CN115101226A - Neutron source intensity determination method, device, equipment and storage medium - Google Patents

Neutron source intensity determination method, device, equipment and storage medium Download PDF

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CN115101226A
CN115101226A CN202210651110.9A CN202210651110A CN115101226A CN 115101226 A CN115101226 A CN 115101226A CN 202210651110 A CN202210651110 A CN 202210651110A CN 115101226 A CN115101226 A CN 115101226A
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source
neutron
determining
neutron source
intensity
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陈丽培
郭述志
杨堃
姚进国
翟伟
王瑞
石昊
朱琳
张哲�
葛同恒
张良
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National Nuclear Demonstration Power Plant Co ltd
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Abstract

The embodiment of the invention discloses a method, a device, equipment and a storage medium for determining the source intensity of a neutron source, wherein the method comprises the following steps: determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down; determining the source intensity of the secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source; the source intensity of the total neutron source changing along with the shutdown time is determined based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combined with the predetermined space influence factor and energy spectrum influence factor, so that safe charging and starting of the unit are guaranteed. By utilizing the method, the source intensity of the neutron source changing along with time after the reactor is stopped is determined by comprehensively considering the different positions and energy spectrums of the primary neutron source and the secondary neutron source and the different influences on the response of the out-of-reactor detector, and combining the predetermined space influence factor and the energy spectrum influence factor, the source intensity of the neutron source is rapidly and accurately determined, and a certain physical basis is provided for the reactor starting.

Description

Neutron source intensity determination method, device, equipment and storage medium
Technical Field
The invention relates to the field of reactor testing, in particular to a method, a device and equipment for determining the source intensity of a neutron source and a storage medium.
Background
During the charging and start-up of a nuclear power plant reactor, the entire process should be under active supervision of neutron detectors in order to ensure critical safety. In-core neutrons are rare during charging and startup, and out-of-core detectors may fail to detect in-core neutron fluence rate levels, either before initial reactor operation or after long shutdown. For this reason, it is usually necessary to install neutron source components in the core, and neutrons generated by these neutron source components can generate enough neutron numbers after subcritical multiplication, so that neutron detectors can detect neutron levels in the core to overcome the measurement blind areas.
Generally, a neutron source is divided into a primary neutron source and a secondary neutron source, which is one of power station nuclear equipment and plays an important role in avoiding a dead zone of core monitoring. The primary neutron source is mainly used in the first cycle of the reactor, and the secondary neutron source is used in the subsequent cycle of the reactor. And the risk of core loading and startup exists when the source strength is insufficient, so that the challenge is formed on the nuclear safety. In the prior art, a plurality of calculations are carried out on a single primary neutron source and a single secondary neutron source, but the primary neutron source and the secondary neutron source have different energy spectrums, different positions and different influences on the response of an out-of-pile detector. The method for determining the source intensity after the window is overhauled is not accurate enough, the change of the source intensity of the neutron source after the window is overhauled cannot be known, and the purposes of guaranteeing safe charging and starting of the unit can be influenced.
Disclosure of Invention
The embodiment of the invention provides a method, a device and equipment for determining the source intensity of a neutron source and a storage medium, which are used for quickly and accurately determining the source intensity of the neutron source and providing a certain physical basis for the startup of a reactor.
In a first aspect, the present embodiment provides a method for determining a source intensity of a neutron source, where the method includes:
determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down;
determining the source intensity of a secondary neutron source which changes along with time after a reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;
and determining the source intensity of a total neutron source which changes along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and in combination with a predetermined space influence factor and a predetermined energy spectrum influence factor so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises the primary neutron source and the secondary neutron source.
In a second aspect, the present embodiment provides a source intensity determining apparatus for a neutron source, including:
the primary source intensity determining module is used for determining the source intensity of the primary neutron source which changes along with time after the reactor is shut down;
the secondary source intensity determining module is used for determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, and the secondary neutron source is an antimony-beryllium source;
and the total source intensity determining module is used for determining the source intensity of a total neutron source changing along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combining with the predetermined space influence factor and energy spectrum influence factor so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises a primary neutron source and a secondary neutron source.
In a third aspect, the present embodiment provides an electronic device, including:
at least one processor; and
a memory communicatively coupled to the at least one processor; wherein the content of the first and second substances,
the memory stores a computer program executable by the at least one processor, the computer program being executable by the at least one processor to enable the at least one processor to perform the method of determining a source intensity of a neutron source according to any of the embodiments of the present invention.
In a fourth aspect, this embodiment provides a computer-readable storage medium storing computer instructions for causing a processor to implement the method for determining a source intensity of a neutron source according to any embodiment of the present invention when the computer instructions are executed.
The embodiment of the invention discloses a method, a device, equipment and a storage medium for determining the source intensity of a neutron source, wherein the method comprises the following steps: determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down; determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source; and determining the source intensity of a total neutron source changing along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combining the predetermined space influence factors and energy spectrum influence factors so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises the primary neutron source and the secondary neutron source. According to the technical scheme, the source intensity of the antimony-beryllium source changing along with time after the reactor is stopped is determined by comprehensively considering that the primary neutron source and the secondary neutron source have different positions, different energy spectrums and different influences on the response of an external detector, and combining the predetermined space influence factor and energy spectrum influence factor, the source intensity of the antimony-beryllium source is quickly and accurately determined, and a certain physical basis is provided for the reactor starting.
It should be understood that the statements in this section do not necessarily identify key or critical features of the embodiments of the present invention, nor do they necessarily limit the scope of the invention. Other features of the present invention will become apparent from the following description.
Drawings
In order to more clearly illustrate the technical solutions in the embodiments of the present invention, the drawings needed to be used in the description of the embodiments will be briefly introduced below, and it is obvious that the drawings in the following description are only some embodiments of the present invention, and it is obvious for those skilled in the art to obtain other drawings based on these drawings without creative efforts.
Fig. 1 is a flowchart of a method for determining a source intensity of a neutron source according to an embodiment of the present invention;
FIG. 1a is a schematic representation of a cross-section of an antimony-beryllium source;
fig. 2 is a schematic structural diagram of a source intensity determining apparatus of a neutron source according to a second embodiment of the present invention;
fig. 3 is a schematic structural diagram of an electronic device according to a third embodiment of the present invention.
Detailed Description
In order to make those skilled in the art better understand the technical solutions of the present invention, the technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the drawings in the embodiments of the present invention, and it is obvious that the described embodiments are only a part of the embodiments of the present invention, and not all of the embodiments. All other embodiments, which can be obtained by a person skilled in the art without making any creative effort based on the embodiments in the present invention, shall fall within the protection scope of the present invention.
It should be noted that the terms "original", "target", and the like in the description and claims of the present invention and the drawings described above are used for distinguishing similar objects and not necessarily for describing a particular sequential or chronological order. It is to be understood that the data so used is interchangeable under appropriate circumstances such that the embodiments of the invention described herein are capable of operation in sequences other than those illustrated or described herein. Furthermore, the terms "comprises," "comprising," and "having," and any variations thereof, are intended to cover a non-exclusive inclusion, such that a process, method, system, article, or apparatus that comprises a list of steps or elements is not necessarily limited to those steps or elements expressly listed, but may include other steps or elements not expressly listed or inherent to such process, method, article, or apparatus.
Example one
Fig. 1 is a flowchart of a method for determining a source intensity of a neutron source according to an embodiment of the present invention, where the method is applicable to a case where the source intensity of the neutron source is calculated after a nuclear reactor is shut down, and the method may be performed by a source intensity determining apparatus, where the source intensity determining apparatus may be implemented in a form of hardware and/or software, and the apparatus may be configured in an electronic device.
At present, a main-flow secondary neutron source is an antimony-beryllium source, and the antimony-beryllium source is formed by cold pressing of uniformly mixed antimony powder and beryllium powder. The antimony-beryllium neutron source does not generate neutrons at first, and is activated into a neutron source only after being irradiated by the neutrons in a reactor. The working principle of a secondary neutron source is the generation of products capable of spontaneous fission by irradiation. The reaction process is as follows: runtime 123 Sb generated by neutron irradiation in a reactor 124 Sb， 124 Sb emits gamma ray in decay process, its half-life period is 60.2 days, gamma ray irradiation 9 Be produces a (γ, n) reaction that produces neutrons with a reaction threshold energy of 1.67 MeV. The reaction formula is as follows:
Figure BDA0003686159000000051
Figure BDA0003686159000000052
consider that the neutron intensity will drop after stopping irradiation for 60 days. For example, if the shutdown overhaul time of a nuclear power plant is prolonged to 60 days or 90 days due to insufficient grid demand, the antimony-beryllium neutron source intensity may be insufficient when the reactor is restarted. There is therefore a need for a method that accurately calculates the source strength of a neutron source as a function of time after a reactor trip.
As shown in fig. 1, the method for determining the source intensity of a neutron source provided in this embodiment may specifically include the following steps:
and S110, determining the source intensity of the primary neutron source which changes along with time after the reactor is shut down.
In conducting the analysis, it was assumed for the sake of convenience of calculation that the antimony-beryllium source was made of natural antimony (including 57.2%) in rod form 121 Sb and 42.8% 123 Sb), and coating a layer of metal beryllium tube. FIG. 1a is a schematic cross-sectional view of an antimony-beryllium source, as shown in FIG. 1a, assuming an outside diameter r of the beryllium tube 2 Inner diameter of r 1 (close adhesion to antimony rod) antimony rod radius r 1 。
It will be appreciated that the primary neutron source is used primarily in the first reactor cycle and the secondary neutron source is used in the subsequent reactor cycle. Consider that the primary neutron source has a source strength that decays over time in subsequent cycles of the reactor, but that it has not completely disappeared. Therefore, compared with the prior art in which separate primary and secondary neutron sources are calculated, the source intensity of the primary neutron source and the source intensity of the secondary neutron source are considered together in the present embodiment. The method is used for determining the source intensity of a primary neutron source after the reactor is shut down.
Specifically, after a reactor is shut down, the source intensity of the primary neutron source changing with time can be expressed as:
Figure BDA0003686159000000053
wherein, N 0 Is the initial source intensity of the primary neutron source, lambda f Is decay constant, t 1 Is the irradiation time in the heap, t 2 Is the time after the heap. It will be appreciated that when the initial source strength of the primary neutron source, the time of irradiation within the stack, and the time after discharge are known, the source strength of the primary neutron source at the corresponding time instant can be determined.
And S120, determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source.
In this embodiment, the secondary neutron source is used in subsequent cycles of the reactor. Wherein, the neutron flux density refers to the number of neutrons passing through a unit area perpendicular to the neutron motion direction in a unit time.
Specifically, the thermal neutron flux density of the antimony-beryllium source is phi, 123 the thermal neutron average macroscopic absorption cross section of Sb is ∑ a ， 124 The nuclear density of Sb is N, N satisfies the following differential equation when the neutron is irradiated:
Figure BDA0003686159000000061
wherein λ is 124 Decay constant of Sb,. lambda. -. 1.332X 10 -7 s -1 。
Let t 1 Denotes the irradiation time of the antimony rod in the stack, t 2 For the time after the stack discharge, in this embodiment, t is kept consistent with the unit of the operating time of the subsequent power station for convenience 1 、t 2 The unit is day, and when t is 0, N is 0, the solution of the equation is:
Figure BDA0003686159000000062
accordingly, the method can be used for solving the problems that, 124 the gamma radioactivity of Sb per unit volume is:
Figure BDA0003686159000000063
further, according to 124 The gamma spectrum of Sb decay is known to have 54% of the gamma ray energy above the (gamma, n) reaction threshold energy of beryllium-9, with 50% being 1.69MeV (MeV in energy units) and 4% being 2.06 MeV. Let the radius of the antimony rod be r 1 The Be tube has an outer diameter r 2 Inner diameter of r 1 (close to the Sb rod), if we neglect the self-absorption of γ by the antimony rod, the γ flux density on the surface of the antimony rod and exceeding the beryllium-9 (γ, n) reaction threshold is:
Figure BDA0003686159000000064
the method is simplified to obtain:
Figure BDA0003686159000000065
neglecting the absorption of γ by beryllium tubes, the γ flux density inside beryllium tubes can be expressed as:
Figure BDA0003686159000000066
further, the average gamma flux density within the beryllium tube can be expressed as:
Figure BDA0003686159000000071
wherein phi (r) is the gamma flux density in the beryllium tube, and the radius of the antimony rod is r 1 The Be tube has an outer diameter of r 2 Inner diameter of r 1 (close to the Sb rods).
Phi (r) in the above step 1 ) Substituting the expression of (a) into the above formula can obtain the average gamma flux density in the beryllium tube as follows:
Figure BDA0003686159000000072
where φ represents the thermal neutron flux density at which the antimony-beryllium source is located, and λ represents 124 Decay constant of Sb, r 1 Denotes the radius of the antimony rod, r 2 Denotes the outer diameter of the beryllium tube, r 1 Denotes the internal diameter of the beryllium tube (close contact with the Sb rod), t 1 Showing the irradiation time of the antimony rods in the stack.
Further, let the (gamma, n) reaction cross-section of beryllium-9 be Σ γ,n And substituting the expression of the average gamma flux density in the beryllium tube, the neutron emissivity (namely the specific intensity) in the beryllium tube per unit length is as follows:
Figure BDA0003686159000000073
where φ represents the thermal neutron flux density at which the antimony-beryllium source is located, Σ a Represent 123 Thermal neutron average macroscopic absorption cross section of Sb, and lambda represents 124 Decay constant of Sb, r 1 Denotes the radius of the antimony rod, r 2 Denotes the outer diameter of the beryllium tube, r 1 Inner diameter of beryllium tube (close contact with Sb rod), t 1 Showing the irradiation time of the antimony rods in the stack.
It will be understood that when r is known 1 And r 2 Proportional relationship between them, and ∑ a The specific strength of the secondary neutron source can be determined. And determining the source intensity of the secondary neutron source which changes along with time after the shutdown 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.
Optionally, the method for determining the source intensity of the secondary neutron source changing with time after the reactor is shut down according to the thermal neutron flux density of the secondary neutron source may be:
a1) and determining the specific strength of the secondary neutron source according to the thermal neutron flux density of the secondary neutron source and a predetermined target relational expression of the outer diameter and the inner diameter of the beryllium tube.
And when the relation between the outer diameter and the inner diameter of the beryllium tube is a target relation formula, the specific intensity of the corresponding secondary neutron source is maximum.
In this embodiment, the ratio of the outer diameter to the inner diameter can be determined according to a predetermined target relation between the outer diameter and the inner diameter of the beryllium tube, that is, the inner diameter of the beryllium tube is r 1 By outside diameter r 2 And (5) characterizing. Further, r is adjusted to 1 The expression of (A) is substituted into a formula of neutron emissivity (namely specific intensity) in a beryllium tube with unit length
Figure BDA0003686159000000081
The maximum value of the specific intensity of the secondary neutron source is given as:
Figure BDA0003686159000000082
wherein, the reaction section Σ of beryllium-9 in this example γ,n Can take the value of 6.2.10 -3 m -1 In addition, the first and second substrates are, in addition, 123 thermal neutron average macroscopic absorption cross section Σ of Sb a This can be calculated by: the density ρ of antimony is known to be 6.69 × 10 3 kg/m 3 The atomic weight A is 121.75, wherein 123 The abundance (atomic%) of Sb ω 43%, 123 thermal neutron average microscopic absorption cross section σ of Sb a 1.63b, then:
Figure BDA0003686159000000083
will be provided with 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, and t is considered due to the fact that different times are involved 2 Time handle S max (t 1 ) As a constant, the following natural decay rule in physics is introduced to obtain:
Figure BDA0003686159000000084
wherein the antimony rod has a radius of r 1 The Be tube has an outer diameter of r 2 Inner diameter of r 1 (close contact with Sb bar), r 1 、r 2 Is given in m, phi is the thermal neutron flux density, and m is given in -2 /s，t 1 Is given in units of days and λ is 124 Decay constant of Sb, t 1 Denotes the irradiation time of the antimony rods in the stack, t 2 The time after the stack is discharged.
b1) And determining the source intensity of the secondary neutron source which changes along with time after the shutdown 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.
Specifically, the specific intensity of the antimony-beryllium source is multiplied by the length of the source to obtain the intensity of the antimony-beryllium source, and if a plurality of antimony-beryllium sources are arranged in the reactor core, the total source intensity of the secondary neutron source is equal to the sum of the intensity of each antimony-beryllium source.
S130, determining the source intensity of a total neutron source changing along with shutdown time based on the source intensity of a primary neutron source and the source intensity of a secondary neutron source and combining predetermined space influence factors and energy spectrum influence factors so as to guarantee safe charging and starting of the unit.
Wherein the total neutron source comprises a primary neutron source and a secondary neutron source.
In the embodiment, the space influence factor k is introduced by considering that the out-of-pile response is different when the neutron source is positioned at different positions 1 (ii) a Considering the different triggering mechanisms of the secondary neutron source and the primary neutron source, the neutron penetration energy is generatedDifferent forces, and the introduction of the energy spectrum factor k to the secondary neutron source 2 . Influence factor k 1 And k 2 Theoretical values can be given according to analysis and calculation and used for first reactor neutron source intensity analysis, and the theoretical values can also be calibrated and determined according to actual engineering test measurement results in different cycle initial stages.
Specifically, after the neutron source space influence factor and the energy spectrum influence factor are considered, the total neutron source intensity is expressed as: s. the T ＝k 1 [S f +k 2 S(t 1 ,t 2 )]Wherein S is f Is the source strength of the primary neutron source, S (t) 1 ,t 2 ) Is the source strength of the secondary neutron source.
The embodiment of the invention discloses a method for determining the source intensity of a neutron source, which comprises the following steps: determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down; determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source; the method comprises the steps of determining the source intensity of a total neutron source changing along with shutdown time based on the source intensity of a primary neutron source and the source intensity of a secondary neutron source and combining predetermined space influence factors and energy spectrum influence factors to guarantee safe charging and starting of a unit, wherein the total neutron source comprises the primary neutron source and the secondary neutron source. According to the technical scheme, the source intensity of the antimony-beryllium source changing along with time after the reactor is stopped is determined by comprehensively considering that the primary neutron source and the secondary neutron source have different positions, different energy spectrums and different influences on the response of an external detector, and combining the predetermined space influence factor and energy spectrum influence factor, the source intensity of the antimony-beryllium source is quickly and accurately determined, and a certain physical basis is provided for the reactor starting.
As a first alternative embodiment of the present invention, the first alternative embodiment is further defined by the method further comprising: according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 and determining a target relational expression of the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum secondary neutron source specific intensity by using the relational expression of the Sb neutron macroscopic absorption sectional area.
In particular, rootUpon neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 determining the relation of Sb neutron macroscopic absorption sectional area 124 Gamma radioactivity of Sb in unit volume. The gamma flux density at the surface of the antimony rod and above the beryllium-9 reaction threshold energy was further determined based on the gamma radioactivity. And determining the average gamma flux density in the beryllium tube according to the gamma flux density. And further determining the specific strength in the beryllium tube per unit length according to the average gamma flux density. And finally, determining a target relational expression of the outer diameter and the inner diameter of the corresponding beryllium tube when the specific strength is at the maximum value.
Further, upon neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 the step of determining a target relational expression of the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum secondary neutron source specific intensity by using the relational expression of the Sb neutron macroscopic absorption sectional area can be expressed as follows:
a2) according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 determining the relation of Sb neutron macroscopic absorption sectional area 124 Gamma radioactivity of Sb in unit volume.
Specifically, the thermal neutron flux density of the antimony-beryllium source is phi, 123 the thermal neutron average macroscopic absorption cross section of Sb is ∑ a ， 124 The nuclear density of Sb is N, N satisfies the following differential equation when the neutron is irradiated:
Figure BDA0003686159000000101
wherein λ is 124 Decay constant of Sb, λ 1.332 × 10 -7 s -1 。
Let t 1 Denotes the irradiation time of the antimony rod in the stack, t 2 For the time after the stack is discharged, in this embodiment, t is kept consistent with the unit of the operating time of the subsequent power station for convenience 1 、t 2 The unit is day, and when t is 0, N is 0, the solution of the equation is:
Figure BDA0003686159000000102
accordingly, the method has the advantages that, 124 the gamma radioactivity of Sb per unit volume is:
Figure BDA0003686159000000103
b2) the gamma flux density on the surface of the antimony rod and exceeding the beryllium-9 reaction threshold energy was determined from the gamma radioactivity.
In particular, according to 124 The gamma spectrum of Sb decay is known to have 54% of the gamma ray energy above the (gamma, n) reaction threshold energy of Be-9, with 50% being 1.69MeV (MeV in energy units) and 4% being 2.06 MeV. Let the radius of the antimony rod be r 1 The external diameter of the beryllium tube is r 2 Inner diameter of r 1 (close to the Sb rod), ignoring gamma self-absorption by the antimony rod, the gamma flux density at the surface of the antimony rod and above the beryllium-9 (gamma, n) reaction threshold is:
Figure BDA0003686159000000111
the method is simplified to obtain:
Figure BDA0003686159000000112
let the external diameter of Be tube Be 2 Inner diameter of r 1 (close to the Sb rod), if the absorption of γ by the beryllium tube is neglected, then the γ flux density inside the Be tube can Be expressed as:
Figure BDA0003686159000000113
c2) from the gamma flux density, the average gamma flux density within the beryllium tube was determined.
Specifically, the average gamma flux density within the Be tube can Be expressed as:
Figure BDA0003686159000000114
wherein phi (r) is the gamma flux in the beryllium tubeDensity, radius of antimony rod r 1 The Be tube has an outer diameter r 2 Inner diameter of r 1 (close to the Sb rod).
Phi (r) in the above step 1 ) Substituting the expression of (a) into the above formula can obtain the average gamma flux density in the beryllium tube as follows:
Figure BDA0003686159000000115
where φ represents the thermal neutron flux density at which the antimony-beryllium source is located, and λ represents 124 Decay constant of Sb, r 1 Denotes the radius of the antimony rod, r 2 Denotes the outer diameter of the beryllium tube, r 1 Denotes the inner diameter of the beryllium tube (close to the Sb rod), t 1 Showing the irradiation time of the antimony rods in the stack.
d2) From the average gamma flux density, the specific strength per unit length of beryllium tube was determined.
Specifically, let Be-9 have a (gamma, n) reaction cross section of ∑ γ,n And substituting the expression of the average gamma flux density in the beryllium tube into the expression, the neutron emissivity (namely the specific strength) in the beryllium tube per unit length is as follows:
Figure BDA0003686159000000121
where φ represents the thermal neutron flux density at which the antimony-beryllium source is located, and λ represents 124 Decay constant of Sb, r 1 Denotes the radius of the antimony rod, r 2 Denotes the outer diameter of the beryllium tube, r 1 Denotes the internal diameter of the beryllium tube (close contact with the Sb rod), t 1 Showing the irradiation time of the antimony rods in the stack.
e2) And determining a target relational expression of the outer diameter and the inner diameter of the corresponding beryllium tube when the specific intensity takes the maximum value.
In this example, it is assumed that the radii of the antimony rod and beryllium tube are pressed in the optimum ratio. In particular, as with S (t) 1 ) Considered as Sb rod radius r 1 To find S (t) 1 ) Maximum of (2) can be
Figure BDA0003686159000000122
The condition that the extreme value is easily obtained is
Figure BDA0003686159000000123
It is understood that the target relation between the external diameter and the internal diameter of the beryllium tube corresponding to the maximum value of the specific intensity can be expressed as follows:
Figure BDA0003686159000000124
by substituting this relationship into the above expression for the specific strength per unit length of beryllium tube, we can obtain:
Figure BDA0003686159000000125
lambda denotes 124 Decay constant of Sb, r 2 Denotes the outer diameter, t, of the beryllium tube 1 Denotes the irradiation time of the antimony rod in the stack, phi denotes the thermal neutron flux density of the antimony rod irradiated, sigma a To represent 123 Thermal neutron average macroscopic absorption cross section of Sb ∑ γ,n Represents the (. gamma.,. n) reaction cross-section of beryllium-9.
In this embodiment, the antimony rods are irradiated with a thermal neutron flux density phi, which is convenient for calculation and can be taken as the average thermal neutron flux density of the reactor core, and the average thermal neutron flux density when the reactor core is operated at a rated power is phi H Then φ can be approximated as follows:
φ＝xφ H for simplicity, the core is considered to be operating at full power, i.e., x is 1. Then the expression of the specific strength of the antimony-beryllium source according to the formula above
Figure BDA0003686159000000126
Determining the secondary neutron source intensity at a certain moment, and then calculating phi at the corresponding moment.
As a second optional embodiment of the present invention, the second optional embodiment further defines that the determining step of the thermal neutron flux density where the secondary neutron source is located includes:
a3) from the reactor core power, the atomic number of uranium-235 per unit volume after the reactor runtime is determined.
Specifically, in this embodiment, the enrichment of uranium-235 after the reactor operation time is determined according to the relationship among the initial uranium-235 quality, the reactor core power, and the operation factor. And determining the abundance of the uranium-235 based on the conversion relation between the negative enrichment degree and the abundance. Further from the abundance of uranium-235, the number of molecules of uranium dioxide per unit volume can be determined. And finally, determining the atomic number of uranium-235 in unit volume according to the molecular number of uranium dioxide.
Further, the step of determining the atomic number of uranium-235 per unit volume after the reactor runtime based on the reactor core power and the reactor runtime can be expressed as:
a31) and determining the enrichment degree of the uranium-235 after the running time of the reactor according to the initial uranium-235 quality, the reactor core power and the running factor.
In this embodiment, the load factor of the power station is considered when the enrichment degree or abundance of the residual nuclear material uranium-235 at a certain stage is calculated for simulating the uranium-235 consumed by the power station operation more truly.
Considering that the primary neutron source still has a large activity for a period of time after the first cycle. The difference of energy spectrums of a primary neutron source and a secondary neutron source is not considered, the numerical value of the total source intensity is calculated, and parameters need to be formulated. Let initial uranium mass be m 1 Initial enrichment of epsilon 1 The number of combustion days is t (unit day), and the uranium dioxide density ρ is 10.42 × 10 3 kg/m 3 The operation factor is 0.93, and the core power P is 4.04 × 10 9 W, uranium-235 fission cross section sigma f 583.5b (b is the fission cross section unit). According to practical values, the mass of the uranium-235 consumed in each 1 megawatt day is 0.00123kg, and the mass of the uranium-235 consumed in the running process until t days can be expressed as:
Δm＝Pt×0.00123×0.93÷10 6 ＝Pt×1.1439×10 -9 (kg)。
the enrichment of uranium-235 by the time of day t can be expressed as:
Figure BDA0003686159000000131
a32) based on the enrichment of uranium-235, the abundance of uranium-235 is determined.
Specifically, according to the conversion relationship between the abundance and the enrichment degree, the abundance of uranium-235 when the current day runs to t can be represented as:
Figure BDA0003686159000000141
a33) the number of uranium dioxide molecules per volume after runtime is determined from the abundance of uranium-235.
The molecular weight of uranium dioxide when the vehicle runs for t days is as follows:
Figure BDA0003686159000000142
wherein c represents the abundance of uranium-235.
The number of uranium dioxide molecules in unit volume at t days is:
Figure BDA0003686159000000143
wherein, the first and the second end of the pipe are connected with each other,
Figure BDA0003686159000000147
denotes the density of uranium dioxide, N 0 Represents the avogalois constant, which is a fixed value.
a34) And determining the atomic number of uranium-235 in the unit volume after the operation time according to the molecular number of uranium dioxide.
The atomic number of uranium-235 in the unit volume at day t is:
Figure BDA0003686159000000144
substituting the abundance expression to obtain:
Figure BDA0003686159000000145
further, substituting the enrichment expression to obtain the atomic number of uranium-235 in the unit volume at t days as follows:
Figure BDA0003686159000000146
in the formula, the letters have the same meanings as above, and are not described herein again.
b3) The macroscopic cross-sectional area of uranium-235 after the reactor run time is determined according to the number of atoms of uranium-235 in a unit volume.
When the reactor runs at full power, uranium-235 in the fuel assembly is continuously consumed, the output power of the reactor is not changed, and the reactivity of the reactor core is required to be maintained by means of lifting rods, boron dilution and the like. Meanwhile, the mass of the uranium-235 consumed every day under the full-power operation condition is fixed, so that a macroscopic cross section of the uranium-235 can be obtained according to the mass of the residual uranium-235, and the average flux density of thermal neutrons required by the full-power operation is judged. When a nuclear power plant normally operates, the power of a reactor usually operates at full power, and after the nuclear power plant is built, the shape of the reactor core is fixed, and the volume of the reactor core is constant, so that the limiting factor of average neutron flux density is the macroscopic absorption cross section of uranium.
Specifically, the fission reaction in the reactor is mainly generated by uranium-235 absorbing thermal neutrons with energy E ═ 0.0253eV, and then the microscopic absorption cross section of uranium-235 is as follows: sigma f 583.5b (b is the unit of cross section).
For uranium-235, the neutron yield per fission was 2.416 and the capture fission ratio was 0.169. The initial macroscopic fission cross-section can be calculated according to the enrichment of the fuel at the beginning of the life: sigma f ＝Nσ f Wherein N is the atomic number of uranium-235 in a unit volume.
And (3) calculating the macroscopic fission section at the end of the service life, wherein the macroscopic fission section is slightly complex, the required fission times are calculated according to the released energy, the consumed nuclear material quality is further obtained, the enrichment degree or abundance of the residual nuclear material uranium-235 at the end of the service life is obtained, and the macroscopic fission section at the end of the service life is finally obtained. Similarly, the macroscopic cross section of the residual uranium-235 at a certain time in the cycle life can be calculated, and then the determined macroscopic cross section of the uranium-235 at the time of the day can be expressed as follows according to the atomic number of the uranium-235 in the unit volume at the time of the day and the microscopic absorption cross section of the uranium-235:
Figure BDA0003686159000000151
the meanings of the letters in the formula are the same as above, and are not described in detail here.
c3) And determining the thermal neutron flux density of the secondary neutron source according to the macroscopic cross section area of the uranium-235, the reactor core volume and the reactor power.
The available energy released per uranium-235 nuclear fission is about 200MeV, so that about 3.12 x 10 is required to release 1J of energy 10 The uranium-235 fission occurs. According to the nuclear physics theory, the power density at any point r of the core is as follows:
Figure BDA0003686159000000161
the power density at each point is integrated assuming that only the fission of the uranium-235 nucleus by thermal neutrons is considered:
Figure BDA0003686159000000162
the average neutron flux density can be found to be:
Figure BDA0003686159000000163
wherein P represents the reactor power, sigma f Represents the macroscopic fission cross section of uranium-235, # (r) represents the flux density at a certain point, and V represents the volume of the core.
It can be understood that after the nuclear power plant is built, the core shape is fixed, the volume is constant, and the core volume is: v pi (d/2) 2 H, where d represents the core equivalent diameter and h represents the core height.
In particular, let P denote the reactor power, Σ f Represents the macroscopic fission cross section of uranium-235, phi (r) represents the flux density at a certain point, V represents the volume of the core, and is substituted into the expression of average neutron flux density to obtain:
Figure BDA0003686159000000164
namely:
Figure BDA0003686159000000165
simplifying the formula yields:
Figure BDA0003686159000000166
exemplarily, the formula after substituting a certain unit initial cycle parameter is obtained on the basis of the original formula:
Figure BDA0003686159000000167
note: the unit is: m/m 2 *s。
As a third alternative embodiment of the present invention, the third alternative embodiment is further defined by the method further comprising: and determining the target overhaul time according to the source strength of the total neutron source and the constraint condition of starting up the reactor, so that the neutron source can be used for starting up the next fuel cycle reactor or other reactors.
The target overhaul time refers to the appropriate overhaul time after the reactor is shut down so as to ensure that the neutron source is used when the next fuel circulation reactor is started or when other reactors are started.
In this embodiment, after a fuel cycle is completed and shutdown is performed, the neutron source is taken out for use in the startup of the reactor of the next fuel cycle or in the startup of other reactors. The method can be used for calculating the change of the source strength of the total neutron source along with the time after the reactor is stopped, and can be used for adjusting the overhaul time after the reactor is stopped under the condition that the source strength required by the reactor is known so as to ensure that the neutron source is used when the next fuel circulation reactor is started or when other reactors are started. For example, if the overhaul time is 2 months, after the source strength of the current neutron source is calculated by using the method provided by the embodiment, if the source strength of the current neutron source does not meet the source strength required for starting the reactor, the overhaul time can be appropriately reduced for the neutron source of this type, so as to meet the requirement for use in starting a fuel cycle reactor or other reactors.
As an optional embodiment of the present invention, the optional embodiment determines the source intensity of the antimony-beryllium source changing with time after the reactor is shut down according to the comprehensive consideration of different positions and energy spectrums of the primary neutron source and the secondary neutron source and different influences on the response of the external detector, and combines the predetermined spatial influence factor and energy spectrum influence factor, so as to realize the fast and accurate determination of the source intensity of the neutron source, and can accurately determine the overhaul time after the reactor is shut down according to the determined source intensity of the neutron source, thereby providing a certain physical basis for the startup of the reactor.
Example two
Fig. 2 is a schematic structural diagram of a source intensity determining apparatus of a neutron source according to a second embodiment of the present invention. The device can be suitable for calculating the source intensity of the neutron source, and the source intensity determining device can be realized in the form of hardware and/or software and can be configured in electronic equipment. As shown in fig. 2, the apparatus includes: a primary source strength determination module 21, a secondary source strength determination module 22, and a total source strength determination module, wherein,
a primary source intensity determining module 21, configured to determine a source intensity of a primary neutron source that changes with time after a reactor shutdown;
the secondary source intensity determining module 22 is used for determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;
and the total source intensity determining module 23 is used for determining the source intensity of the total neutron source changing along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combining the predetermined space influence factor and the energy spectrum influence factor so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises the primary neutron source and the secondary neutron source.
Further, the secondary source intensity determining module 22 includes:
the secondary neutron source specific strength determining unit is used for determining the specific strength of the secondary neutron source by combining a predetermined target relational expression of the outer diameter and the inner diameter of the beryllium tube according to the thermal neutron flux density of the secondary neutron source, and the specific strength of the corresponding secondary neutron source is the maximum value when the relation of the outer diameter and the inner diameter of the beryllium tube is the target relational expression;
and the secondary neutron source strength determining unit is used for determining the source strength of the secondary neutron source which changes along with time after the reactor is stopped according to the specific strength of the secondary neutron source, the length of the secondary neutron source and the number of the secondary neutron sources.
Optionally, the apparatus further comprises:
a target relation determination module for determining a target relation according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 and determining a target relational expression of the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum secondary neutron source specific intensity by using the relational expression of the Sb neutron macroscopic absorption sectional area.
Further, the target relation determination module is specifically configured to:
according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 determining the relation of Sb neutron macroscopic absorption sectional area 124 Gamma radioactivity of Sb per unit volume;
determining a gamma flux density on the surface of the antimony rod and exceeding the beryllium-9 reaction threshold energy based on the gamma radioactivity;
determining the average gamma flux density in the beryllium tube according to the gamma flux density;
determining the specific strength in the beryllium tube per unit length according to the average gamma flux density;
and determining a target relational expression of the outer diameter and the inner diameter of the corresponding beryllium tube when the specific intensity takes the maximum value.
Optionally, the apparatus further comprises: a neutron flux density determination module comprising:
the atomic number determining unit is used for determining the atomic number of uranium-235 in unit volume after the operation time of the reactor according to the nuclear power of the reactor core;
the macroscopic cross-sectional area determining unit is used for determining the macroscopic cross-sectional area of the uranium-235 after the operation time of the reactor according to the atomic number of the uranium-235 in the unit volume;
and the neutron flux density determining unit is used for determining the thermal neutron flux density of the secondary neutron source according to the macroscopic cross-sectional area of the uranium-235, the reactor core volume and the reactor power.
Further, the atomic number determination unit is specifically configured to:
determining the enrichment degree of uranium-235 after the running time of the reactor according to the initial uranium-235 quality, the reactor core power and the running factor;
determining the abundance of uranium-235 based on the enrichment of uranium-235;
determining the number of uranium dioxide molecules in unit volume after running time according to the abundance of uranium-235;
and determining the atomic number of uranium-235 in the unit volume after the operation time according to the molecular number of uranium dioxide.
Optionally, the apparatus further comprises: a target overhaul time determination module to:
and determining the target overhaul time according to the source strength of the total neutron source and the constraint condition of starting up the reactor, so that the neutron source can be used for starting up the next fuel cycle reactor or other reactors.
The neutron source intensity determining device provided by the embodiment of the invention can execute the neutron source intensity determining method provided by any embodiment of the invention, and has corresponding functional modules and beneficial effects of the executing method.
EXAMPLE III
Fig. 3 is a schematic structural diagram of an electronic device according to a third embodiment of the present invention. Electronic devices are intended to represent various forms of digital computers, such as laptops, desktops, workstations, personal digital assistants, servers, blade servers, mainframes, and other appropriate computers. The electronic device may also represent various forms of mobile devices, such as personal digital assistants, cellular phones, smart phones, wearable devices (e.g., helmets, glasses, watches, etc.), and other similar computing devices. The components shown herein, their connections and relationships, and their functions, are meant to be exemplary only, and are not meant to limit implementations of the inventions described and/or claimed herein.
As shown in fig. 3, the electronic device 30 includes at least one processor 31, and a memory communicatively connected to the at least one processor 31, such as a Read Only Memory (ROM)32, a Random Access Memory (RAM)33, and the like, wherein the memory stores a computer program executable by the at least one processor, and the processor 31 may perform various suitable actions and processes according to the computer program stored in the Read Only Memory (ROM)32 or the computer program loaded from a storage unit 38 into the Random Access Memory (RAM) 33. In the RAM 33, various programs and data necessary for the operation of the electronic apparatus 30 can also be stored. The processor 31, the ROM 32, and the RAM 33 are connected to each other via a bus 34. An input/output (I/O) interface 35 is also connected to bus 34.
A plurality of 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, and the like; a storage unit 38 such as a magnetic disk, optical disk, or the like; and a communication unit 39 such as a network card, modem, 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.
The processor 31 may be a variety of general and/or special purpose processing components having 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 suitable processor, controller, microcontroller, or the like. The processor 31 performs the various methods and processes described above, such as the neutron source intensity determination method.
In some embodiments, the neutron source intensity determination method may be implemented as a computer program that is tangibly embodied on a computer-readable storage medium, such as storage unit 38. In some embodiments, part or all of the computer program may be loaded and/or installed onto the electronic device 30 via the ROM 32 and/or the communication unit 39. When the computer program is loaded into RAM 33 and executed by processor 31, one or more steps of the neutron source intensity determination method described above may be performed. Alternatively, in other embodiments, the processor 31 may be configured to perform the neutron source intensity determination method by any other suitable means (e.g., by means of firmware).
Various implementations of the systems and techniques described here above may be implemented in digital electronic circuitry, integrated circuitry, Field Programmable Gate Arrays (FPGAs), Application Specific Integrated Circuits (ASICs), Application Specific Standard Products (ASSPs), system on a chip (SOCs), load programmable logic devices (CPLDs), computer hardware, firmware, software, and/or combinations thereof. These various embodiments may include: implemented in one or more computer programs that are executable and/or interpretable on a programmable system including at least one programmable processor, which may be special or general purpose, receiving data and instructions from, and transmitting data and instructions to, a storage system, at least one input device, and at least one output device.
A computer program 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, special purpose computer, or other programmable data processing apparatus, such that the computer programs, when executed by the processor, cause the functions/acts specified in the flowchart and/or block diagram block or blocks to be performed. A computer program can execute entirely on a machine, partly on a machine, as a stand-alone software package partly on a machine and partly on a remote machine or entirely on a remote machine or server.
In the context of the present invention, a computer-readable storage medium may be a tangible medium that can contain, or store a computer program for use by or in connection with an instruction execution system, apparatus, or device. A computer readable storage medium may include, but is not limited to, an electronic, magnetic, optical, electromagnetic, infrared, or semiconductor system, apparatus, or device, or any suitable combination of the foregoing. Alternatively, the computer readable storage medium may be a machine readable signal medium. More specific examples of a machine-readable storage medium would include an electrical connection based on one or more wires, a portable computer diskette, a hard disk, a Random Access Memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or flash memory), an optical fiber, a portable compact disc read-only memory (CD-ROM), an optical storage device, a magnetic storage device, or any suitable combination of the foregoing.
To provide for interaction with a user, the systems and techniques described here 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 a user; and a keyboard and a pointing device (e.g., a mouse or a trackball) by which a user may provide input to the electronic device. Other kinds of devices may also be used to provide for interaction with a user; for example, 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 may be received in any form, including acoustic, speech, or tactile input.
The systems and techniques described here can be implemented in a computing system that includes a back-end component (e.g., as a data server), or that includes a middleware component (e.g., an application server), or that includes a front-end component (e.g., a user computer having a graphical user interface or a web browser through which a user can interact with an implementation of the systems and techniques described here), or any combination of such back-end, middleware, 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: local Area Networks (LANs), Wide Area Networks (WANs), blockchain networks, and the internet.
The computing system may include clients and servers. A client and server are generally remote from each other and typically interact through a communication network. The relationship of client and server arises by virtue of computer programs running on the respective computers and having a client-server relationship to each other. The server can be a cloud server, also called a cloud computing server or a cloud host, and is a host product in a cloud computing service system, so that the defects of high management difficulty and weak service expansibility in the traditional physical host and VPS service are overcome.
It should be understood that various forms of the flows shown above may be used, with steps reordered, added, or deleted. For example, the steps described in the present invention may be executed in parallel, sequentially, or in different orders, and the present invention is not limited herein as long as the desired result of the technical solution of the present invention can be achieved.
The above-described embodiments should not be construed as limiting the scope of the invention. Those skilled in the art will appreciate that various modifications, combinations, sub-combinations and substitutions are possible, depending on design requirements and other factors. Any modification, equivalent replacement, and improvement made within the spirit and principle of the present invention should be included in the protection scope of the present invention.

Claims (10)

1. A method for determining the source intensity of a neutron source is characterized by comprising the following steps:
determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down;
determining the source intensity of a secondary neutron source which changes along with time after a reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, wherein the secondary neutron source is an antimony-beryllium source;
and determining the source intensity of a total neutron source changing along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combining the predetermined space influence factors and energy spectrum influence factors so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises the primary neutron source and the secondary neutron source.
2. The method of claim 1, wherein determining the source strength of the secondary neutron source as a function of time after a reactor shutdown based on the determined thermal neutron flux density at which the secondary neutron source is located comprises:
determining the specific intensity of the secondary neutron source by combining a predetermined target relational expression of the outer diameter and the inner diameter of the beryllium tube according to the thermal neutron flux density of the secondary neutron source, wherein the specific intensity of the corresponding secondary neutron source is the maximum value when the relation of the outer diameter and the inner diameter of the beryllium tube is the target relational expression;
and determining the source intensity of the secondary neutron source which changes along with time after the shutdown 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 of claim 2, further comprising:
according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at which the source of antimony-beryllium is located, and 124 and determining a target relational expression of the outer diameter and the inner diameter of the beryllium tube corresponding to the maximum secondary neutron source specific intensity by using the relational expression of the Sb neutron macroscopic absorption sectional area.
4. The method of claim 3, wherein the time of neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density at the source of antimony-beryllium, and 124 determining a target relational expression of the outer diameter and the inner diameter of a beryllium tube corresponding to the maximum secondary neutron source specific intensity by using a relational expression of Sb neutron macroscopic absorption sectional areas, wherein the target relational expression comprises the following steps:
according to neutron irradiation 124 Nuclear density of Sb, thermal neutron flux density of Sb-Be source and 124 determining the relation of Sb neutron macroscopic absorption sectional area 124 Gamma radioactivity of Sb per unit volume;
determining a gamma flux density on the surface of the antimony rod and exceeding the beryllium-9 reaction threshold energy based on the gamma radioactivity;
determining the average gamma flux density in the beryllium tube according to the gamma flux density;
determining the specific strength in the beryllium tube per unit length according to the average gamma flux density;
and determining a target relational expression of the outer diameter and the inner diameter of the corresponding beryllium tube when the specific intensity takes the maximum value.
5. The method of claim 1, wherein the step of determining the thermal neutron flux density at which the secondary neutron source is located comprises:
determining the atomic number of uranium-235 in a unit volume after the operation time of the reactor according to the nuclear power of the reactor core;
determining the macroscopic sectional area of the uranium-235 after the operation time of the reactor according to the atomic number of the uranium-235 in the unit volume;
and determining the thermal neutron flux density of the secondary neutron source according to the macroscopic cross section area of the uranium-235, the reactor core volume and the reactor power.
6. The method of claim 5, wherein determining the atomic number of uranium-235 per unit volume after the reactor runtime based on the reactor core power comprises:
determining the enrichment degree of uranium-235 after the running time of the reactor according to the initial uranium-235 quality, the reactor core power and the running factor;
determining the abundance of uranium-235 based on the enrichment of uranium-235;
determining the number of uranium dioxide molecules in unit volume after running time according to the abundance of the uranium-235;
and determining the atomic number of uranium-235 in unit volume after the operation time according to the molecular number of the uranium dioxide.
7. The method of claim 1, further comprising:
and determining the target overhaul time by combining the starting constraint conditions according to the source strength of the total neutron source so as to enable the neutron source to be used when the next fuel circulation reactor is started or other reactors are started.
8. A source intensity determining apparatus for a neutron source, comprising:
the primary source intensity determining module is used for determining the source intensity of a primary neutron source which changes along with time after the reactor is shut down;
the secondary source intensity determining module is used for determining the source intensity of a secondary neutron source which changes along with time after the reactor is stopped according to the determined thermal neutron flux density of the secondary neutron source, and the secondary neutron source is an antimony-beryllium source;
and the total source intensity determining module is used for determining the source intensity of a total neutron source changing along with the shutdown time based on the source intensity of the primary neutron source and the source intensity of the secondary neutron source and combining with the predetermined space influence factor and energy spectrum influence factor so as to ensure safe charging and starting of the unit, wherein the total neutron source comprises 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 coupled to the at least one processor; wherein the content of the first and second substances,
the memory stores a computer program executable by the at least one processor, the computer program being executable by the at least one processor to enable the at least one processor to perform the method of neutron source intensity determination of any of claims 1-7.
10. A computer-readable storage medium storing computer instructions for causing a processor to implement the method of determining a source intensity of a neutron source of any of claims 1-7 when executed.
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