A method, device, equipment and storage medium for determining neutron source intensity

Abstract

Embodiments of the present invention disclose a method, apparatus, device, and storage medium for determining neutron source intensity. The method includes: determining the source intensity of a primary neutron source as it changes over time after a reactor shutdown; determining the source intensity of a secondary neutron source as it changes over time after a reactor shutdown based on the determined thermal neutron flux density of a secondary neutron source; and determining the source intensity of a total neutron source as it changes over shutdown time based on the source intensity of the primary and secondary neutron sources, combined with predetermined spatial and energy spectrum factors, to ensure safe loading and startup of the unit. This method comprehensively considers the different locations and energy spectra of the primary and secondary neutron sources, which have different effects on the response of the off-core detectors, and combines the predetermined spatial and energy spectrum factors to determine the source intensity of the neutron source as it changes over time after a reactor shutdown, thereby achieving rapid and accurate determination of the neutron source intensity and providing a physical basis for reactor startup.

Description

Technical Field

[0001] The present invention relates to the field of reactor testing, and in particular to a method, device, equipment and storage medium for determining neutron source intensity. Background Art

[0002] To ensure criticality safety, nuclear power plant reactor loading and startup require effective monitoring by neutron detectors throughout the entire process. Before initial operation or after extended shutdown, the core is low in neutrons during loading and startup, making it difficult for external detectors to detect the neutron fluence rate within the core. To address this, neutron source assemblies are typically installed within the core. These neutrons, after subcritical multiplication, generate a sufficient number of neutrons for the neutron detectors to detect the neutron levels within the core, overcoming the measurement blind spot.

[0003] Generally speaking, neutron sources are divided into primary neutron sources and secondary neutron sources. They are one of the nuclear equipment in power plants and play an important role in avoiding blind spots in core monitoring. Primary neutron sources are mainly used in the first cycle of the reactor, while secondary neutron sources are used in subsequent cycles of the reactor. Insufficient source strength will cause risks in core loading and startup, posing a challenge to nuclear safety. Existing technologies have more calculations for separate primary and secondary neutron sources, but primary and secondary neutron sources have different energy spectra and locations, and their impact on the response of external detectors is also different. In addition, the method for determining the source strength after the overhaul window is not accurate enough, and it is impossible to understand the changes in the neutron source strength after the overhaul, which will affect the purpose of ensuring the safe loading and startup of the unit. Summary of the Invention

[0004] The embodiments of the present invention provide a method, device, equipment and storage medium for determining the source intensity of a neutron source, so as to achieve rapid and accurate determination of the source intensity of a neutron source and provide a certain physical basis for reactor startup.

[0005] In a first aspect, this embodiment provides a method for determining neutron source intensity, the method comprising:

[0006] Determine the source intensity of the primary neutron source as a function of time after reactor shutdown;

[0007] determining a source intensity of a secondary neutron source that varies with time after a reactor shutdown based on a determined thermal neutron flux density at a secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;

[0008] Based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, the source strength of the total neutron source that varies with the shutdown time is determined to ensure safe loading and startup of the unit, wherein the total neutron source includes a primary neutron source and a secondary neutron source.

[0009] In a second aspect, this embodiment provides a neutron source intensity determination device, the device comprising:

[0010] A primary source strength determination module is used to determine the source strength of the primary neutron source that changes with time after the reactor is shut down;

[0011] A secondary source strength determination module is used to determine the source strength of the secondary neutron source that changes with time after the reactor is shut down based on the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;

[0012] The total source strength determination module is used to determine the source strength of the total neutron source that varies with the shutdown time based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, so as to ensure safe loading and startup of the unit, wherein the total neutron source includes a primary neutron source and a secondary neutron source.

[0013] In a third aspect, this embodiment provides an electronic device, comprising:

[0014] at least one processor; and

[0015] a memory communicatively connected to the at least one processor; wherein,

[0016] The memory stores a computer program executable by the at least one processor, and the computer program is executed by the at least one processor so that the at least one processor can execute the neutron source intensity determination method described in any embodiment of the present invention.

[0017] In a fourth aspect, this embodiment provides a computer-readable storage medium, wherein the computer-readable storage medium stores computer instructions, and the computer instructions are used to enable a processor to implement the neutron source intensity determination method described in any embodiment of the present invention when executed.

[0018] An embodiment of the present invention discloses a method, device, equipment and storage medium for determining the source intensity of a neutron source. The method comprises: determining the source intensity of a primary neutron source that varies with time after a reactor shutdown; determining the source intensity of a secondary neutron source that varies with time after a reactor shutdown based on a determined thermal neutron flux density where a secondary neutron source is located, wherein the secondary neutron source is an antimony-beryllium source; and determining the source intensity of a total neutron source that varies with shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source, in combination with predetermined spatial influencing factors and energy spectrum influencing factors, to ensure safe loading and startup of the unit, wherein the total neutron source includes a primary neutron source and a secondary neutron source. The above technical solution comprehensively considers the different positions and energy spectra of the primary neutron source and the secondary neutron source, and their different impacts on the external detector response. Combined with the predetermined spatial influencing factors and energy spectrum influencing factors, it determines the source intensity of the antimony-beryllium source that changes with time after the reactor shutdown, thereby achieving rapid and accurate determination of the neutron source intensity and providing a certain physical basis for reactor startup.

[0019] It should be understood that the content described in this section is not intended to identify the key or important features of the embodiments of the present invention, nor is it intended to limit the scope of the present invention. Other features of the present invention will become readily understood through the following description. BRIEF DESCRIPTION OF THE DRAWINGS

[0020] In order to more clearly illustrate the technical solutions in the embodiments of the present invention, the following briefly introduces the drawings required for use in the description of the embodiments. Obviously, the drawings described below are only some embodiments of the present invention. For ordinary technicians in this field, other drawings can be obtained based on these drawings without creative work.

[0021] Figure 1 A flow chart of a method for determining neutron source intensity provided in Example 1 of the present invention;

[0022] Figure 1a is a schematic diagram of a cross section of an antimony-beryllium source;

[0023] Figure 2 A schematic structural diagram of a neutron source intensity determination device provided in a second embodiment of the present invention;

[0024] Figure 3 This is a structural diagram of an electronic device provided in Example 3 of the present invention. DETAILED DESCRIPTION

[0025] In order to enable those skilled in the art to better understand the solutions of the present invention, the technical solutions in the embodiments of the present invention will be clearly and completely described below in conjunction with the drawings in the embodiments of the present invention. Obviously, the embodiments described are only part of the embodiments of the present invention, not all of the embodiments. Based on the embodiments of the present invention, all other embodiments obtained by ordinary technicians in this field without making creative efforts should fall within the scope of protection of the present invention.

[0026] It should be noted that the terms "original", "target", etc. in the description and claims of the present invention and the above-mentioned drawings are used to distinguish similar objects and are not necessarily used to describe a specific order or sequence. It should be understood that the numbers used in this way can be interchanged where appropriate so that the embodiments of the present invention described herein can be implemented in an order other than those illustrated or described herein. In addition, the terms "including" and "having" and any variations thereof are intended to cover non-exclusive inclusions. For example, a process, method, system, product or device that includes a series of steps or units is not necessarily limited to those steps or units that are clearly listed, but may include other steps or units that are not clearly listed or that are inherent to these processes, methods, products or devices.

[0027] Example 1

[0028] Figure 1 This is a flow chart of a method for determining a neutron source intensity provided in Example 1 of the present invention. This method is applicable to situations where the neutron source intensity is calculated after a nuclear reactor is shut down. This method can be performed by a neutron source intensity determination device. The neutron source intensity determination device can be implemented in the form of hardware and / or software, and the device can be configured in an electronic device.

[0029] The current mainstream secondary neutron source is the antimony-beryllium source, which is formed by cold pressing a uniform mixture of antimony powder and beryllium powder. The antimony-beryllium neutron source does not produce neutrons at the beginning, and only becomes a neutron source after being irradiated by neutrons in the reactor. The working principle of the secondary neutron source is to generate products that can spontaneously fission through irradiation. The reaction process is: 123 Sb is generated after neutron irradiation in the reactor 124 Sb, 124 Sb emits gamma rays during the decay process, and its half-life is 60.2 days. 9 Be produces a (γ, n) reaction to generate neutrons, and the reaction threshold energy is 1.67MeV. The reaction formula is:

[0030]

[0031]

[0032] Considering that neutron intensity decreases 60 days after irradiation cessation, for example, if a nuclear power plant's outage for overhaul is extended to 60 or 90 days due to insufficient grid demand, the antimony-beryllium neutron source intensity may be insufficient upon restart. Therefore, a method is needed to accurately calculate the temporal variation of neutron source intensity after a reactor shutdown.

[0033] like Figure 1 As shown, the method for determining the neutron source intensity provided in the first embodiment may specifically include the following steps:

[0034] S110. Determine the source intensity of the primary neutron source that changes with time after the reactor is shut down.

[0035] In the analysis, it is assumed that the antimony-beryllium source is made of natural antimony (including 57.2% of 121 Sb and 42.8% 123 Sb), covered with a layer of metal beryllium tube. Figure 1a is a schematic diagram of the cross section of the antimony-beryllium source, as shown in Figure 1a As shown, it is assumed that the outer diameter of the beryllium tube is r2, the inner diameter is r1 (close to the antimony rod), and the radius of the antimony rod is r1.

[0036] It is understood that the primary neutron source is primarily used in the first reactor cycle, while the secondary neutron source is used in subsequent reactor cycles. Considering that the source strength of the primary neutron source decays over time during the subsequent reactor cycles, its source strength does not completely disappear. Therefore, compared to the prior art, which requires more calculations for the primary and secondary neutron sources separately, this embodiment comprehensively considers the source strength of both the primary and secondary neutron sources. This step is used to determine the source strength of the primary neutron source after reactor shutdown.

[0037] Specifically, after the reactor is shut down, the source intensity of the primary neutron source that changes with time can be expressed as: Among them, N0 is the initial source intensity of the primary neutron source, λ f is the decay constant, t1 is the in-pile irradiation time, and t2 is the time after leaving the pile. It is understandable that when the initial source intensity of the primary neutron source, the in-pile irradiation time, and the time after leaving the pile are known, the source intensity of the primary neutron source at the corresponding moment can be determined.

[0038] S120. Determine the source intensity of the secondary neutron source that changes with time after reactor shutdown based on the determined thermal neutron flux density of the secondary neutron source, where the secondary neutron source is an antimony-beryllium source.

[0039] In this embodiment, the secondary neutron source is used in the subsequent cycle of the reactor. The neutron flux density refers to the number of neutrons passing through a unit area perpendicular to the direction of neutron movement per unit time.

[0040] Specifically, let the thermal neutron flux density of the antimony-beryllium source be φ, 123 The average macroscopic absorption cross section of thermal neutrons of Sb is Σ a , 124 The nuclear density of Sb is N, then when neutron irradiated, N satisfies the following differential equation: Among them, λ is 124 The decay constant of Sb is λ = 1.332 × 10 -7 s -1 .

[0041] Let t1 be the irradiation time of the antimony rod in the pile, and t2 be the time after the rod is taken out of the pile. In this embodiment, to keep the unit consistent with the subsequent power plant operation time, t1 and t2 are in days, and assume that N = 0 when t = 0. The solution of the equation is:

[0042] Accordingly, 124 The γ activity of Sb per unit volume is:

[0043] Further, according to 124 The gamma energy spectrum of Sb decay shows that 54% of the gamma rays have energies above the (γ, n) reaction threshold energy of beryllium-9, 50% of which are at 1.69 MeV (MeV is the unit of energy) and 4% are at 2.06 MeV. Let r1 be the radius of the antimony rod, r2 be the outer diameter of the Be tube, and r1 be the inner diameter (closely attached to the Sb rod). If we ignore the self-absorption of gamma rays by the antimony rod, the gamma flux density on the antimony rod surface that exceeds the (γ, n) reaction threshold energy of beryllium-9 is:

[0044]

[0045] Simplifying it, we get:

[0046]

[0047] If the absorption of γ by the beryllium tube is neglected, the γ flux density in the beryllium tube can be expressed as:

[0048]

[0049] Furthermore, the average γ flux density in the beryllium tube can be expressed as:

[0050] Where φ(r) is the gamma flux density in the beryllium tube, the radius of the antimony rod is r1, the outer diameter of the Be tube is r2, and the inner diameter is r1 (close to the Sb rod).

[0051] Substituting the expression of φ(r1) in the above step into the above formula, the average γ flux density in the beryllium tube can be obtained as: Where φ represents the thermal neutron flux density where the antimony-beryllium source is located, and λ represents 124 The decay constant of Sb, r1 represents the radius of the antimony rod, r2 represents the outer diameter of the beryllium tube, r1 represents the inner diameter of the beryllium tube (close to the Sb rod), and t1 represents the irradiation time of the antimony rod in the stack.

[0052] Furthermore, let the (γ, n) reaction cross section of beryllium-9 be Σ γ,n , and substituting the above expression of the average γ flux density in the beryllium tube into it, the neutron emission rate (i.e., specific intensity) per unit length of the beryllium tube is:

[0053] Where φ represents the thermal neutron flux density where the antimony-beryllium source is located, Σ a express 123 The average macroscopic absorption cross section of thermal neutrons of Sb, λ represents 124 The decay constant of Sb, r1 represents the radius of the antimony rod, r2 represents the outer diameter of the beryllium tube, r1 represents the inner diameter of the beryllium tube (close to the Sb rod), and t1 represents the irradiation time of the antimony rod in the stack.

[0054] It is understandable that when the proportional relationship between r1 and r2 and Σ a The specific intensity of the secondary neutron source can be determined. Based on the specific intensity of the secondary neutron source, the length of the secondary neutron source, and the number of secondary neutron sources, the source intensity of the secondary neutron source that changes with time after the shutdown of the reactor can be determined.

[0055] Optionally, based on the thermal neutron flux density where the secondary neutron source is located, a method for determining the source intensity of the secondary neutron source that changes with time after the reactor shutdown may be:

[0056] a1) Determine the specific intensity of the secondary neutron source based on the thermal neutron flux density at the location of the secondary neutron source and a predetermined target relationship between the outer diameter and the inner diameter of the beryllium tube.

[0057] Among them, when the relationship between the outer diameter and the inner diameter of the beryllium tube is the target relationship, the specific intensity of the corresponding secondary neutron source is the maximum value.

[0058] In this embodiment, the ratio of the outer diameter to the inner diameter can be determined based on the predetermined target relationship between the outer diameter and the inner diameter of the beryllium tube. That is, the inner diameter of the beryllium tube is represented by the outer diameter r2. Then, the expression of r1 is substituted into the formula of the neutron emission rate (i.e., specific intensity) per unit length of the beryllium tube. The maximum value of the specific intensity of the secondary neutron source is expressed as:

[0059] Among them, the reaction cross section of beryllium-9 in this embodiment Σ γ,n The value can be 6.2·10 -3 m -1 ,in addition,123 The average macroscopic absorption cross section Σ for thermal neutrons of Sb a It can be calculated as follows: the density of antimony is known to be ρ = 6.69 × 10 3 kg / m 3 , atomic weight A=121.75, where 123 The abundance of Sb (atomic percentage) ω = 43%, 123 The average microscopic absorption cross section σ of thermal neutrons of Sb a =1.63b, then:

[0060]

[0061] Will 123 The average macroscopic absorption cross section of thermal neutrons of Sb is substituted into the formula of the maximum value of the specific intensity of the secondary neutron source. Since different times are involved, Sb is replaced by t2 when considering t2. max (t1) is treated as a constant and then the natural decay law in physics is introduced to obtain:

[0062] Where, the radius of the antimony rod is r1, the outer diameter of the Be tube is r2, and the inner diameter is r1 (close to the Sb rod). The units of r1 and r2 are m, and φ is the thermal neutron flux density in m. -2 / s, the unit of t1 is day, and λ is 124 The decay constant of Sb, t1 represents the irradiation time of the antimony rod in the pile, and t2 is the time after it is taken out of the pile.

[0063] b1) Determine the source intensity of the secondary neutron source that changes with time after shutdown based on the specific intensity, length and number of the secondary neutron source.

[0064] Specifically, the intensity of the antimony-beryllium source is obtained by multiplying the specific intensity of the antimony-beryllium source by the length of the source. If there are several antimony-beryllium sources installed in the core, the total source intensity of the secondary neutron source is equal to the sum of the source intensities of each antimony-beryllium source.

[0065] S130. Based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, determine the source strength of the total neutron source that changes with the shutdown time to ensure safe loading and startup of the unit.

[0066] Among them, the total neutron source includes primary neutron source and secondary neutron source.

[0067] In this embodiment, a spatial influence factor k1 is introduced to account for the varying out-of-core responses when the neutron source is located at different locations. A spectrum factor k2 is added to the secondary neutron source intensity to account for the different triggering mechanisms of the secondary and primary neutron sources, resulting in different neutron penetration capabilities. The influence factors k1 and k2 can be calculated to give theoretical values ​​for use in the analysis of the first-in-the-cycle neutron source intensity. They can also be calibrated and determined based on actual engineering test measurement results at the beginning of different cycles.

[0068] Specifically, after considering the neutron source space influence factor and energy spectrum influence factor, the total neutron source intensity is expressed as: S T =k1[S f +k2S(t1,t2)], where S f is the source intensity of the primary neutron source, and S(t1,t2) is the source intensity of the secondary neutron source.

[0069] An embodiment of the present invention discloses a method for determining neutron source intensity, comprising: determining the intensity of a primary neutron source that varies with time after a reactor shutdown; determining the intensity of a secondary neutron source that varies with time after a reactor shutdown based on the determined thermal neutron flux density of a secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source; and determining the intensity of a total neutron source that varies with shutdown time based on the intensity of the primary neutron source and the intensity of the secondary neutron source, combined with predetermined spatial influence factors and energy spectrum influence factors, to ensure safe loading and startup of the unit, wherein the total neutron source includes the primary neutron source and the secondary neutron source. The above technical solution comprehensively considers the different positions and energy spectra of the primary and secondary neutron sources, and their different effects on external detector responses, and combines predetermined spatial influence factors and energy spectrum influence factors to determine the intensity of the antimony-beryllium source that varies with time after a reactor shutdown, thereby achieving rapid and accurate determination of the neutron source intensity and providing a certain physical basis for reactor startup.

[0070] As a first optional embodiment of the embodiment of the present invention, this first optional embodiment further defines that the method further includes: according to the neutron irradiation time 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the Sb neutron macroscopic absorption cross-sectional area and the target relationship between the outer diameter and inner diameter of the beryllium tube corresponding to the maximum value of the secondary neutron source intensity is determined.

[0071] Specifically, according to the neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the macroscopic absorption cross-section of Sb neutrons and the 124The gamma activity of Sb per unit volume. Further, based on the gamma activity, the gamma flux density on the antimony rod surface that exceeds the beryllium-9 reaction threshold energy is determined. Based on the gamma flux density, the average gamma flux density within the beryllium tube is determined. Furthermore, based on the average gamma flux density, the specific intensity per unit length of the beryllium tube is determined. Finally, the target relationship between the outer and inner diameters of the beryllium tube corresponding to the maximum specific intensity is determined.

[0072] Furthermore, according to the neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The steps for determining the target relationship between the outer diameter and inner diameter of the beryllium tube corresponding to the maximum value of the secondary neutron source intensity can be expressed as follows:

[0073] a2) According to neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the macroscopic absorption cross-section of Sb neutrons and the 124 The gamma activity of Sb per unit volume.

[0074] Specifically, let the thermal neutron flux density of the antimony-beryllium source be φ, 123 The average macroscopic absorption cross section of thermal neutrons of Sb is Σ a , 124 The nuclear density of Sb is N, then when neutron irradiated, N satisfies the following differential equation: Among them, λ is 124 The decay constant of Sb is λ = 1.332 × 10 -7 s -1 .

[0075] Let t1 be the irradiation time of the antimony rod in the pile, and t2 be the time after the rod is taken out of the pile. In this embodiment, to keep the unit consistent with the subsequent power plant operation time, t1 and t2 are in days, and assume that N = 0 when t = 0. The solution of the equation is:

[0076] Accordingly, 124 The γ activity of Sb per unit volume is:

[0077] b2) Based on the gamma activity, determine the gamma flux density on the surface of the antimony rod that exceeds the threshold energy of the beryllium-9 reaction.

[0078] Specifically, according to 124The gamma energy spectrum of Sb decay shows that 54% of the gamma rays have energies above the Be-9 (γ, n) reaction threshold energy. Of these, 50% are at 1.69 MeV (MeV is the unit of energy) and 4% are at 2.06 MeV. Let r1 be the radius of the antimony rod, r2 be the outer diameter of the beryllium tube, and r1 be the inner diameter (closely attached to the Sb rod). If we ignore the antimony rod's self-absorption of gamma rays, the gamma flux density on the antimony rod surface that exceeds the Be-9 (γ, n) reaction threshold energy is:

[0079]

[0080] Simplifying it, we get:

[0081]

[0082] Assuming the outer diameter of the Be tube is r2 and the inner diameter is r1 (close to the Sb rod), if the absorption of γ by the Be tube is ignored, the γ flux density in the Be tube can be expressed as:

[0083]

[0084] c2) Determine the average gamma flux density in the beryllium tube based on the gamma flux density.

[0085] Specifically, the average γ flux density in the Be tube can be expressed as:

[0086] Where φ(r) is the gamma flux density in the beryllium tube, the radius of the antimony rod is r1, the outer diameter of the Be tube is r2, and the inner diameter is r1 (close to the Sb rod).

[0087] Substituting the expression of φ(r1) in the above step into the above formula, the average γ flux density in the beryllium tube can be obtained as: Where φ represents the thermal neutron flux density where the antimony-beryllium source is located, and λ represents 124 The decay constant of Sb, r1 represents the radius of the antimony rod, r2 represents the outer diameter of the beryllium tube, r1 represents the inner diameter of the beryllium tube (close to the Sb rod), and t1 represents the irradiation time of the antimony rod in the stack.

[0088] d2) Determine the specific intensity per unit length of the beryllium tube based on the average gamma flux density.

[0089] Specifically, let the (γ, n) reaction cross section of Be-9 be Σ γ,n , and substituting the above expression of the average γ flux density in the beryllium tube into it, the neutron emission rate (i.e., specific intensity) per unit length of the beryllium tube is:

[0090] Where φ represents the thermal neutron flux density where the antimony-beryllium source is located, and λ represents 124The decay constant of Sb, r1 represents the radius of the antimony rod, r2 represents the outer diameter of the beryllium tube, r1 represents the inner diameter of the beryllium tube (close to the Sb rod), and t1 represents the irradiation time of the antimony rod in the stack.

[0091] e2) Determine the target relationship between the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum specific strength.

[0092] In this embodiment, it is assumed that the radius of the antimony rod and the beryllium tube are pressed in an optimal ratio. Specifically, if S(t1) is regarded as a function of the radius r1 of the Sb rod, in order to obtain the maximum value of S(t1), we can let It is easy to get the condition for taking extreme values: It can be understood that the target relationship between the outer diameter and inner diameter of the beryllium tube corresponding to the maximum specific strength can be expressed as: Substituting this relationship into the above expression for the specific strength per unit length of the beryllium tube, we can obtain:

[0093] λ represents 124 The decay constant of Sb, r2 represents the outer diameter of the beryllium tube, t1 represents the irradiation time of the antimony rod in the pile, φ represents the thermal neutron flux density of the antimony rod, Σ a express 123 The average macroscopic absorption cross section of thermal neutrons of Sb, Σ γ,n represents the (γ, n) reaction cross section of beryllium-9.

[0094] In this embodiment, the thermal neutron flux density of the antimony rod is φ, which can be taken as the average thermal neutron flux density of the reactor core for ease of calculation. If the average thermal neutron flux density during rated power operation is φ H , then φ can be approximately calculated as follows:

[0095] φ=xφ H For simplicity, it is assumed that the core is operating at full power, that is, x = 1. Then, according to the above formula, the specific intensity of the antimony-beryllium source is expressed as Determine the secondary neutron source strength at a certain moment, and then just calculate φ at the corresponding moment.

[0096] As a second optional embodiment of the embodiment of the present invention, this second optional embodiment further defines that the step of determining the thermal neutron flux density where the secondary neutron source is located includes:

[0097] a3) Based on the nuclear power of the reactor core, determine the number of uranium-235 atoms per unit volume after the reactor has been in operation for a certain period of time.

[0098] Specifically, in this embodiment, the enrichment of uranium-235 after the reactor's operating time is determined can be determined based on the relationship between the initial uranium-235 mass, reactor core power, and operating factors. Furthermore, the uranium-235 enrichment can be determined based on the conversion relationship between negative enrichment and abundance. Furthermore, the number of uranium dioxide molecules per unit volume can be determined based on the uranium-235 abundance. Finally, the number of uranium-235 atoms per unit volume can be determined based on the number of uranium dioxide molecules.

[0099] Furthermore, based on the reactor core nuclear power and the reactor operating time, the steps for determining the number of uranium-235 atoms per unit volume after the reactor operating time can be expressed as:

[0100] a31) Determine the uranium-235 enrichment after the reactor has been in operation for a period of time based on the initial uranium-235 mass, reactor core power, and operating factors.

[0101] In this embodiment, in order to more realistically simulate the uranium-235 consumed by the operation of the power plant, the power plant load factor is taken into account when calculating the enrichment or abundance of the remaining nuclear material uranium-235 at a certain stage.

[0102] Considering that the primary neutron source still has a high activity for a period of time after the first cycle, the total source intensity needs to be calculated without considering the difference in the energy spectrum between the primary and secondary neutron sources. Let the initial uranium mass be m1, the initial enrichment be ε1, the number of burning days be t (unit: day), and the uranium dioxide density ρ = 10.42×10 3 kg / m 3 , the operating factor is 0.93, assuming the core power P = 4.04×10 9 W, uranium-235 fission cross section σ f =583.5b (b is the unit of fission cross section). According to the practical value, the mass of uranium-235 consumed for each 1 megawatt-day is 0.00123kg. The mass of uranium-235 consumed after t days of operation can be expressed as:

[0103] Δm=Pt×0.00123×0.93÷10 6 =Pt×1.1439×10 -9 (kg).

[0104] The enrichment of uranium-235 at day t can be expressed as:

[0105]

[0106] a32) Determine the abundance of uranium-235 based on its enrichment.

[0107] Specifically, according to the conversion relationship between abundance and enrichment, the uranium-235 abundance at day t can be expressed as:

[0108] a33) Based on the uranium-235 abundance, determine the number of uranium dioxide molecules per unit volume after the operating time.

[0109] The molecular weight of uranium dioxide after running for t days is:

[0110] Here, c represents the abundance of uranium-235.

[0111] Then the number of uranium dioxide molecules per unit volume at day t is:

[0112]

[0113] in, represents the density of uranium dioxide, and N0 represents Avogadro's constant, which is a fixed value.

[0114] a34) Based on the number of molecules of uranium dioxide, determine the number of uranium-235 atoms per unit volume after the run time.

[0115] Then the number of uranium-235 atoms per unit volume at day t is:

[0116]

[0117] Substituting the abundance expression into the equation, we obtain:

[0118]

[0119] Furthermore, substituting the enrichment expression into the equation, we can obtain the number of uranium-235 atoms per unit volume at day t:

[0120] In the formula, the meanings of the letters are the same as above and will not be repeated here.

[0121] b3) Determine the macroscopic cross-sectional area of ​​uranium-235 after the reactor has been in operation, based on the number of uranium-235 atoms per unit volume.

[0122] When a reactor operates at full power, the uranium-235 in the fuel assemblies is continuously consumed, while the reactor's output power remains constant. This necessitates maintaining the core's reactivity through methods such as rod extraction and boron dilution. Furthermore, the mass of uranium-235 consumed daily under full power operation is fixed. Therefore, the remaining mass of uranium-235 can be used to derive its macroscopic cross-section, and thus the average thermal neutron flux density required to meet full power. During normal operation, a nuclear power plant typically operates at full power. After the plant is built, the core's shape and volume are fixed, so the limiting factor for the average neutron flux density is the uranium's macroscopic absorption cross-section.

[0123] Specifically, the fission reaction in the reactor is mainly produced by uranium-235 absorbing thermal neutrons with energy E = 0.0253 eV. At this time, the microscopic absorption cross section of uranium-235 is: f =583.5b (b is the unit of cross section).

[0124] For uranium-235, the neutron yield per fission is 2.416, and the capture fission ratio is 0.169. Based on the fuel enrichment at the beginning of its life, the initial macroscopic fission cross section can be calculated: Σ f =Nσ f , where N is the number of uranium-235 atoms per unit volume.

[0125] Calculating the macroscopic fission cross section at the end of the life cycle is somewhat complex. It requires calculating the number of fissions required based on the energy released, which in turn derives the mass of nuclear material consumed and the enrichment or abundance of uranium-235 remaining at the end of the life cycle, ultimately yielding the macroscopic fission cross section at the end of the life cycle. Similarly, the macroscopic cross section of uranium-235 remaining at a certain point in the cycle life cycle can be calculated. Based on the number of uranium-235 atoms per unit volume at day t and the microscopic absorption cross section of uranium-235, the macroscopic cross section of uranium-235 at day t can be expressed as: The meanings of the letters in the formula are the same as above and will not be repeated here.

[0126] c3) Determine the thermal neutron flux density at the secondary neutron source based on the macroscopic cross-sectional area of ​​uranium-235, the reactor core volume, and the reactor power.

[0127] The available energy released by each uranium-235 nuclear fission is about 200 MeV, so releasing 1J of energy requires about 3.12×10 10 Secondary uranium-235 nuclear fission. According to nuclear physics theory, the power density at any point r in the core is:

[0128] Assuming that only the fission of uranium-235 nuclei caused by thermal neutrons is considered, the power density at each point is integrated:

[0129] The average neutron flux density can be obtained as:

[0130] Where P represents the reactor power, Σ f represents the macroscopic fission cross section of uranium-235, φ(r) represents the flux density at a certain point, and V represents the volume of the core.

[0131] It is understandable that after the nuclear power plant is built, the core shape is fixed and the volume is also constant, the core volume is: V = π·(d / 2) 2 h, where d is the equivalent diameter of the core and h is the height of the core.

[0132] Specifically, let P represent the reactor power, Σ f represents the macroscopic fission cross section of uranium-235, φ(r) represents the flux density at a certain point, and V represents the volume of the core. Substituting into the expression of the average neutron flux density, we obtain:

[0133]

[0134] Right now:

[0135]

[0136] Simplifying the formula we get:

[0137]

[0138] For example, the formula obtained by substituting the first cycle parameters of a unit into the original formula is: Note: Unit: / m 2 *s.

[0139] As a third optional embodiment of the embodiment of the present invention, this third optional embodiment further defines that the method also includes: determining the target overhaul time based on the source strength of the total neutron source and the stacking constraint conditions so that the neutron source can be used for the startup of the next fuel cycle reactor or the startup of other reactors.

[0140] The target overhaul time refers to the appropriate overhaul time after the reactor is shut down to ensure that the neutron source is available for the startup of the next fuel cycle reactor or other reactors.

[0141] In this embodiment, after a fuel cycle is completed and the reactor is shut down, the aforementioned neutron source is removed for use during the startup of the next fuel cycle reactor or other reactors. By using this method to calculate how the total neutron source intensity changes over time after reactor shutdown, the time for post-shutdown overhaul can be adjusted, given the source intensity required for startup, to ensure that the neutron source is available for startup of the next fuel cycle reactor or other reactors. For example, if the current overhaul time is two months, and the current neutron source intensity is calculated using the method provided in this embodiment, if the current neutron source intensity does not meet the required intensity for startup, the overhaul time for this type of neutron source can be appropriately shortened to ensure that it is available for startup of the fuel cycle reactor or other reactors.

[0142] As an optional embodiment of the embodiment of the present invention, this optional embodiment comprehensively considers that the primary neutron source and the secondary neutron source have different positions and energy spectra, and therefore have different impacts on the external detector response, and combines predetermined spatial influencing factors and energy spectrum influencing factors to determine the source strength of the antimony-beryllium source that changes with time after the reactor is shut down, thereby achieving rapid and accurate determination of the neutron source strength. Based on the determined neutron source strength, the overhaul time of the reactor after shutdown can be accurately determined, providing a certain physical basis for reactor startup.

[0143] Example 2

[0144] Figure 2 This is a schematic diagram of the structure of a neutron source intensity determination device provided in the second embodiment of the present invention. The device is applicable to the case of calculating the neutron source intensity. The neutron source intensity determination device can be implemented in the form of hardware and / or software, and the device can be configured in an electronic device. Figure 2 As shown, the device includes: a primary source strength determination module 21, a secondary source strength determination module 22 and a total source strength determination module, wherein:

[0145] A primary source strength determination module 21 is used to determine the source strength of the primary neutron source that changes with time after the reactor is shut down;

[0146] A secondary source strength determination module 22 is configured to determine the source strength of the secondary neutron source that changes with time after the reactor is shut down based on the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;

[0147] The total source strength determination module 23 is used to determine the source strength of the total neutron source that varies with the shutdown time based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, to ensure safe loading and startup of the unit, wherein the total neutron source includes the primary neutron source and the secondary neutron source.

[0148] Furthermore, the secondary source strength determination module 22 includes:

[0149] a secondary neutron source specific intensity determination unit, configured to determine the specific intensity of the secondary neutron source based on the thermal neutron flux density where the secondary neutron source is located and a predetermined target relationship between the outer diameter and the inner diameter of the beryllium tube, wherein the specific intensity of the secondary neutron source is a maximum value when the relationship between the outer diameter and the inner diameter of the beryllium tube satisfies the target relationship;

[0150] The secondary neutron source intensity determination unit is used to determine the intensity of the secondary neutron source that changes with time after the reactor is shut down based on the specific intensity, length and number of the secondary neutron source.

[0151] Optionally, the device further includes:

[0152] Target relationship determination module is used to determine the target relationship according to the neutron irradiation time. 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the Sb neutron macroscopic absorption cross-sectional area and the target relationship between the outer diameter and inner diameter of the beryllium tube corresponding to the maximum value of the secondary neutron source intensity is determined.

[0153] Furthermore, the target relationship determination module is specifically used to:

[0154] According to neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the macroscopic absorption cross-section of Sb neutrons and the 124 The gamma activity of Sb per unit volume;

[0155] Based on the gamma activity, determine the gamma flux density on the surface of the antimony rod that exceeds the reaction threshold energy of beryllium-9;

[0156] Based on the gamma flux density, determine the average gamma flux density in the beryllium tube;

[0157] Based on the average gamma flux density, determine the specific intensity per unit length of the beryllium tube;

[0158] Determine the target relationship between the outer diameter and inner diameter of the beryllium tube corresponding to the maximum specific strength.

[0159] Optionally, the device further includes: a neutron flux density determination module, including:

[0160] an atomic number determination unit for determining the number of uranium-235 atoms per unit volume after the reactor has been in operation for a period of time based on the nuclear power of the reactor core;

[0161] a macroscopic cross-sectional area determination unit for determining the macroscopic cross-sectional area of ​​uranium-235 after the reactor operation time based on the number of uranium-235 atoms per unit volume;

[0162] The neutron flux density determination unit is used to determine the thermal neutron flux density at the secondary neutron source based on the macroscopic cross-sectional area of ​​uranium-235, the reactor core volume, and the reactor power.

[0163] Furthermore, the atomic number determination unit is specifically used to:

[0164] Determine the uranium-235 enrichment after the reactor has been in operation based on the initial uranium-235 mass, reactor core power, and operating factors;

[0165] Determine the abundance of uranium-235 based on its enrichment;

[0166] Based on the abundance of uranium-235, the number of uranium dioxide molecules per unit volume after the operation time is determined;

[0167] Based on the number of uranium dioxide molecules, the number of uranium-235 atoms per unit volume after the running time is determined.

[0168] Optionally, the device further includes: a target overhaul time determination module, configured to:

[0169] The target overhaul time is determined based on the source intensity of the total neutron source and the startup constraints so that the neutron source can be used for the startup of the next fuel cycle reactor or other reactors.

[0170] The neutron source intensity determination device provided in the embodiment of the present invention can execute the neutron source intensity determination method provided in any embodiment of the present invention, and has the corresponding functional modules and beneficial effects of the execution method.

[0171] Example 3

[0172] Figure 3 A schematic diagram of the structure of an electronic device provided for embodiment three of the present invention. The electronic device is intended to represent various forms of digital computers, such as laptop computers, desktop computers, workstations, personal digital assistants, servers, blade servers, mainframe computers, and other suitable computers. The electronic device may also represent various forms of mobile devices, such as personal digital processing, cellular phones, smart phones, wearable devices (such as helmets, glasses, watches, etc.) and other similar computing devices. The components shown herein, their connections and relationships, and their functions are merely examples and are not intended to limit the implementation of the present invention described and / or required herein.

[0173] like Figure 3As shown, the electronic device 30 includes at least one processor 31 and a memory, such as a read-only memory (ROM) 32, a random access memory (RAM) 33, etc., which is communicatively connected to the at least one processor 31. The memory stores a computer program that can be executed by the at least one processor. The processor 31 can perform various appropriate actions and processes according to the computer program stored in the read-only memory (ROM) 32 or the computer program loaded from the storage unit 38 into the random access memory (RAM) 33. Various programs and data required for the operation of the electronic device 30 can also be stored in the RAM 33. The processor 31, ROM 32, and RAM 33 are connected to each other via a bus 34. An input / output (I / O) interface 35 is also connected to the bus 34.

[0174] Multiple components in the electronic device 30 are connected to the I / O interface 35, including an input unit 36, such as a keyboard, a mouse, etc.; an output unit 37, such as various types of displays, speakers, etc.; a storage unit 38, such as a magnetic disk, an optical disk, etc.; and a communication unit 39, such as a network card, a modem, a wireless communication transceiver, etc. The communication unit 39 allows the electronic device 30 to exchange information / data with other devices via a computer network such as the Internet and / or various telecommunication networks.

[0175] Processor 31 can be any general-purpose and / or specialized processing component with processing and computing capabilities. Some examples of processor 31 include, but are not limited to, a central processing unit (CPU), a graphics processing unit (GPU), various specialized artificial intelligence (AI) computing chips, various processors running machine learning model algorithms, a digital signal processor (DSP), and any appropriate processor, controller, microcontroller, etc. Processor 31 executes the various methods and processes described above, such as the method for determining neutron source intensity.

[0176] In some embodiments, the neutron source intensity determination method can be implemented as a computer program, which is tangibly contained in a computer-readable storage medium, such as a storage unit 38. In some embodiments, part or all of the computer program can be loaded and / or installed on the electronic device 30 via the ROM 32 and / or the communication unit 39. When the computer program is loaded into the RAM 33 and executed by the processor 31, one or more steps of the neutron source intensity determination method described above can be performed. Alternatively, in other embodiments, the processor 31 can be configured to execute the neutron source intensity determination method in any other appropriate manner (e.g., by means of firmware).

[0177] Various embodiments of the systems and techniques described herein can be implemented in digital electronic circuit systems, integrated circuit systems, field programmable gate arrays (FPGAs), application specific integrated circuits (ASICs), application specific standard products (ASSPs), system-on-chip systems (SOCs), programmable logic devices (CPLDs), computer hardware, firmware, software, and / or combinations thereof. These various embodiments can include being implemented in one or more computer programs that are executable and / or interpreted on a programmable system that includes at least one programmable processor, which can be a special purpose or general purpose programmable processor that can receive data and instructions from a storage system, at least one input device, and at least one output device, and transmit data and instructions to the storage system, the at least one input device, and the at least one output device.

[0178] Computer programs for implementing the methods of the present invention may be written in any combination of one or more programming languages. These computer programs may be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable data processing device, such that when the computer program is executed by the processor, the functions / operations specified in the flowcharts and / or block diagrams are implemented. The computer program may be executed entirely on the machine, partially on the machine, as a stand-alone software package, partially on the machine and partially on a remote machine, or entirely on a remote machine or server.

[0179] In the context of the present invention, computer-readable storage media can be tangible media that can contain or store a computer program for use with an instruction execution system, device or equipment or used in combination with an instruction execution system, device or equipment. Computer-readable storage media can include but are not limited to electronic, magnetic, optical, electromagnetic, infrared or semiconductor systems, devices or equipment, or any suitable combination of the foregoing. Alternatively, computer-readable storage media can be machine-readable signal media. More specific examples of machine-readable storage media can include electrical connections based on one or more lines, portable computer disks, hard disks, random access memories (RAM), read-only memories (ROM), erasable programmable read-only memories (EPROM or flash memory), optical fibers, portable compact disk read-only memories (CD-ROM), optical storage devices, magnetic storage devices, or any suitable combination of the foregoing.

[0180] To provide interaction with a user, the systems and techniques described herein can be implemented on an electronic device having: a display device (e.g., a CRT (cathode ray tube) or LCD (liquid crystal display) monitor) for displaying information to the user; and a keyboard and pointing device (e.g., a mouse or trackball) through which the user can provide input to the electronic device. Other types of devices can also be used to provide interaction with the user; for example, the feedback provided to the user can be any form of sensory feedback (e.g., visual feedback, auditory feedback, or tactile feedback); and input from the user can be received in any form (including acoustic input, voice input, or tactile input).

[0181] The systems and techniques described herein can be implemented in a computing system that includes back-end components (e.g., as a data server), or a computing system that includes middleware components (e.g., an application server), or a computing system that includes front-end components (e.g., a user computer with a graphical user interface or web browser through which a user can interact with implementations of the systems and techniques described herein), or a computing system that includes any combination of such back-end components, middleware components, or front-end components. The components of the system can be interconnected by any form or medium of digital data communication (e.g., a communication network). Examples of communication networks include: a local area network (LAN), a wide area network (WAN), a blockchain network, and the Internet.

[0182] A computing system may include clients and servers. The clients and servers are typically remote from each other and typically interact via a communication network. This client-server relationship arises through computer programs running on the respective computers, creating a client-server relationship. The server may be a cloud server, also known as a cloud computing server or cloud host. This server is a hosting product within the cloud computing service ecosystem that addresses the management difficulties and limited scalability of traditional physical hosting and VPS services.

[0183] It should be understood that the various forms of the processes shown above can be used to reorder, add, or delete steps. For example, the steps described in the present invention can be performed in parallel, sequentially, or in a different order, as long as the desired results of the technical solution of the present invention can be achieved. This is not limited herein.

[0184] The above specific embodiments do not limit the scope of protection of the present invention. Those skilled in the art will appreciate that various modifications, combinations, sub-combinations, and substitutions may be made based on design requirements and other factors. Any modifications, equivalent substitutions, and improvements made within the spirit and principles of the present invention are intended to be included within the scope of protection of the present invention.

Claims

1. A method for determining neutron source intensity, characterized in that: include: Determine the source intensity of the primary neutron source as a function of time after reactor shutdown; determining a source intensity of the secondary neutron source that varies with time after reactor shutdown based on a determined thermal neutron flux density at a secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source, and the thermal neutron flux density is related to the number of uranium-235 atoms per unit volume after the reactor has been in operation, wherein the number of uranium-235 atoms is determined based on the enrichment of uranium-235, the abundance of uranium-235, and the number of uranium dioxide molecules per unit volume after the reactor has been in operation; Based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, the source strength of the total neutron source that varies with the shutdown time is determined to ensure safe loading and startup of the unit, wherein the total neutron source includes a primary neutron source and a secondary neutron source.

2. The method according to claim 1, characterized in that Determining the source intensity of the secondary neutron source that changes with time after the reactor shutdown based on the determined thermal neutron flux density where the secondary neutron source is located includes: Determining the specific intensity of the secondary neutron source based on the thermal neutron flux density where the secondary neutron source is located and in combination with a predetermined target relationship between the outer diameter and the inner diameter of the beryllium tube, wherein the specific intensity of the secondary neutron source is a maximum value when the relationship between the outer diameter and the inner diameter of the beryllium tube satisfies the target relationship; The source intensity of the secondary neutron source that changes with time after the shutdown is determined according to the specific intensity of the secondary neutron source, the length of the secondary neutron source and the number of the secondary neutron sources.

3. The method according to claim 2, characterized in that Also includes: According to neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The target relationship between the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum value of the secondary neutron source intensity is determined by the relationship between the Sb neutron macroscopic absorption cross-sectional area and the secondary neutron source intensity.

4. The method according to claim 3, characterized in that According to the neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the Sb neutron macroscopic absorption cross-sectional area and the target relationship between the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum value of the secondary neutron source intensity ratio is determined, including: According to neutron irradiation 124 The nuclear density of Sb, the thermal neutron flux density of the antimony-beryllium source, and 124 The relationship between the macroscopic absorption cross-section of Sb neutrons and the 124 The gamma activity of Sb per unit volume; Determining a gamma flux density on the surface of the antimony rod that exceeds the reaction threshold energy of beryllium-9 based on the gamma radioactivity; determining an average gamma flux density in the beryllium tube based on the gamma flux density; determining a specific intensity per unit length of the beryllium tube based on the average gamma flux density; A target relationship between the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum value of the specific strength is determined.

5. The method according to claim 1, wherein The step of determining the thermal neutron flux density where the secondary neutron source is located comprises: Determine the number of uranium-235 atoms per unit volume after the reactor has been in operation, based on the reactor core nuclear power; determining a macroscopic cross-sectional area of ​​uranium-235 after the reactor operation time based on the number of uranium-235 atoms within the unit volume; The thermal neutron flux density at the secondary neutron source is determined based on the macroscopic cross-sectional area of ​​the uranium-235, the reactor core volume, and the reactor power.

6. The method according to claim 5, characterized in that Determining the number of uranium-235 atoms per unit volume after the reactor has been in operation for a certain period of time based on the reactor core nuclear power includes: Determine the uranium-235 enrichment after the reactor has been in operation based on the initial uranium-235 mass, the reactor core power, and the operating factor; determining the abundance of uranium-235 based on the uranium-235 enrichment; Determine the number of uranium dioxide molecules per unit volume after the operating time based on the uranium-235 abundance; The number of uranium-235 atoms per unit volume after the running time is determined based on the number of uranium dioxide molecules.

7. The method according to claim 1, characterized in that Also includes: The target overhaul time is determined based on the source strength of the total neutron source and in combination with the startup constraints, so that the neutron source can be used for the startup of the next fuel cycle reactor or other reactors.

8. A neutron source intensity determination device, characterized in that: include: A primary source strength determination module is used to determine the source strength of the primary neutron source that changes with time after the reactor is shut down; a secondary source strength determination module, configured to determine a source strength of a secondary neutron source that varies with time after reactor shutdown based on a determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source, the thermal neutron flux density is related to the number of uranium-235 atoms per unit volume after the reactor has been in operation, and the number of uranium-235 atoms is determined based on the enrichment of uranium-235, the abundance of uranium-235, and the number of uranium dioxide molecules per unit volume after the reactor has been in operation; The total source strength determination module is used to determine the source strength of the total neutron source that varies with the shutdown time based on the source strength of the primary neutron source and the source strength of the secondary neutron source, combined with predetermined spatial influencing factors and energy spectrum influencing factors, so as to ensure safe loading and startup of the unit, wherein the total neutron source includes a primary neutron source and a secondary neutron source.

9. An electronic device, characterized in that: The electronic device comprises: at least one processor; and a memory communicatively connected to the at least one processor; wherein, The memory stores a computer program executable by the at least one processor. The computer program is executed by the at least one processor so that the at least one processor can execute the neutron source intensity determination method according to any one of claims 1 to 7.

10. A computer-readable storage medium, characterized in that The computer-readable storage medium stores computer instructions, and the computer instructions are used to enable a processor to implement the neutron source intensity determination method according to any one of claims 1 to 7 when executed.

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