determining the intensity of a neutron source involve large computational loads, are time-consuming, complex, and prone to errors, making it difficult to effectively monitor and control the reactor startup process.
By obtaining the target characteristic parameters and fitting function of the fuel assembly where the antimony-beryllium neutron source is located, the nuclear reaction cross section for the generation of antimony-124 from antimony-123 is determined. The source strength of the antimony-beryllium neutron source is calculated based on the activity and fitting function, avoiding Monte Carlo modeling and directly calculating the activity and source strength of antimony-124.
This reduces the computational burden of determining the neutron source intensity, improves computational efficiency and accuracy, and ensures the safety and controllability of the reactor startup process.
Smart Images
Figure CN122197266A_ABST
Abstract
Description
Technical Field
[0001] This disclosure relates to the field of nuclear reactor technology, and in particular to a method, apparatus, equipment, medium, and procedure for determining the intensity of a neutron source. Background Technology
[0002] Currently, during reactor loading and startup, real-time reactivity monitoring is crucial to ensure subcritical state control, primarily through measuring the number of fission neutrons in the reactor core. Because the number of neutrons generated by spontaneous fission of nuclear fuel assemblies is extremely small, and the scattering and absorption of neutrons by coolant water, along with the shielding effect of internal metal components and thick pressure vessels, makes it difficult for external detectors to effectively capture neutron signals, creating a detection "blind zone." Therefore, a "neutron source" is introduced at the initial stage of reactor loading to increase the neutron "background level," enabling detectors to clearly record changes in fission neutron flux. This allows for real-time monitoring of the magnitude and rate of positive reactivity introduction, ensuring that the core effective growth factor Keff remains within the expected control range, thus guaranteeing the safety and controllability of the reactor startup process.
[0003] During reactor startup, the required secondary neutron source must reach a certain lower limit of source strength to ensure effective monitoring and control. For example, the antimony-beryllium (Sb-Be) neutron source must be irradiated for a sufficient duration at a specific power level to allow sufficient (γ,n) reactions to occur inside and accumulate a sufficient neutron yield in order to meet the requirements for the neutron background level during the startup phase and ensure that the reactor reaches the critical state safely and controllably.
[0004] Currently, the method for determining the intensity of a neutron source typically involves statistically analyzing the reactor power levels at various stages of operation and the characteristic parameters of different fuel assemblies, then performing Monte Carlo modeling, and finally calculating the neutron source intensity based on the Monte Carlo calculation model. However, this method suffers from problems such as high computational cost, long processing time, and a high susceptibility to errors. Summary of the Invention
[0005] To address the aforementioned technical problems, this disclosure provides a method, apparatus, device, medium, and program product for determining the intensity of a neutron source.
[0006] A first aspect of this disclosure provides a method for determining the intensity of a neutron source, comprising: The target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located are obtained. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. The target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source was determined based on the target characteristic parameters and the first fitting function. The activity of antimony-124 was determined based on the target nuclear reaction cross section; The target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and a second fitting function, which is used to characterize the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source.
[0007] In some embodiments of this disclosure, the target characteristic parameters include the neutron flux corresponding to the fuel assembly and the ratio of fast neutrons to thermal neutrons.
[0008] In some embodiments of this disclosure, obtaining the first fitting function includes: Multiple characteristic parameters corresponding to the fuel assembly were obtained, as well as the nuclear reaction cross section for the generation of antimony-124 from antimony-123 at different burnup states and different axial heights; the axial height is the height of the core block structure in the antimony-beryllium neutron source rod bundle along the axial direction in the fuel assembly guide tube; Sensitivity analysis was performed on multiple characteristic parameters and nuclear reaction cross sections to obtain the analysis results; The target characteristic parameter is determined from multiple characteristic parameters based on the analysis results. The target characteristic parameter is determined based on the correlation between multiple characteristic parameters and the nuclear reaction cross section. A regression analysis algorithm was used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, and the first fitting function was obtained.
[0009] In some embodiments of this disclosure, before determining the target source strength of the neutron source based on the activity of antimony-124 and a second fitting function, the method further includes: Obtain the second fitting function; Obtaining the second fitting function includes: The target activity of antimony-124 is calculated based on the nuclear reaction cross section; the photon transport is calculated based on the target activity of antimony-124 to obtain the photon nuclear reaction rate of beryllium-9; the total number of neutrons produced by the photon nuclear reaction is determined based on the photon nuclear reaction rate; and the total number of neutrons is determined as the source strength of the antimony-beryllium neutron source. A regression analysis algorithm was used to establish the relationship between target activity and source strength, resulting in a second fitting function.
[0010] In some embodiments of this disclosure, the activity of antimony-124 is determined based on the target nuclear reaction cross section, including: Obtain the density of antimony-123 in the antimony-beryllium neutron source and the corresponding neutron flux of the fuel assembly; The activity of antimony-124 was determined based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
[0011] In some embodiments of this disclosure, the target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and a second fitting function, including: The activity of antimony-124 is input into the second fitting function for calculation to obtain the target source strength.
[0012] A second aspect of this disclosure provides a neutron source intensity determination apparatus, comprising: The information acquisition module is used to acquire the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. The nuclear reaction cross section determination module is used to determine the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 in the antimony-beryllium neutron source based on the target characteristic parameters and the first fitting function. The activity determination module is used to determine the activity of antimony-124 based on the target nuclear reaction cross section; The neutron source intensity determination module is used to determine the target source intensity of the antimony-beryllium neutron source based on the activity of antimony-124 and a second fitting function. The second fitting function is used to characterize the relationship between the activity of antimony-124 and the source intensity of the antimony-beryllium neutron source.
[0013] In some embodiments of this disclosure, the target characteristic parameters include the neutron flux corresponding to the fuel assembly and the ratio of fast neutrons to thermal neutrons.
[0014] In some embodiments of this disclosure, the information acquisition module is specifically used to acquire multiple characteristic parameters corresponding to the fuel assembly, as well as the nuclear reaction cross section of antimony-123 generating antimony-124 at different burnup states and different axial heights; the axial height is the height of the core block structure in the antimony-beryllium neutron source rod bundle along the axial direction in the fuel assembly guide tube; Sensitivity analysis was performed on multiple characteristic parameters and nuclear reaction cross sections to obtain the analysis results; The target characteristic parameter is determined from multiple characteristic parameters based on the analysis results. The target characteristic parameter is determined based on the correlation between multiple characteristic parameters and the nuclear reaction cross section. A regression analysis algorithm was used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, and the first fitting function was obtained.
[0015] In some embodiments of this disclosure, the neutron source strength determination device further includes a function acquisition module.
[0016] The function acquisition module is used to acquire the second fitting function before determining the target source strength of the neutron source based on the activity of antimony-124 and the second fitting function.
[0017] The function acquisition module is specifically used to calculate the target activity of antimony-124 based on the nuclear reaction cross section; to calculate the photon transport based on the target activity of antimony-124 to obtain the photon nuclear reaction rate of beryllium-9; to determine the total number of neutrons produced by the photon nuclear reaction based on the photon nuclear reaction rate; and to determine the source strength of the antimony-beryllium neutron source based on the total number of neutrons. A regression analysis algorithm was used to establish the relationship between target activity and source strength, resulting in a second fitting function.
[0018] In some embodiments of this disclosure, the activity determination module is specifically used to obtain the density of antimony-123 in the antimony-beryllium neutron source and the neutron flux corresponding to the fuel assembly; The activity of antimony-124 was determined based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
[0019] In some embodiments of this disclosure, the neutron source intensity determination module is specifically used to input the activity of antimony-124 into a second fitting function for calculation to obtain the target source intensity.
[0020] A third aspect of this disclosure provides an electronic device, including: processor; Memory, used to store executable instructions; The processor is used to read executable instructions from memory and execute the executable instructions to implement the neutron source strength determination method provided in the first aspect above.
[0021] A fourth aspect of this disclosure provides a computer-readable storage medium storing a computer program that, when executed by a processor, causes the processor to implement the neutron source strength determination method provided in the first aspect.
[0022] A fifth aspect of this disclosure provides a computer program product comprising a computer program or instructions that, when executed by a processor, implement the neutron source strength determination method of the first aspect described above.
[0023] The technical solution provided in this disclosure has the following advantages compared with the prior art: The neutron source intensity determination method, apparatus, device, medium, and program products provided in this disclosure can obtain the target characteristic parameters and a first fitting function of the fuel assembly where the antimony-beryllium neutron source is located. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. After obtaining the target characteristic parameters and the first fitting function, the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source is determined based on the target characteristic parameters and the first fitting function. Then, the activity of antimony-124 is determined based on the target nuclear reaction cross section. The target source intensity of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and a second fitting function, wherein the second fitting function is used to characterize the relationship between the activity of antimony-124 and the source intensity of the antimony-beryllium neutron source. Therefore, based on the relationship between the target characteristic parameters and the nuclear reaction cross section in the first fitting function, the target nuclear reaction cross section for the antimony-123 to antimony-124 generation of the antimony-beryllium neutron source can be directly determined. After determining the activity of antimony-124 based on the target nuclear reaction cross section, the target source strength of the antimony-beryllium neutron source can be determined based on the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source in the second fitting function. This avoids the problems of large computational load, long time consumption, and complex and error-prone process in the process of performing Monte Carlo program modeling and calculating the source strength of the antimony-beryllium neutron source based on the Monte Carlo calculation model, thereby reducing the computational load for source strength determination and improving the efficiency and accuracy of source strength determination. Attached Figure Description
[0024] The accompanying drawings, which are incorporated in and form a part of this specification, illustrate embodiments consistent with this disclosure and, together with the description, serve to explain the principles of this disclosure.
[0025] To more clearly illustrate the technical solutions in the embodiments of this disclosure or the prior art, the accompanying drawings used in the description of the embodiments or the prior art will be briefly introduced below. Obviously, for those skilled in the art, other drawings can be obtained based on these drawings without creative effort.
[0026] Figure 1 This is a flowchart of a method for determining the intensity of a neutron source provided in an embodiment of this disclosure; Figure 2 This is a schematic diagram of a Monte Carlo computational model provided in an embodiment of this disclosure; Figure 3 This is a flowchart of another method for determining the intensity of a neutron source provided in this embodiment of the disclosure; Figure 4 This is a schematic diagram of a neutron source intensity determination device provided in an embodiment of this disclosure; Figure 5 This is a schematic diagram of the structure of an electronic device provided in an embodiment of this disclosure. Detailed Implementation
[0027] To better understand the above-mentioned objectives, features, and advantages of this disclosure, the solutions disclosed herein will be further described below. It should be noted that, unless otherwise specified, the embodiments and features described herein can be combined with each other.
[0028] Numerous specific details are set forth in the following description in order to provide a full understanding of this disclosure, but this disclosure may also be implemented in other ways different from those described herein; obviously, the embodiments in the specification are only some, and not all, of the embodiments of this disclosure.
[0029] It should be understood that the steps described in the method embodiments of this disclosure may be performed in different orders and / or in parallel. Furthermore, the method embodiments may include additional steps and / or omit the steps shown. The scope of this disclosure is not limited in this respect.
[0030] It should be noted that, in this document, relational terms such as "first" and "second" are used merely to distinguish one entity or operation from another, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Furthermore, the terms "comprising," "including," or any other variations thereof are intended to cover non-exclusive inclusion, such that a process, method, article, or apparatus that comprises a list of elements includes not only those elements but also other elements not expressly listed, or elements inherent to such a process, method, article, or apparatus. Without further limitations, an element defined by the phrase "comprising one..." does not exclude the presence of other identical elements in the process, method, article, or apparatus that includes said element.
[0031] It should be noted that the terms "a" and "a plurality of" used in this disclosure are illustrative rather than restrictive, and those skilled in the art should understand that, unless otherwise expressly indicated in the context, they should be understood as "one or more".
[0032] Typically, during reactor startup, the secondary neutron source needs to reach a certain lower limit of source strength to ensure effective monitoring and control. For example, the antimony-beryllium (Sb-Be) neutron source must be irradiated for a sufficient duration at a specific power level to allow sufficient (γ,n) reactions to occur inside and accumulate enough neutron yield in order to meet the neutron background level requirements during the startup phase and ensure that the reactor reaches criticality safely and controllably.
[0033] Currently, the method for determining the source strength of a neutron source typically involves statistically analyzing the reactor power level at various stages of operation and the characteristic parameters of each segment of different fuel assemblies, performing Monte Carlo modeling, and then calculating the neutron source strength based on the Monte Carlo calculation model. However, antimony-beryllium neutron sources are usually constructed as rod bundle structures and inserted into the guide tubes of fuel assemblies, with a certain range of axial height. The key parameters affecting the size of the antimony-beryllium neutron source in different assemblies can vary significantly, requiring segmented calculation of the source strength of the neutron source in different axial segments. The steps of the conventional algorithm are as follows: Based on the power level operation of the secondary source irradiated cycle reactor, divide the time period according to the approximate power, and statistically analyze the power level and operating time of each segment; extract data such as neutron flux rate, nuclear fuel nuclide composition, coolant density, and boron concentration of each segment of the fuel assembly corresponding to the antimony-beryllium core block of the secondary source assembly from the reactor core calculation program; perform Monte Carlo modeling on the fuel assembly where the secondary source is located based on the above information, and calculate the nuclear reaction cross section of antimony-123 to antimony-124 at different axial height segments and different burnup states; calculate the antimony-124 activity at the shutdown time according to the calculation formula and parameters of the antimony-123(n,γ) antimony-124 nuclear reaction equation, divide the cycle power difference, and calculate the photon intensity released by antimony-124 at the expected loading time of the next cycle; then use the Monte Carlo model, based on the photon intensity obtained in the previous step, calculate the photon flux rate of each energy group through photon transport, calculate the beryllium-9(γ,n) nuclear reaction rate, and obtain the final neutron source strength. Therefore, the Monte Carlo component model calculated in axial segments would have more than five parts. If different burn-out points are also considered, the amount of data and computation required would be even greater, resulting in a long processing time and a complex, error-prone process. To address this issue, this disclosure provides a method for determining the intensity of a neutron source, which will be described below with reference to specific embodiments.
[0034] Figure 1 This is a flowchart of a method for determining the intensity of a neutron source according to an embodiment of this disclosure. The method can be executed by a neutron source intensity determination device, which can be implemented in software and / or hardware. The neutron source intensity determination device can be configured in an electronic device, such as a server or terminal, wherein the terminal specifically includes a mobile phone, computer or tablet computer, etc.
[0035] like Figure 1 As shown, the method for determining the neutron source intensity provided in this embodiment includes the following steps.
[0036] S110. Obtain the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located.
[0037] In this embodiment of the disclosure, the antimony-beryllium neutron source, namely the Sb-Be neutron source, is a secondary neutron source. After the Sb-Be metal hybrid material is irradiated with neutrons, the antimony isotope antimony-123 is activated by reactor neutrons to antimony-124. The latter decays and emits gamma rays. The gamma rays bombard the beryllium isotope beryllium-9 in the hybrid material and release neutrons.
[0038] In this embodiment of the disclosure, the first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section.
[0039] In this embodiment of the disclosure, the target characteristic parameters may include the neutron flux corresponding to the fuel assembly and the ratio of fast neutrons to thermal neutrons.
[0040] The neutron flux of a fuel assembly is the number of neutrons passing through a unit area of the fuel assembly per unit time. It is a core parameter reflecting the power level of the nuclear reactor and the fission activity of the fuel assembly. The neutron number includes the number of thermal neutrons and the number of fast neutrons. Thermal neutrons and fast neutrons are two different energy states of neutrons in a nuclear reactor; thermal neutrons can be understood as neutrons with lower energy, for example, neutrons with energy below 0.025 eV; fast neutrons can be understood as neutrons with higher energy, for example, neutrons with energy above 1 MeV.
[0041] The nuclear reaction cross section is used to represent the probability of a specific nuclear reaction occurring between an incident particle and a target nucleus.
[0042] In some embodiments of this disclosure, the electronic device can respond to a command to determine the source strength of the antimony-beryllium neutron source, obtain the identification information of the fuel assembly, and obtain the target characteristic parameters and a first fitting function of the fuel assembly from a preset database based on the identification information of the fuel assembly.
[0043] In other embodiments of this disclosure, the electronic device can respond to a command to determine the source strength of the antimony-beryllium neutron source, obtain identification information of the fuel assembly, obtain multiple characteristic parameters of the fuel assembly from a preset database based on the identification information of the fuel assembly, wherein the multiple characteristic parameters include target characteristic parameters, construct a first fitting function based on the multiple characteristic parameters, and then obtain the target characteristic parameters and the first fitting function.
[0044] S120. Based on the target characteristic parameters and the first fitting function, determine the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source.
[0045] Specifically, after acquiring the target characteristic parameters and the first fitting function, the electronic device inputs the target characteristic parameters into the first fitting function to calculate the nuclear reaction cross section, thereby obtaining the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source.
[0046] S130. Determine the activity of antimony-124 based on the target nuclear reaction cross section.
[0047] In this embodiment of the disclosure, activity can be understood as the number of radioactive atomic nuclei that decay per unit time, that is, the number of particles released by a radioactive nuclide due to spontaneous decay per unit time. It directly reflects the decay rate of a radioactive substance.
[0048] Specifically, after obtaining the target nuclear reaction cross section for the generation of antimony-124 from antimony-123, the electronic device substitutes the target nuclear reaction cross section into the calculation formula for nuclear reaction cross section and activity to calculate the activity of antimony-124.
[0049] The target source strength of the antimony-beryllium neutron source is determined by S140, based on the activity of antimony-124 and a second fitting function.
[0050] In this embodiment of the disclosure, the second fitting function is used to characterize the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source.
[0051] Source strength can be understood as the number of neutrons emitted by a neutron source per unit time, and it is a core parameter for measuring the radiation output capability of a neutron source. In the embodiments of this disclosure, the source strength is the number of neutrons emitted by an antimony-beryllium neutron source per unit time.
[0052] Specifically, after obtaining the activity of antimony-124, the electronic device obtains a second fitting function, and calculates the source strength based on the activity of antimony-124 and the second fitting function to obtain the target source strength of the antimony-beryllium neutron source.
[0053] In this embodiment of the disclosure, the target characteristic parameters and a first fitting function of the fuel assembly containing the antimony-beryllium neutron source can be obtained. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. After obtaining the target characteristic parameters and the first fitting function, the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source is determined based on the target characteristic parameters and the first fitting function. Then, the activity of antimony-124 is determined based on the target nuclear reaction cross section. The target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and the second fitting function, wherein the second fitting function is used to characterize the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source. Therefore, based on the relationship between the target characteristic parameters and the nuclear reaction cross section in the first fitting function, the target nuclear reaction cross section for the antimony-123 to antimony-124 generation of the antimony-beryllium neutron source can be directly determined. After determining the activity of antimony-124 based on the target nuclear reaction cross section, the target source strength of the antimony-beryllium neutron source can be determined based on the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source in the second fitting function. This avoids the problems of large computational load, long time consumption, and complex and error-prone process in the process of performing Monte Carlo program modeling and calculating the source strength of the antimony-beryllium neutron source based on the Monte Carlo calculation model, thereby reducing the computational load for source strength determination and improving the efficiency and accuracy of source strength determination.
[0054] Based on the above embodiments of this disclosure, a first fitting function is constructed, i.e., a method for obtaining the first fitting function for the first time. Specifically, this may include: obtaining multiple characteristic parameters corresponding to the fuel assembly, and the nuclear reaction cross section of antimony-123 generating antimony-124 at different burnup states and different axial heights; performing sensitivity analysis on the multiple characteristic parameters and the nuclear reaction cross section to obtain analysis results; determining the target characteristic parameter from the multiple characteristic parameters based on the analysis results; and establishing the relationship between the target characteristic parameter and the nuclear reaction cross section using a regression analysis algorithm to obtain the first fitting function.
[0055] In this embodiment of the disclosure, multiple characteristic parameters include the thermal power of the fuel assembly, neutron flux, the ratio of fast neutrons to thermal neutrons, burnup, and coolant temperature.
[0056] Thermal power is the power generated by the fuel assembly. Burnup is the total energy released by the fuel assembly in the nuclear reactor, reflecting the extent to which fuel is consumed by fission reactions during reactor operation. Coolant temperature is the temperature of the water in the fuel assembly.
[0057] In this embodiment of the disclosure, the fuel consumption state point can be understood as a state point under different fuel consumption depths. For example, the fuel consumption state point can be three state points: initial fuel consumption, intermediate fuel consumption, and maximum fuel consumption, or it can be a state point under other fuel consumption conditions. There is no limitation here.
[0058] The axial height is the axial height of the core structure in the antimony-beryllium neutron source rod bundle within the fuel assembly guide tube.
[0059] In this embodiment of the disclosure, sensitivity analysis calculates how changes in other parameters affect the output result, i.e., the nuclear reaction cross section, using a specific quantitative parameter. The analysis results include the degree and pattern of influence of each characteristic parameter on the nuclear reaction cross section. Based on the analysis results, it can be determined that changes in other characteristic parameters besides the target characteristic parameter can affect the nuclear reaction cross section through changes in the target characteristic parameter.
[0060] In the embodiments of this disclosure, the target characteristic parameters are determined based on the correlation between multiple characteristic parameters and the nuclear reaction cross section.
[0061] In this embodiment of the disclosure, the regression analysis algorithm can be any existing regression analysis algorithm, such as the least squares method, and is not limited herein.
[0062] In some embodiments of this disclosure, the electronic device can obtain multiple characteristic parameters of the fuel assembly from a preset database for each type of fuel assembly based on the fuel assembly's identification information. After obtaining the multiple characteristic parameters, based on a preset calculation model and the multiple characteristic parameters, neutron transport calculation is used to obtain the nuclear reaction cross section of the selected type of fuel assembly for generating antimony-124 at different burnup state points and different axial heights.
[0063] In other embodiments of this disclosure, the electronic device can obtain multiple characteristic parameters of the fuel assembly from a preset database based on the fuel assembly's identification information for each type of fuel assembly, as well as the nuclear reaction cross section for generating antimony-124 from antimony-123 at different burnup states and different axial heights.
[0064] Furthermore, after obtaining the nuclear reaction cross section, sensitivity analysis is performed on multiple characteristic parameters and the nuclear reaction cross section to obtain analysis results. Based on these results, key parameters affecting the nuclear reaction cross section are then identified from among the multiple characteristic parameters, and target characteristic parameters are determined based on these key parameters. Finally, a regression analysis algorithm is used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, resulting in the first fitting function.
[0065] The preset calculation model can be a Monte Carlo calculation model, or any other calculation model that can calculate the nuclear reaction cross section based on the characteristic parameters of the fuel assembly; no restrictions are imposed here.
[0066] Figure 2 This is a schematic diagram of a Monte Carlo computational model provided in an embodiment of this disclosure, as shown below. Figure 2As shown, in this Monte Carlo calculation model, 10 (white dots) represents fuel rods; 20 (the area outside the fuel rods in each square) represents coolant; 30 (black dots) represents control components (components that control the nuclear reaction rate); and 40 (gray dots with black circles) represents secondary neutron sources such as antimony-beryllium neutron sources.
[0067] In this embodiment of the disclosure, before determining the target source strength of the neutron source based on the activity of antimony-124 and the second fitting function, the method may further include: obtaining the second fitting function.
[0068] Obtaining the second fitting function may specifically include: calculating the target activity of antimony-124 based on the nuclear reaction cross section; calculating the photon transport based on the target activity of antimony-124 to obtain the photon nuclear reaction rate corresponding to beryllium-9; determining the total number of neutrons generated by the photon nuclear reaction based on the photon nuclear reaction rate; determining the total number of neutrons as the source strength of the antimony-beryllium neutron source; and establishing the relationship between the target activity and the source strength using a regression analysis algorithm to obtain the second fitting function.
[0069] In this embodiment of the disclosure, the photon nuclear reaction rate can be understood as the total number of times a photon reacts with an atomic nucleus per unit time and unit volume.
[0070] In this embodiment, after obtaining the nuclear reaction cross-sections for the generation of antimony-124 from antimony-123 at different burnup states and axial heights, the electronic device calculates the amount of antimony-124 produced in segments based on the nuclear reaction cross-section and the formula for calculating the nuclear reaction cross-section and activity, according to the difference in cycle power level, to obtain the target activity of antimony-124. Based on the target activity of antimony-124, photon transport calculations are performed to obtain the photon nuclear reaction rate corresponding to beryllium-9. The product of the photon nuclear reaction rate and the number of neutrons produced in each nuclear reaction is determined as the total number of neutrons produced by the nuclear reaction, and the total number of neutrons is determined as the source strength of the antimony-beryllium neutron source. A regression analysis algorithm is used to establish the relationship between the target activity and the source strength to obtain a second fitting function.
[0071] In this embodiment, key parameters affecting the nuclear reaction cross section can be determined through sensitivity analysis, thereby determining the target characteristic parameters. A regression analysis algorithm is used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, as well as the relationship between activity and source intensity, to obtain a first fitting function and a second fitting function. This lays the foundation for subsequent calculation of the source intensity of the antimony-beryllium neutron source. The source intensity of the antimony-beryllium neutron source can be directly calculated based on the target characteristic parameters, simplifying the calculation of the source intensity of the antimony-beryllium neutron source and improving the calculation efficiency and accuracy.
[0072] In this embodiment of the disclosure, determining the activity of antimony-124 based on the target nuclear reaction cross section may specifically include: obtaining the density of antimony-123 in the antimony-beryllium neutron source and the neutron flux corresponding to the fuel assembly; and determining the activity of antimony-124 based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
[0073] Specifically, the electronic device can obtain the density of antimony-123 and the neutron flux corresponding to the fuel assembly from a preset database, calculate the product of the target nuclear reaction cross section and the density of antimony-123 to obtain a first value, and calculate the first value and the second value of the neutron flux, and determine the second value as the activity of antimony-124.
[0074] Furthermore, the target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and the second fitting function. Specifically, this may include inputting the activity of antimony-124 into the second fitting function for calculation to obtain the target source strength.
[0075] In this embodiment of the disclosure, the source strength of the antimony-beryllium neutron source is calculated based on the second fitting function and the activity of antimony-124, which improves the convenience, accuracy and efficiency of the source strength calculation.
[0076] Figure 3 This is a flowchart of another method for determining the intensity of a neutron source provided in this embodiment of the disclosure, as follows: Figure 3 As shown, the method for determining the source strength of a neutron source may include the following steps: S310. Obtain the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located.
[0077] S320. Based on the target characteristic parameters and the first fitting function, determine the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source.
[0078] S330: Obtain the density of antimony-123 in the antimony-beryllium neutron source and the corresponding neutron flux of the fuel assembly; determine the activity of antimony-124 based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
[0079] S340. Input the activity of antimony-124 into the second fitting function for calculation to obtain the target source strength of the antimony-beryllium neutron source.
[0080] It should be noted that the specific implementation methods of steps S310 to S340 are similar to the implementation methods of the corresponding steps in the above embodiments of this disclosure, and will not be repeated here.
[0081] In this embodiment, the target source strength of the antimony-beryllium neutron source is calculated based on the target characteristic parameters of the fuel assembly, a first fitting function, and a second fitting function. This simplifies and speeds up the entire calculation of the antimony-beryllium neutron source strength, eliminating the need for processing a large number of physical property parameters required for Monte Carlo modeling, and significantly reducing the overall calculation time. According to existing calibration results, the difference in source strength between the conventional complete calculation and the convenient calculation of this invention fully meets the accuracy requirements for starting the neutron source. Therefore, this reduces the computational load for source strength determination and improves the efficiency and accuracy of source strength determination.
[0082] Figure 4 This is a schematic diagram of a neutron source intensity determination device provided in an embodiment of this disclosure.
[0083] In this embodiment, the neutron source strength determination device can be housed within an electronic device and is understood as a functional module within the aforementioned electronic device. Specifically, the electronic device can be a server or a terminal, wherein the terminal specifically includes mobile phones, computers, or tablet computers, etc., without limitation.
[0084] like Figure 4 As shown, the neutron source intensity determination device 400 may include an information acquisition module 410, a nuclear reaction cross section determination module 420, an activity determination module 430, and a neutron source intensity determination module 440.
[0085] The information acquisition module 410 can be used to acquire the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. The nuclear reaction cross section determination module 420 can be used to determine the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source based on the target characteristic parameters and the first fitting function. The activity determination module 430 can be used to determine the activity of antimony-124 based on the target nuclear reaction cross section; The neutron source intensity determination module 440 can be used to determine the target source intensity of the antimony-beryllium neutron source based on the activity of antimony-124 and a second fitting function. The second fitting function is used to characterize the relationship between the activity of antimony-124 and the source intensity of the antimony-beryllium neutron source.
[0086] In this embodiment of the disclosure, the target characteristic parameters and a first fitting function of the fuel assembly containing the antimony-beryllium neutron source can be obtained. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. After obtaining the target characteristic parameters and the first fitting function, the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source is determined based on the target characteristic parameters and the first fitting function. Then, the activity of antimony-124 is determined based on the target nuclear reaction cross section. The target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and the second fitting function, wherein the second fitting function is used to characterize the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source. Therefore, based on the relationship between the target characteristic parameters and the nuclear reaction cross section in the first fitting function, the target nuclear reaction cross section for the antimony-123 to antimony-124 generation of the antimony-beryllium neutron source can be directly determined. After determining the activity of antimony-124 based on the target nuclear reaction cross section, the target source strength of the antimony-beryllium neutron source can be determined based on the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source in the second fitting function. This avoids the problems of large computational load, long time and low accuracy in the process of performing Monte Carlo program modeling and calculating the source strength of the antimony-beryllium neutron source based on the Monte Carlo calculation model, thereby reducing the computational load for source strength determination and improving the efficiency and accuracy of source strength determination.
[0087] In some embodiments of this disclosure, the target characteristic parameters include the neutron flux corresponding to the fuel assembly and the ratio of fast neutrons to thermal neutrons.
[0088] In some embodiments of this disclosure, the information acquisition module 410 can be specifically used to acquire multiple characteristic parameters corresponding to the fuel assembly, as well as the nuclear reaction cross section of antimony-123 generating antimony-124 at different burnup states and different axial heights; the axial height is the height of the core block structure in the antimony-beryllium neutron source rod bundle along the axial direction in the fuel assembly guide tube; Sensitivity analysis was performed on multiple characteristic parameters and nuclear reaction cross sections to obtain the analysis results; The target characteristic parameter is determined from multiple characteristic parameters based on the analysis results. The target characteristic parameter is determined based on the correlation between multiple characteristic parameters and the nuclear reaction cross section. A regression analysis algorithm was used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, and the first fitting function was obtained.
[0089] In some embodiments of this disclosure, the neutron source strength determination device 400 may further include a function acquisition module.
[0090] The function acquisition module can be used to acquire the second fitting function before determining the target source strength of the neutron source based on the activity of antimony-124 and the second fitting function.
[0091] The function acquisition module can be specifically used to calculate the target activity of antimony-124 based on the nuclear reaction cross section; to calculate the photon transport based on the target activity of antimony-124 to obtain the photon nuclear reaction rate of beryllium-9; to determine the total number of neutrons produced by the photon nuclear reaction based on the photon nuclear reaction rate; and to determine the source strength of the antimony-beryllium neutron source based on the total number of neutrons. A regression analysis algorithm was used to establish the relationship between target activity and source strength, resulting in a second fitting function.
[0092] In some embodiments of this disclosure, the activity determination module 430 can be specifically used to obtain the density of antimony-123 in the antimony-beryllium neutron source and the neutron flux corresponding to the fuel assembly; The activity of antimony-124 was determined based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
[0093] In some embodiments of this disclosure, the neutron source intensity determination module 440 can be specifically used to input the activity of antimony-124 into a second fitting function for calculation to obtain the target source intensity.
[0094] It should be noted that, Figure 4 The neutron source strength determination device 400 shown can execute the various steps in the above method embodiments and realize the various processes and effects in the above method embodiments, which will not be elaborated here.
[0095] Figure 5 This is a schematic diagram of the structure of an electronic device provided in an embodiment of this disclosure.
[0096] In this embodiment of the disclosure, Figure 5 The electronic devices shown can be servers or terminals, and terminals specifically include mobile phones, computers, or tablets, etc., without limitation.
[0097] like Figure 5 As shown, the electronic device may include a processor 510 and a memory 520 storing computer program instructions.
[0098] Specifically, the processor 510 may include a central processing unit (CPU), an application-specific integrated circuit (ASIC), or one or more integrated circuits that can be configured to implement the embodiments of this disclosure.
[0099] Memory 520 may include a large-capacity storage device for information or instructions. For example, and not limitingly, memory 520 may include a hard disk drive (HDD), a floppy disk drive, flash memory, optical disk, magneto-optical disk, magnetic tape, or a Universal Serial Bus (USB) drive, or a combination of two or more of these. Where appropriate, memory 520 may include removable or non-removable (or fixed) media. Where appropriate, memory 520 may be internal or external to the integrated gateway device. In a particular embodiment, memory 520 is a non-volatile solid-state memory. In a particular embodiment, memory 520 includes read-only memory (ROM). Where appropriate, the ROM may be a mask-programmed ROM, a programmable ROM (PROM), an erasable PROM (Electrically Programmable ROM, EPROM), an electrically erasable programmable PROM (EEPROM), an electrically alterable ROM (EAROM), or flash memory, or a combination of two or more of these.
[0100] The processor 510 reads and executes computer program instructions stored in the memory 520 to perform the steps of the neutron source strength determination method provided in this embodiment of the disclosure.
[0101] In one example, the electronic device may also include a transceiver 530 and a bus 540. Wherein, as... Figure 5 As shown, the processor 510, memory 520 and transceiver 530 are connected via bus 540 and communicate with each other.
[0102] Bus 540 may include hardware, software, or both. For example, and not limitingly, the bus may include an Accelerated Graphics Port (AGP) or other graphics bus, an Extended Industry Standard Architecture (EISA) bus, a Front Side Bus (FSB), a Hyper Transport (HT) interconnect, an Industrial Standard Architecture (ISA) bus, an Infinite Bandwidth Interconnect, a Low Pin Count (LPC) bus, a memory bus, a MicroChannel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a PCI-Express (PCI-X) bus, a Serial Advanced Technology Attachment (SATA) bus, a Video Electronics Standards Association Local Bus (VLB) bus, or other suitable buses, or a combination of two or more of these. Where appropriate, bus 540 may include one or more buses.
[0103] This disclosure also provides a computer-readable storage medium that can store a computer program. When the computer program is executed by a processor, the processor implements the neutron source strength determination method provided in this disclosure.
[0104] The aforementioned storage medium may, for example, include a memory 520 containing computer program instructions, which can be executed by a processor 510 of an electronic device to complete the neutron source strength determination method provided in this embodiment. Optionally, the storage medium may be a non-transitory computer-readable storage medium, such as a ROM, random access memory (RAM), compact disc ROM (CD-ROM), magnetic tape, floppy disk, and optical data storage device.
[0105] This disclosure also provides a computer program product, which includes a computer program or instructions. When the computer program or instructions are executed by a processor, they implement the neutron source strength determination method provided in this disclosure and can achieve the various processes and effects in the above embodiments of this disclosure, which will not be elaborated here.
[0106] The above description is merely a specific embodiment of this disclosure, enabling those skilled in the art to understand or implement it. Various modifications to these embodiments will be readily apparent to those skilled in the art, and the general principles defined herein may be implemented in other embodiments without departing from the spirit or scope of this disclosure. Therefore, this disclosure is not to be limited to the embodiments described herein, but is to be accorded the widest scope consistent with the principles and novel features disclosed herein.
Claims
1. A method for determining the intensity of a neutron source, characterized in that, include: Obtain the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. Based on the target characteristic parameters and the first fitting function, the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source is determined; The activity of antimony-124 was determined based on the target nuclear reaction cross section; The target source strength of the antimony-beryllium neutron source is determined based on the activity of antimony-124 and a second fitting function, wherein the second fitting function is used to characterize the relationship between the activity of antimony-124 and the source strength of the antimony-beryllium neutron source.
2. The method according to claim 1, characterized in that, The target characteristic parameters include the neutron flux corresponding to the fuel assembly and the ratio of fast neutrons to thermal neutrons.
3. The method according to claim 1, characterized in that, Obtaining the first fitting function includes: Multiple characteristic parameters corresponding to the fuel assembly are obtained, as well as the nuclear reaction cross section for the generation of antimony-124 from antimony-123 at different burnup states and different axial heights; the axial height is the height of the core block structure in the antimony-beryllium neutron source rod bundle along the axial direction in the guide tube of the fuel assembly; Sensitivity analysis was performed on the multiple characteristic parameters and the nuclear reaction cross section to obtain the analysis results; Based on the analysis results, a target characteristic parameter is determined from the plurality of characteristic parameters, wherein the target characteristic parameter is determined based on the correlation between the plurality of characteristic parameters and the nuclear reaction cross section; A regression analysis algorithm is used to establish the relationship between the target characteristic parameters and the nuclear reaction cross section, thus obtaining the first fitting function.
4. The method according to claim 3, characterized in that, Before determining the target source strength of the neutron source based on the activity of the antimony-124 and the second fitting function, the method further includes: Obtain the second fitting function; The step of obtaining the second fitting function includes: The target activity of antimony-124 is calculated based on the nuclear reaction cross section; the photon transport is calculated based on the target activity of antimony-124 to obtain the photon nuclear reaction rate of beryllium-9; the total number of neutrons generated by the photon nuclear reaction is determined based on the photon nuclear reaction rate; and the total number of neutrons is determined as the source strength of the antimony-beryllium neutron source. The regression analysis algorithm is used to establish the relationship between the target activity and the source strength, and the second fitting function is obtained.
5. The method according to claim 1, characterized in that, The determination of the activity of antimony-124 based on the target nuclear reaction cross section includes: Obtain the density of antimony-123 in the antimony-beryllium neutron source and the neutron flux corresponding to the fuel assembly; The activity of antimony-124 is determined based on the target nuclear reaction cross section, the density of antimony-123, and the neutron flux.
6. The method according to claim 1, characterized in that, The determination of the target source strength of the antimony-beryllium neutron source based on the activity of antimony-124 and the second fitting function includes: The activity of antimony-124 is input into the second fitting function for calculation to obtain the target source strength.
7. A device for determining the intensity of a neutron source, characterized in that, include: The information acquisition module is used to acquire the target characteristic parameters and the first fitting function of the fuel assembly where the antimony-beryllium neutron source is located. The first fitting function is used to characterize the relationship between the target characteristic parameters and the nuclear reaction cross section. The nuclear reaction cross section determination module is used to determine the target nuclear reaction cross section for the generation of antimony-124 from antimony-123 corresponding to the antimony-beryllium neutron source based on the target characteristic parameters and the first fitting function. An activity determination module is used to determine the activity of antimony-124 based on the target nuclear reaction cross section; The neutron source intensity determination module is used to determine the target source intensity of the antimony-beryllium neutron source based on the activity of antimony-124 and a second fitting function. The second fitting function is used to characterize the relationship between the activity of antimony-124 and the source intensity of the antimony-beryllium neutron source.
8. An electronic device, characterized in that, include: processor; Memory, used to store executable instructions; The processor is configured to read the executable instructions from the memory and execute the executable instructions to implement the neutron source strength determination method according to any one of claims 1-6.
9. A computer-readable storage medium, characterized in that, The storage medium stores a computer program, which, when executed by a processor, causes the processor to implement the neutron source strength determination method according to any one of claims 1-6.
10. A computer program product, the computer program product comprising a computer program or instructions, characterized in that, When the computer program or instructions are executed by the processor, they implement the method for determining the neutron source strength as described in any one of claims 1-6.