Self-powered neutron detector emitter burnup lifetime analysis method and apparatus

Abstract

This invention discloses a method and apparatus for analyzing the burnup lifetime of a self-powered neutron detector emitter, comprising: performing generational processing on the total neutron fluence absorbed by the self-powered neutron detector throughout its service life, and simultaneously performing layered processing on the emitter of the self-powered neutron detector; performing burnup analysis calculations on the neutron fluence of each generation and on each emitter layer; during the calculation of the neutron fluence of each generation, iteratively calculating the atomic nucleus density of each emitter layer for different service periods using the burnup calculation formula, thereby obtaining the burnup distribution information of the emitter under the total neutron fluence condition during the service period; wherein, during the iterative calculation process, for a certain iteration step, the atomic nucleus density of each emitter layer and the incident neutron fluence considering the self-shielding effect are given by the previous iteration calculation process. This invention can greatly improve the accuracy of burnup analysis of self-powered neutron detectors, providing more substantial and accurate data support for the design, maintenance, and replacement strategies of self-powered neutron detectors.

Description

Technical Field

[0001] This invention relates to the field of reactor neutron detector design and analysis technology, specifically to a method and equipment for analyzing the burnup lifetime of a self-powered neutron detector emitter. Background Technology

[0002] A self-powered neutron detector consists of several parts, including an emitter, an insulating layer, a collecting electrode, and a signal transmission core. The emitter material undergoes a nuclear reaction with neutrons in the reactor to produce beta electrons, thus enabling neutron detection. The intensity of the beta current signal is proportional to the neutron fluence rate in the reactor; therefore, the reactor neutron fluence rate (power) level can be monitored by measuring the output current signal of the self-powered neutron detector. Commonly used emitter materials for self-powered neutron detectors include rhodium, vanadium, silver, and hafnium—metals with large neutron absorption cross-sections that react with neutrons to produce beta and gamma rays. Because the detector is constantly exposed to high neutron fluence irradiation in the reactor, the atomic density of the neutron conversion materials (rhodium, vanadium, silver, hafnium, etc.) within the emitter decreases with increasing service time. This increased burnup leads to a decrease in the detector's neutron sensitivity, ultimately causing complete detector failure. Therefore, emitter burnup is a key factor affecting the service life of a self-powered neutron detector, and its impact must be considered during the design and development of the detector to accurately assess its service life within the reactor.

[0003] Currently, the commonly used formula for analyzing the burnup of self-powered neutron detectors is: Where N0 is the initial number of nucleons in the emitter, and K Φ σ is the self-shielding factor of the emitter material against neutrons, and σ is the neutron reaction cross section. To calculate the cumulative neutron flux, the remaining emitter nucleon density can be determined by calculating the cumulative neutron flux under different service life conditions. However, the self-shielding factor K increases due to the consumption of emitter nuclei under different service life conditions. Φ This will also change accordingly. The existing burnup analysis methods mentioned above ignore the influence of this effect, resulting in a significant difference between the theoretical burnup lifetime of self-sufficient neutron detectors and the actual situation. Summary of the Invention

[0004] The purpose of this invention is to improve the accuracy of burnup analysis for self-sustaining neutron detectors, providing more comprehensive and accurate data support for the design and maintenance / replacement strategies of self-sustaining neutron detectors during service. This invention aims to provide a method and apparatus for analyzing the burnup lifetime of a self-sustaining neutron detector emitter. It proposes a novel method for generational calculation of the total incident neutron fluence and layered processing of the emitter. This method considers the differences in neutron self-shielding effects within the emitter, as well as the changes in neutron self-shielding effects under different cumulative neutron fluence conditions. It can be used to quickly and accurately analyze the specific distribution information of emitter burnup under different cumulative neutron fluence irradiation conditions during different service periods within a reactor, thereby greatly improving the accuracy of burnup analysis for self-sustaining neutron detectors and providing more comprehensive and accurate data support for the design and maintenance / replacement strategies of self-sustaining neutron detectors during service.

[0005] This invention is achieved through the following technical solution:

[0006] In a first aspect, the present invention provides a method for analyzing the burnup lifetime of a self-powered neutron detector emitter, the method comprising:

[0007] The total neutron fluence absorbed by the self-powered neutron detector during its entire service life is processed in generations, and the emitter of the self-powered neutron detector is processed in layers. Burnup analysis and calculation are carried out in each generation of neutron fluence and in each layer of emitter material. In the process of calculating the neutron fluence of each generation, the atomic nucleus density of each layer of emitter at different service periods is calculated iteratively by the burnup calculation formula, so as to obtain the burnup distribution information of the emitter under the total neutron fluence condition during the service period, and realize the analysis and evaluation of the burnup lifetime of the self-powered neutron detector.

[0008] In the iterative calculation of burnup using the burnup formula, for a given iteration step, the atomic nucleus density of each layer of the emitter and the incident neutron fluence considering the self-shielding effect are given by the previous iteration. Since the neutron fluence calculated in each iteration step is low, the influence of neutron self-shielding changes in each iteration step can be ignored. Furthermore, the emitter is divided into several layers, each with a thickness of approximately a few micrometers, and it is assumed that the atomic nucleus density of each layer is uniformly distributed. For a given emitter layer, the incident neutron fluence needs to consider the influence of the self-shielding effect of the external emitter layers, and this neutron fluence is used to calculate the emitter atomic nucleus density at the end of the current iteration.

[0009] Furthermore, during the calculation of neutron flux in each generation, the neutron self-shielding factor within each emitter layer remains unchanged, while the neutron self-shielding factor differs between different emitter layers, and the neutron self-shielding factor is continuously updated as the neutron generation changes.

[0010] Furthermore, the total neutron fluence absorbed by the self-powered neutron detector during its entire service life is processed in generations, including:

[0011] The total neutron flux absorbed by the self-powered neutron detector during its entire service life is divided into several generations. The decrease in the number of nucleons in the emitter caused by the nuclear reaction between the neutrons and the emitter in each generation is very low, that is, the effect of burnup on the reaction rate can be ignored within each generation.

[0012] Furthermore, the emitter of the self-powered neutron detector undergoes layered processing, including:

[0013] The material of the self-powered neutron detector is layered, with each layer only on the order of micrometers in thickness. Within each layer, the self-shielding effect of the material on the incident neutron can be considered to remain unchanged.

[0014] The incident neutron fluence required for the burnup calculation of each emitter layer takes into account the absorption of the initial incident neutron fluence by the preceding layers, that is, the influence of the self-shielding effect is added between the emitter materials of each layer.

[0015] Furthermore, the formula for calculating the burnup of each launcher layer is as follows:

[0016] N(t) = N0exp -Iσt

[0017] I = (A + A') / 2

[0018] A′=Aexp (-N×d×σ)

[0019] In the formula, N(t) is the burnup of each emitter layer; N0 is the initial number of emitter nucleons; I is the incident neutron flux required for burnup calculation of each emitter layer in each iteration step; σ is the neutron reaction cross section; t is time; A is the parallel incident neutron flux intensity; A′ is the flux intensity of the neutron flux after penetrating the emitter material of this layer, i.e., A of the next layer; N is the atomic nucleus surface density of the emitter material; d is the emitter layer thickness; σ is the cross section where neutrons react with the emitter nuclei.

[0020] Furthermore, the burnup distribution of each emitter layer under each generation's neutron flux condition is analyzed using the burnup calculation formula for each generation; the burnup calculation formula for each generation is as follows:

[0021] N(i) = N0exp -Iσi

[0022] N(i+1)=N(i)exp -Iσ

[0023]

[0024] In the formula, N(i) is the atomic nucleus density of the emitter material in a certain emitter layer, i is the neutron generation; N0 is the initial number of nucleons in the emitter; I is the incident neutron fluence required for burnup calculation of each emitter layer in each iteration step; σ is the neutron reaction cross section; N(i+1) is the atomic nucleus density of the emitter material in a certain emitter layer, i+1 generation; and a is the number of incident neutrons in each generation.

[0025] Furthermore, the change of atomic nucleus density of each emitter layer material with the cumulative neutron flux is iteratively calculated according to the N(i+1) calculation formula until the total neutron flux is reached, thus completing the entire burnup analysis process of the self-powered neutron detector; and the entire burnup analysis process takes a relatively short time.

[0026] Furthermore, this method is applied to self-powered neutron detector emitters used in reactors, including but not limited to... 103 Rh、 50 V. 59 Materials such as Co, Pt, and Hf.

[0027] Secondly, the present invention provides a computer device, including a memory, a processor, and a computer program stored in the memory and executable on the processor, wherein the processor executes the computer program to implement the above-described method for analyzing the burnup lifetime of a self-powered neutron detector emitter.

[0028] Thirdly, the present invention provides a computer-readable storage medium storing a computer program that, when executed by a processor, implements the above-described method for analyzing the burnup lifetime of a self-powered neutron detector emitter.

[0029] Compared with the prior art, the present invention has the following advantages and beneficial effects:

[0030] This invention discloses a method and apparatus for analyzing the burnup lifetime of a self-powered neutron detector emitter. The invention proposes a novel method for generational calculation of the total incident neutron fluence and layered processing of the emitter. It performs generational processing of the total neutron fluence absorbed by the self-powered neutron detector throughout its entire service life, and simultaneously layers the emitter. Burnup analysis calculations are conducted for each generation of neutron fluence and within each layer of emitter material. During the calculation of each generation of neutron fluence, the atomic nucleus density of each layer of emitter at different service lifespans is iteratively calculated using the burnup calculation formula. This method provides information on emitter burnup distribution under total neutron flux conditions during service. It considers the differences in neutron self-shielding effects within the emitter, as well as the changes in neutron self-shielding effects under different cumulative neutron flux conditions. It can be used to quickly and accurately analyze the specific burnup distribution information of self-powered neutron detectors under different cumulative neutron flux irradiation conditions during different service periods in the reactor. This can greatly improve the accuracy of burnup analysis of self-powered neutron detectors and provide more substantial and accurate data support for the design of self-powered neutron detectors and maintenance and replacement strategies during service. Attached Figure Description

[0031] The accompanying drawings, which are included to provide a further understanding of embodiments of the invention and form part of this application, do not constitute a limitation thereof. In the drawings:

[0032] Figure 1 This is a schematic diagram of the self-powered neutron detector structure, which is a method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to the present invention.

[0033] Figure 2 This is a comparison of the burnup analysis results of a self-powered neutron detector using the burnup lifetime analysis method for the emitter of a self-powered neutron detector according to the present invention. Detailed Implementation

[0034] To make the objectives, technical solutions, and advantages of the present invention clearer, the present invention will be further described in detail below with reference to the embodiments and accompanying drawings. The illustrative embodiments and descriptions of the present invention are only used to explain the present invention and are not intended to limit the present invention.

[0035] The commonly used formula for analyzing the burnup of self-powered neutron detectors is as follows: Where N0 is the initial number of nucleons in the emitter, and K Φ σ is the self-shielding factor of the emitter material against neutrons, and σ is the neutron reaction cross section. To calculate the cumulative neutron flux, the remaining emitter nucleon density can be determined by calculating the cumulative neutron flux under different service life conditions. However, the self-shielding factor K increases due to the consumption of emitter nuclei under different service life conditions. ΦThis will also change accordingly. The existing burnup analysis methods mentioned above ignore the influence of this effect, resulting in a significant difference between the theoretical burnup lifetime of self-sufficient neutron detectors and the actual situation.

[0036] Therefore, to address the above problems, this invention designs a method and apparatus for analyzing the burnup lifetime of a self-powered neutron detector emitter. The design concept of this invention is to perform generational processing on the cumulative neutron flux absorbed by the self-powered neutron detector during its service life, for example, approximately 10... 10 ~10 11 Each generation has a neutron injection of approximately 10. 12 Simultaneously, the entire emitter material is layered, with each layer being approximately a few micrometers thick. In this way, during the calculation of neutron flux in each generation, the neutron self-shielding factor within each emitter layer remains unchanged, while the neutron self-shielding factors between different emitter layers are different. Furthermore, the neutron self-shielding factor is continuously updated as the neutron generation changes, ultimately providing information on the emitter burnup distribution under the total neutron flux conditions during service.

[0037] (1) The self-powered neutron detector emitter burnup lifetime analysis method of the present invention analyzes the consumption of atomic nuclei density of the self-powered neutron detector emitter material through numerical calculation. The principle formula is as follows:

[0038]

[0039] Where N0 is the initial number of nucleons in the emitter, and K Φ σ is the self-shielding factor of the emitter material against neutrons, and σ is the neutron reaction cross section. The nucleus density of the remaining emitter can be determined by calculating the cumulative neutron flux under different service life conditions.

[0040] (2) The core design point of the self-powered neutron detector emitter burnup lifetime analysis method of this invention is to perform layered processing on the emitter of the self-powered neutron detector. The thickness of each emitter layer is about a few micrometers. The difference in neutron self-shielding effect within each emitter layer is not significant. It can be assumed that the self-shielding factor within each emitter layer is constant, and its value is half of the self-shielding effect at the maximum thickness of each emitter layer. Assuming the emitter layer thickness is d, the atomic nucleus surface density of the emitter material is N, the parallel incident neutron flux intensity is A, and the cross section of the nuclear reaction between the neutron and the emitter atomic nucleus is σ, then the flux intensity A′ of the neutron flux after penetrating the emitter material layer is:

[0041] A′=Aexp (-N×d×σ) (2)

[0042] A′ serves as the next layer A, and the change in current intensity reflects the change in the self-shielding factor.

[0043] As can be seen from formula (2), even if the emitter layer is very thin, the intensity of the neutron flux will always decrease after penetration, and the self-shielding effect will always exist. On the other hand, since the emitter layer is very thin, it can be assumed that the self-shielding factor is linearly distributed inside the emitter layer, and its mean value can be used as the self-shielding factor required for burnup analysis calculation. Therefore, considering the self-shielding effect, the actual neutron flux used for burnup calculation of each emitter layer is:

[0044] I=(A+A′) / 2 (3)

[0045] Therefore, the burnup calculation formula (1) for each emitter layer can be expressed as:

[0046] N(t) = N0exp -Iσt (4)

[0047] (3) The total neutron fluence absorbed by the self-powered neutron detector during its service life is processed in generations, with the number of incident neutrons in each generation being a (approximately 10^6). 12 Then formula (4) can be expressed as:

[0048] N(i) = N0exp -Iσi (5)

[0049] Where i is the neutron generation, and I is related to the number of incident neutrons in each generation and the atomic nucleus density of that emitter layer. For a given emitter layer, since the number of incident neutrons is related to the total neutron fluence (approximately 10^- ... 22 Compared to the previous generation, the decrease in the number of atomic nuclei in the emitter material caused by each generation of neutron incidence is extremely small. Therefore, it can be assumed that the self-shielding effect within each emitter layer remains unchanged under each generation of neutron incidence conditions. However, the self-shielding effect between different neutron generations and different emitter layers is iteratively updated based on the actual remaining atomic nuclei density of the emitter material. Therefore, for a given emitter layer, the relationship between the atomic nuclei density N(i) of the current emitter material and the atomic nuclei density N(i+1) of the next generation emitter material is as follows:

[0050] N(i+1)=N(i)exp -Iσ (6)

[0051] And according to formula (3), I can be expressed as:

[0052]

[0053] According to formula (7), the incident neutron flux required for each emitter layer to carry out burnout calculation in each iteration step can be derived; according to formula (6), the change of atomic nucleus density of each emitter layer material with the cumulative neutron flux can be calculated iteratively until the total neutron flux is reached, thus completing the entire self-powered neutron detector burnout analysis process.

[0054] The above analysis shows that the self-powered neutron detector emitter burnup lifetime analysis method proposed in this invention is a novel method that calculates the total incident neutron fluence in generations and processes the emitter in layers. This method considers the differences in neutron self-shielding effects inside the emitter, as well as the changes in neutron self-shielding effects under different cumulative neutron fluence conditions. It can more accurately assess the burnup lifetime of self-powered neutron detectors in reactors during long-term operation, and provides a more accurate and substantial basis for the design and replacement of self-powered neutron detectors.

[0055] Example 1

[0056] like Figure 1 and Figure 2 As shown, this invention provides a method for analyzing the burnup lifetime of a self-powered neutron detector emitter, the method comprising:

[0057] The total neutron fluence absorbed by the self-powered neutron detector during its entire service life is processed in generations, and the emitter of the self-powered neutron detector is processed in layers. Burnup analysis and calculation are carried out in each generation of neutron fluence and in each layer of emitter material. In the process of calculating the neutron fluence of each generation, the atomic nucleus density of each layer of emitter at different service periods is calculated iteratively by the burnup calculation formula, so as to obtain the burnup distribution information of the emitter under the total neutron fluence condition during the service period, and realize the analysis and evaluation of the burnup lifetime of the self-powered neutron detector.

[0058] In the iterative calculation of burnup using the burnup formula, for a given iteration step, the atomic nucleus density of each layer of the emitter and the incident neutron fluence considering the self-shielding effect are given by the previous iteration. Since the neutron fluence calculated in each iteration step is low, the influence of neutron self-shielding changes in each iteration step can be ignored. Furthermore, the emitter is divided into several layers, each with a thickness of approximately a few micrometers, and it is assumed that the atomic nucleus density of each layer is uniformly distributed. For a given emitter layer, the incident neutron fluence needs to consider the influence of the self-shielding effect of the external emitter layers, and this neutron fluence is used to calculate the emitter atomic nucleus density at the end of the current iteration.

[0059] This embodiment focuses on a specific self-powered neutron detector. The emitter is divided into several layers, and the total neutron flux absorbed throughout its service life is divided into several generations. The atomic nucleus density of each emitter layer is iteratively calculated using a burnup calculation formula for different service periods, thereby enabling the analysis and evaluation of the self-powered neutron detector's burnup lifetime. During the iterative burnup calculation, for each iteration step, the atomic nucleus density of each emitter layer and the incident neutron flux considering the self-shielding effect are given by the previous iteration. Since the neutron flux calculated in each iteration step is relatively low, the influence of neutron self-shielding changes in each iteration step can be ignored. Furthermore, the emitter is divided into several layers, each approximately a few micrometers thick, and it is assumed that the atomic nucleus density of each layer is uniformly distributed. For a given emitter layer, the incident neutron flux needs to consider the influence of the self-shielding effect of the external emitter layers. This neutron flux is used to calculate the emitter atomic nucleus density at the end of the current iteration time, ultimately achieving the calculation of the number of remaining atomic nuclei in the emitter for different service lengths.

[0060] Specifically, this embodiment takes a rhodium self-energized neutron detector as an example, in which rhodium-103 accounts for 100% of the emitter, and its structure is as follows. Figure 1 As shown, it consists of three parts: an emitter, an insulating layer, and a collector. The emitter has a diameter of 2 mm and a nuclear density of 0.7188 × 10⁻⁶. 23 n / cm 3 Assume the neutron flux (all thermal neutrons) density of the neutron field where the detector is located is 1 × 10⁻⁶. 13 n / cm 2 ·s -1 The reaction cross section between rhodium-103 and thermal neutrons is 146 Å. The thermal neutrons and detector operated continuously in this neutron field for 600 months, receiving a cumulative neutron fluence of 1.55 × 10⁻⁶. 22 .

[0061] The analysis and calculation tool used in this embodiment is MATLAB.

[0062] The specific steps for analysis are as follows:

[0063] The cumulative neutron fluence over 600 months is divided into 60,000,000 generations, meaning the number of incident neutrons in each generation is 2.6 × 10⁻⁶. 14 All neutrons are incident perpendicularly along the outer surface of the emitter; the emitter is divided into 100 layers, each 10 micrometers thick. N(m,i) represents the atomic nucleus density distribution of the m-th emitter layer after irradiation by the i-th generation of neutrons, and the initial atomic nucleus density of each emitter layer is N0 = 0.7188 × 10⁻⁶. 23 n / cm 3I(m) represents the neutron fluence required for burnup calculation of the m-th emitter (this fluence takes into account the self-shielding effect of each emitter layer and the current emitter layer). The incident neutron fluence on the outer surface of the emitter in each iteration is 2.6 × 10⁻⁶. 14 This embodiment is implemented using MATLAB. For the m-th layer emitter, the incident neutron fluence is A(m), and the transmitted neutron fluence is A(m+1). For the i-th iteration calculation, then:

[0064] The neutron flux required for the i-th iteration is: I = (A(m+1) + A(m)) / 2;

[0065] Then, after the i-th iteration, the atomic nucleus density of the m-th emitter is:

[0066] The above steps allow for iterative calculation of the nuclear density distribution of each emitter layer under different cumulative neutron fluence conditions.

[0067] Therefore, after i-generation neutron irradiation, the burnup rate of the self-powered neutron detector is:

[0068] like Figure 2 The figure shows the distribution of burnup of the self-powered neutron detector emitter over its service life (in months), calculated using the method of the present invention. For comparison, Figure 2 The paper also presents burnup analysis results without stratifying the emitter, using a fixed self-shielding factor. For a rhodium self-powered neutron detector with an emitter diameter of 2 mm, the self-shielding factor is 0.733. It can be observed that the difference between the two burnup analysis methods gradually increases with the service life of the self-powered neutron detector, reaching a deviation of 10% after 600 months of service. Furthermore, stratifying the emitter yields a higher burnup rate. This is because as the emitter burns down, the density of rhodium-103 nuclei decreases, weakening the self-shielding absorption of neutrons by the outer emitter material and reducing the self-shielding effect. The burnup analysis method with a fixed self-shielding factor of 0.733 does not consider this effect, leading to increased error.

[0069] The above analysis shows that the self-powered neutron detector emitter burnup lifetime analysis method proposed in this invention can more accurately assess the burnup lifetime of a self-powered neutron detector during long-term service in a reactor, providing a more accurate and substantial basis for the design and replacement of self-powered neutron detectors.

[0070] Example 2

[0071] The difference between this embodiment and Embodiment 1 is that this embodiment provides a computer device, including a memory, a processor, and a computer program stored in the memory and executable on the processor. When the processor executes the computer program, it implements the above-described method for analyzing the burnup lifetime of a self-powered neutron detector emitter.

[0072] Meanwhile, the present invention also provides a computer-readable storage medium storing a computer program, which, when executed by a processor, implements the above-described method for analyzing the burnup lifetime of a self-powered neutron detector emitter.

[0073] Those skilled in the art will understand that embodiments of this application can be provided as methods, systems, or computer program products. Therefore, this application can take the form of a completely hardware embodiment, a completely software embodiment, or an embodiment combining software and hardware aspects. Furthermore, this application can take the form of a computer program product embodied on one or more computer-usable storage media (including but not limited to disk storage, CD-ROM, optical storage, etc.) containing computer-usable program code.

[0074] This application is described with reference to flowchart illustrations and / or block diagrams of methods, apparatus (systems), and computer program products according to embodiments of this application. It will be understood that each block of the flowchart illustrations and / or block diagrams, and combinations of blocks in the flowchart illustrations and / or block diagrams, can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, special-purpose computer, embedded processor, or other programmable data processing apparatus to produce a machine, such that the instructions, which execute via the processor of the computer or other programmable data processing apparatus, generate instructions for implementing the flowchart... Figure 1 One or more processes and / or boxes Figure 1 A device that provides the functions specified in one or more boxes.

[0075] These computer program instructions may also be stored in a computer-readable storage medium that can direct a computer or other programmable data processing device to function in a particular manner, such that the instructions stored in the computer-readable storage medium produce an article of manufacture including instruction means, which are implemented in a process Figure 1 One or more processes and / or boxes Figure 1 The function specified in one or more boxes.

[0076] These computer program instructions may also be loaded onto a computer or other programmable data processing equipment to cause a series of operational steps to be performed on the computer or other programmable equipment to produce a computer-implemented process, thereby providing instructions that execute on the computer or other programmable equipment for implementing the process. Figure 1One or more processes and / or boxes Figure 1 The steps of the function specified in one or more boxes.

[0077] The specific embodiments described above further illustrate the purpose, technical solution, and beneficial effects of the present invention. It should be understood that the above description is only a specific embodiment of the present invention and is not intended to limit the scope of protection of the present invention. Any modifications, equivalent substitutions, improvements, etc., made within the spirit and principles of the present invention should be included within the scope of protection of the present invention.

Claims

1. A method for analyzing the burnup lifetime of a self-powered neutron detector emitter, characterized in that, The method includes: The total neutron fluence absorbed by the self-powered neutron detector during its entire service life is processed in generations, and the emitter of the self-powered neutron detector is processed in layers. Burnup analysis and calculation are performed for each generation of neutron fluence and each layer of emitter. In the process of calculating the neutron fluence of each generation, the atomic nucleus density of each layer of emitter at different service periods is calculated iteratively by the burnup calculation formula to obtain the burnup distribution information of the emitter under the total neutron fluence condition during the service period. In the process of fuel consumption iterative calculation using the fuel consumption formula, for a certain iteration step, the atomic nucleus density of each layer of the emitter and the incident neutron fluence considering the self-shielding effect are obtained from the previous iteration calculation process.

2. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 1, characterized in that, During each generation of neutron flux calculation, the neutron self-shielding factor within each emitter layer remains constant, while the neutron self-shielding factor differs between different emitter layers, and the neutron self-shielding factor is continuously updated as the neutron generation changes.

3. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 1, characterized in that, The total neutron fluence absorbed by the self-powered neutron detector during its entire service life is processed in generations, including: The total neutron flux absorbed by the self-powered neutron detector during its entire service life is divided into several generations. The number of nucleons in the emitter decreases due to the nuclear reaction between the neutrons and the emitter in each generation. That is, the effect of burnup on the reaction rate is not considered in each generation.

4. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 1, characterized in that, The emitter of a self-powered neutron detector undergoes layered processing, including: The material of the self-powered neutron detector is layered, with each layer having a thickness on the order of micrometers. Within each layer, the self-shielding effect of the material on the incident neutron is assumed to remain unchanged. The incident neutron fluence required for burnout calculation of each emitter layer takes into account the absorption of the initial incident neutron fluence by the preceding layers, i.e., the influence of the self-shielding effect added between the emitter materials of each layer.

5. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 4, characterized in that, The formula for calculating the burnup of each launcher layer is as follows: N(t)=N0exp -Iσt I = (A + A') / 2 A′=Aexp (-N×d×σ) In the formula, N(t) is the burnup of each emitter layer; N0 is the initial number of emitter nucleons; I is the incident neutron flux required for burnup calculation of each emitter layer in each iteration step; σ is the neutron reaction cross section; t is time; A is the parallel incident neutron flux intensity; A′ is the flux intensity of the neutron flux after penetrating the emitter material of this layer, i.e., A of the next layer; N is the atomic nucleus surface density of the emitter material; d is the emitter layer thickness; σ is the cross section where neutrons react with the emitter nuclei.

6. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 1, characterized in that, The burnup distribution of each emitter layer under each generation's neutron flux condition is analyzed using the burnup calculation formula for each generation. The burnup calculation formula for each generation is as follows: N(i)=N0exp -Iσi N(i+1)=N(i)exp -Iσ In the formula, N(i) is the atomic nucleus density of the emitter material in a certain emitter layer, i is the neutron generation; N0 is the initial number of nucleons in the emitter; I is the incident neutron fluence required for burnup calculation of each emitter layer in each iteration step; σ is the neutron reaction cross section; N(i+1) is the atomic nucleus density of the emitter material in a certain emitter layer, i+1 generation; and a is the number of incident neutrons in each generation.

7. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 6, characterized in that, The change in atomic nucleus density of each emitter layer material with the cumulative neutron flux is calculated iteratively according to the N(i+1) calculation formula until the total neutron flux is reached, thus completing the entire burnup analysis process of the self-powered neutron detector.

8. The method for analyzing the burnup lifetime of a self-powered neutron detector emitter according to claim 1, characterized in that, This method is applied to the emitters of self-powered neutron detectors used in reactors, including... 103 Rh、 50 V. 59 Co, Pt, and Hf materials.

9. A computer device comprising a memory, a processor, and a computer program stored in the memory and executable on the processor, characterized in that, When the processor executes the computer program, it implements a method for analyzing the burnup lifetime of a self-powered neutron detector emitter as described in any one of claims 1 to 8.

10. A computer-readable storage medium storing a computer program, characterized in that, When the computer program is executed by the processor, it implements a method for analyzing the burnup lifetime of a self-powered neutron detector emitter as described in any one of claims 1 to 8.

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