Neutron flux measurement method and device, neutron flux calibration method and system

Abstract

The application discloses a neutron flux measurement method and device and a neutron flux calibration method and system, and relates to the technical field.The measurement method is applied to a neutron flux measurement system, and the neutron flux measurement system comprises a fuel rod sleeve of a micro reactor core critical device; M grooves are arranged on the fuel rod sleeve, and the grooves are used for accommodating detection activation sheets; M is a positive integer; the method comprises the following steps: acquiring a first correction factor, a first nuclear density, a first detection efficiency and M first information corresponding to the M grooves; and determining M absolute neutron fluxes corresponding to the M grooves according to the first correction factor, the first nuclear density, the first detection efficiency and the M first information.According to the embodiment of the application, the method not only meets the structural requirements of a compact reactor core, but also avoids the disturbance of carriers and supports to the neutron field in the reactor core, and the number of times of starting the reactor can be reduced.

Description

Technical Field

[0001] This application belongs to the field of nuclear industry technology, specifically relating to a neutron flux measurement method and device, and a neutron flux calibration method and system. Background Technology

[0002] Portable micro nuclear energy devices represent a new trend in advanced nuclear energy applications both domestically and internationally. They hold enormous market potential for specific applications such as power supply in remote areas and localized microgrids on isolated islands. Gas-cooled microreactors, developed based on prism-type high-temperature gas-cooled reactor technology, are one of the mainstream reactor types. Critical devices are crucial for verifying critical reactor physics theories, reactor physics calculation models and programs, and the nuclear characteristics of new reactor types.

[0003] In related technologies, the detector activation sheet is usually placed at the test location inside the reactor. After irradiation for a certain period of time, it is taken out and its activation intensity is measured to determine the absolute flux level at a specific point. However, the above method has at least the following shortcomings: (1) It is usually necessary to carry out absolute flux measurement and relative distribution measurement at multiple neutron flux level steps, which requires multiple reactor opening tests; (2) A large number of detector activation sheets need to be arranged inside the reactor. The activation sheets and their carriers or supports cause a large disturbance to the neutron field inside the reactor. Moreover, the active region of the microreactor critical device is only tens of centimeters, which makes it difficult to provide a sufficient number of measurement channels with a large enough space. Summary of the Invention

[0004] The technical problem to be solved by this application is to provide a neutron flux measurement method and apparatus, a neutron flux calibration method and system to address the above-mentioned deficiencies in the existing technology. Using this neutron flux measurement method, the structural requirements of a compact reactor core are met, the disturbance of the neutron field inside the reactor by the carrier and support is avoided, and the number of reactor openings can be reduced.

[0005] In a first aspect, embodiments of this application provide a neutron flux measurement method, applied to a neutron flux measurement system, the neutron flux measurement system including a fuel rod sleeve of a microreactor critical device; the fuel rod sleeve is provided with M grooves for accommodating a detector activation plate; M is a positive integer; The method includes: The system acquires a first correction factor, a first nucleus density, a first detection efficiency, and M pieces of first information corresponding to each of the M grooves. The first correction factor is a correction factor related to the radiation time of the activated probe within the reactor core. The first nucleus density is the nucleus density per unit volume of the activated probe. The first detection efficiency is the detection efficiency of the measuring instrument. The first information includes a first measurement count, a second correction factor, a third correction factor, a first microscopic cross-section, and a first volume. The first measurement count is the measurement count of the activated probe within the groove in the measuring instrument. The second correction factor is a correction factor related to the waiting time between the activated probe being removed from the reactor core and being placed into the measuring instrument. The third correction factor is a correction factor related to the measurement time of the activated probe within the groove in the measuring instrument. The first microscopic cross-section is the average microscopic cross-section of the material and energy of the activated probe within the groove, where neutrons within the neutron energy range detectable by the activated probe can undergo nuclear reactions. The first volume is the volume of the activated probe within the groove. Based on the first correction factor, the first nucleon density, the first detection efficiency, and M pieces of first information, determine the M absolute neutron fluxes corresponding to the M slots.

[0006] In some embodiments of the first aspect, based on a first correction factor, a first nucleon density, a first detection efficiency, and M pieces of first information, M absolute neutron fluxes corresponding to the M slots are determined, including: Substituting the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information into formula (1), the M absolute neutron fluxes corresponding to the M slots are calculated. Formula (1) includes: (1) in, Let M be the absolute neutron flux at the i-th groove, where i is a positive integer less than or equal to M; The first measurement count corresponds to the i-th groove; For the highest detection efficiency; The first correction factor; This is the second correction factor corresponding to the i-th groove; The third correction factor corresponds to the i-th groove; N is the first nucleon density; This is the i-th first microscopic section; Let i be the first volume.

[0007] In some embodiments of the first aspect, before obtaining the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of first information corresponding to the M grooves, the method further includes: Obtain the relative distribution curve of neutron flux within the microreactor critical device; The number and location of the grooves are determined based on the relative distribution curve of neutron flux within the reactor.

[0008] In some embodiments of the first aspect, the size of the detector activation sheet is related to a first perturbation limit, a first neutron flux level, a first nucleon density, a second microscopic cross section, and the range of neutron energies that the detector activation sheet can detect; Among them, the first perturbation limit is the perturbation limit of the activated sheet on the neutron field in the reactor; the first neutron flux level is the neutron flux level corresponding to neutron energy E; and the second microscopic cross section is the microscopic cross section of the material in the activated sheet undergoing a nuclear reaction with neutrons of neutron energy E.

[0009] In some embodiments of the first aspect, the neutron energy range includes an upper limit and a lower limit for neutron energy; The size of the detector activation sheet satisfies formula (2); Formula (2) includes: (2) in, To detect the size of the activated sheet; This is the first disturbance limit; The first neutron flux level; N is the first nucleon density; This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy.

[0010] In some embodiments of the first aspect, the energy group of the neutron flux measured by the detector activated sheet is related to a first nucleon density, a first neutron flux level, a second microsection, and a first scale. Wherein, the first neutron flux level is the neutron flux level corresponding to neutron energy E; the second microscopic cross section is the microscopic cross section for detecting nuclear reactions between the material in the activated sheet and neutrons with neutron energy E; and the first ratio is the proportion of the nuclear reaction rate between the activated sheet and neutrons with neutron energy within the neutron energy range that the activated sheet can detect.

[0011] In some embodiments of the first aspect, the energy group of the neutron flux measured by the detector activated sheet satisfies formula (3). Formula (3) includes: (3) Where N is the first nucleon density; This represents the first neutron flux level. This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy. The first proportion.

[0012] Based on the same inventive concept, in a second aspect, embodiments of this application also provide a neutron flux calibration method, comprising: Determine M absolute neutron fluxes based on the neutron flux measurement method of any one of the first aspects; Based on M absolute neutron fluxes and multiple nuclear detection signals corresponding to multiple detector activation plates, a curve showing the correspondence between absolute neutron fluxes and nuclear detection signals is calibrated.

[0013] Based on the same inventive concept, in a third aspect, embodiments of this application provide a neutron flux measurement device, applied to a neutron flux measurement system, the neutron flux measurement system including a fuel rod sleeve of a microreactor critical device; the fuel rod sleeve is provided with M grooves for accommodating a detection activation plate; M is a positive integer; The device includes: The first acquisition module is used to acquire a first correction factor, a first nucleon density, a first detection efficiency, and M pieces of first information corresponding to the M grooves; the first correction factor is a correction factor related to the radiation time of the activated detector in the pile; the first nucleon density is the nucleon density per unit volume of the activated detector; the first detection efficiency is the detection efficiency of the measuring instrument; the first information includes a first measurement count, a second correction factor, a third correction factor, a first micro-section, and a first volume; the first measurement count is the measurement count of the activated detector in the groove in the measuring instrument; the second correction factor is a correction factor related to the waiting time of the activated detector in the groove from being removed from the pile to being placed in the measuring instrument; the third correction factor is a correction factor related to the measurement time of the activated detector in the groove in the measuring instrument; the first micro-section is the average micro-section of the nuclear reaction between the material and energy of the activated detector in the groove and neutrons within the neutron energy range that the activated detector can detect; the first volume is the volume of the activated detector in the groove. The first determining module is used to determine the M absolute neutron fluxes corresponding to the M slots based on the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information.

[0014] Based on the same inventive concept, in a fourth aspect, embodiments of this application also provide a neutron flux calibration system, comprising: The third aspect is a neutron flux measuring device used to determine M absolute neutron fluxes; The calibration device is used to calibrate the correspondence curve between the absolute neutron flux and the nuclear detection signal based on M absolute neutron fluxes and multiple nuclear detection signals obtained one-to-one from multiple detector activation plates.

[0015] Based on the same inventive concept, in a fifth aspect, embodiments of this application provide an electronic device, the electronic device comprising: at least one processor; and a memory communicatively connected to the at least one processor; wherein the memory stores one or more computer programs executable by the at least one processor, the one or more computer programs being executed by the at least one processor to enable the at least one processor to perform the above-described neutron flux measurement method or neutron flux calibration method.

[0016] Based on the same inventive concept, in a sixth aspect, embodiments of this application provide a computer-readable storage medium having a computer program stored thereon, wherein the computer program, when executed by a processor, implements the above-described neutron flux measurement method or neutron flux calibration method.

[0017] Based on the same inventive concept, in a seventh aspect, embodiments of this application provide a computer program product that includes computer-readable code or a non-volatile computer-readable storage medium carrying computer-readable code. When the computer-readable code is run in the processor of an electronic device, the processor in the electronic device executes the above-described neutron flux measurement method or neutron flux calibration method.

[0018] According to the neutron flux measurement method and apparatus, and neutron flux calibration method and system provided in the embodiments of this application, M grooves are provided on the fuel rod sleeve. These grooves are used to accommodate the detection activation plates. That is, grooves for arranging the detection activation plates are provided on the fuel sleeve of the microreactor critical device, avoiding the need for a carrier and support for the detection activation plates. This satisfies the structural requirements of a compact reactor core and avoids disturbance of the neutron field within the reactor by the carrier and support. In addition, by using M first information corresponding to each of the M grooves, M absolute neutron fluxes corresponding to each of the M grooves are determined. That is, the absolute neutron fluxes at different detection points are measured simultaneously by the detection activation plates, which can reduce the number of reactor startups.

[0019] It should be understood that the description in this section is not intended to identify key or essential features of the embodiments of this application, nor is it intended to limit the scope of this application. Other features of this application will become readily apparent from the following description. Attached Figure Description

[0020] The accompanying drawings are provided to further illustrate the present application and form part of the specification. They are used together with the embodiments of the present application to explain the application and do not constitute a limitation thereof. The above and other features and advantages will become more apparent to those skilled in the art from the detailed example embodiments described with reference to the accompanying drawings, in which: Figure 1 This illustration shows a flowchart of a neutron flux measurement method provided in an embodiment of this application. Figure 2This diagram illustrates the radial direction of the groove detection point provided in an embodiment of this application. Figure 3 This illustration shows an axial schematic diagram of the groove detection point provided in an embodiment of this application; Figure 4 This illustrates the axial distribution trend of neutron flux predicted by a core model simulated using a high-fidelity neutronics program provided in an embodiment of this application. Figure 5 This illustrates the axial distribution trend of neutron flux predicted by a core model simulated using a high-fidelity neutronics program provided in an embodiment of this application. Figure 6 The in-pile neutron energy spectrum curves provided in the embodiments of this application are shown; Figure 7 This application provides an embodiment of the invention. 197 The curve of Au(n,γ) reaction rate as a function of neutron energy; Figure 8 This application provides an embodiment of the results showing the relative distribution of axial neutron flux in the reactor core. Figure 9 This illustration shows a structural schematic diagram of a neutron flux measurement device provided in an embodiment of this application; Figure 10 This illustration shows a structural diagram of an electronic device provided in an embodiment of this application. Detailed Implementation

[0021] To enable those skilled in the art to better understand the technical solutions of this application, exemplary embodiments of this application are described below in conjunction with the accompanying drawings, including various details of the embodiments of this application to aid understanding. These should be considered merely exemplary. Therefore, those skilled in the art should recognize that various changes and modifications can be made to the embodiments described herein without departing from the scope and spirit of this application. Similarly, for clarity and conciseness, descriptions of well-known functions and structures are omitted in the following description.

[0022] Where there is no conflict, the various embodiments of this application and the features thereof may be combined with each other.

[0023] As used herein, the term “and / or” includes any and all combinations of one or more related enumerated entries.

[0024] The terminology used herein is for the purpose of describing particular embodiments only and is not intended to limit the application. As used herein, the singular forms “a” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will also be understood that when the terms “comprising” and / or “made of” are used in this specification, the presence of the stated feature, integral, step, operation, element, and / or component is specified, but the presence or addition of one or more other features, integrals, steps, operations, elements, components, and / or groups thereof is not excluded. Terms such as “connected” or “linked” are not limited to physical or mechanical connections but can include electrical connections, whether direct or indirect.

[0025] Unless otherwise specified, all terms used herein (including technical and scientific terms) have the same meaning as commonly understood by one of ordinary skill in the art. It will also be understood that terms such as those defined in commonly used dictionaries should be interpreted as having a meaning consistent with their meaning in the context of the relevant art and this application, and will not be interpreted as having an idealized or overly formal meaning, unless expressly so defined herein.

[0026] It should be understood that the term "and / or" used in this article is merely a description of the relationship between related objects, indicating that three relationships can exist. For example, A and / or B can represent: A existing alone, A and B existing simultaneously, and B existing alone. Additionally, the character " / " in this article generally indicates that the preceding and following related objects have an "or" relationship.

[0027] The neutron flux measurement method provided in this application embodiment can be applied to a neutron flux measurement system and is applicable to the neutron flux calibration process. The neutron flux measurement system may include a fuel rod sleeve for a microreactor critical device. The fuel rod sleeve has M grooves, which can be used to accommodate detector activation plates. It is understood that each groove can accommodate at least one detector activation plate, and the number of detector activation plates accommodated in different grooves may be the same or different; this is not limited here.

[0028] Exemplarily, the neutron flux measurement system may further include a precision balance, an activity measuring instrument, a timer, and a computer. The precision balance can measure the mass of the activated probe sheet and then calculate its volume. The activity measuring instrument can measure the activity of the activated probe sheet after activation. The timer can record the irradiation time, the waiting time before measurement, and the measurement time. The computer can process the data, that is, it can execute the neutron flux measurement method or neutron flux calibration method provided in the embodiments of this application.

[0029] It should be noted that this neutron flux measurement method can be performed by a neutron flux measurement device and electronic equipment. The following explanation will take the execution of this neutron flux measurement method by electronic equipment as an example.

[0030] like Figure 1 As shown, the neutron flux measurement method provided in this application includes steps S110 to S120.

[0031] S110. Obtain the first correction factor, the first nucleon density, the first detection efficiency, and M pieces of first information corresponding to the M grooves; the first correction factor is a correction factor related to the radiation time of the detection activation sheet in the pile; the first nucleon density is the nucleon density per unit volume of the detection activation sheet; the first detection efficiency is the detection efficiency of the measuring instrument; the first information includes the first measurement count, the second correction factor, the third correction factor, the first micro-section, and the first volume; the first measurement count is the measurement count of the detection activation sheet in the groove in the measuring instrument; the second correction factor is a correction factor related to the waiting time of the detection activation sheet in the groove from being taken out of the pile to being placed in the measuring instrument; the third correction factor is a correction factor related to the measurement time of the detection activation sheet in the groove in the measuring instrument; the first micro-section is the average micro-section of the nuclear reaction between the material and energy of the detection activation sheet in the groove and neutrons within the neutron energy range that the detection activation sheet can detect; the first volume is the volume of the detection activation sheet in the groove.

[0032] S120. Based on the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of first information, determine the M absolute neutron fluxes corresponding to the M slots.

[0033] According to the neutron flux measurement method provided in the embodiments of this application, M grooves are provided on the fuel rod sleeve. These grooves are used to accommodate the detection activation plates. That is, grooves for arranging the detection activation plates are provided on the fuel sleeve of the microreactor critical device, avoiding the need for a carrier and support for the detection activation plates. This satisfies the structural requirements of a compact reactor core and avoids disturbance of the neutron field within the reactor by the carrier and support. In addition, M absolute neutron fluxes corresponding to M grooves are determined by M pieces of first information corresponding to each of the M grooves. That is, the absolute neutron fluxes at different detection points are measured simultaneously by the detection activation plates, which can reduce the number of reactor startups.

[0034] The specific implementation methods for each of the above steps are described below.

[0035] In step S110, for example, the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information corresponding to the M grooves can all be pre-stored in the electronic device for direct retrieval later, or can be input through the input interface of the electronic device, which is not limited here.

[0036] For example, the first correction factor A satisfies: e is the natural constant, and T is the irradiation time. To detect the decay constant of the activated sheet after activation.

[0037] For example, the second correction factor satisfy: , t wi For waiting time.

[0038] For example, the third correction factor satisfy: , t ci For measuring time.

[0039] For example, a gas-cooled microreactor critical device is used as an example. The gas-cooled microreactor critical device uses graphite as the moderator, reflector, and structural material. The fuel is a cylindrical fuel pellet composed of particulate fuel and a ceramic matrix material. Multiple fuel pellets are placed inside a graphite sleeve to form a fuel rod. Grooves for activating detection plates are provided on the graphite sleeve of the fuel rod; the grooves are only 10 mm in diameter and 1 mm deep. For fuel rods that do not contain detection points, the graphite sleeve does not have a groove structure. This limited number of grooves avoids the use of traditional activation plate carriers and supports, meeting the structural requirements of a compact core, and also avoids significant disturbance to the neutron field within the reactor.

[0040] For example, the radial schematic diagram of the groove detection point provided in the embodiments of this application is as follows: Figure 2 As shown, the axial schematic diagram of the groove detection point is as follows. Figure 3 As shown. It should be noted that, Figure 3 Taking M as an example of 7 is not intended to limit this application.

[0041] For example, by simultaneously irradiating the detector activation sheet at different detection points during the measurement process, it is possible to simultaneously measure the relative distribution and absolute flux of neutrons within the reactor. This can reduce the number of reactor start-up tests by at least half.

[0042] In some embodiments, before obtaining the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of first information corresponding to the M grooves, the method further includes: Obtain the relative distribution curve of neutron flux within the microreactor critical device; The number and location of the grooves are determined based on the relative distribution curve of neutron flux within the reactor.

[0043] In this embodiment, the number of grooves (i.e., the value of M) and their placement are determined by the relative distribution curve of neutron flux within the reactor, so that the purpose of neutron flux measurement can be achieved with as few detection points as possible.

[0044] For example, a high-fidelity neutronics program can be used to simulate the core model to predict the relative neutron flux distribution curve within the reactor. More specifically, a three-dimensional, detailed core model of the microreactor critical device can be established using the general Monte Carlo (RMC) program, and the relative neutron flux distribution curve within the reactor can be theoretically calculated.

[0045] For example, Figure 4 These are curves showing the relative distribution of neutron flux along the core axis in different energy ranges, calculated by the RMC program. It can be seen that due to differences in the structure of materials such as moderators and fuel, the axial distribution curves of neutron flux in different energy ranges of the gas-cooled microreactor vary significantly. Figure 4 In the neutron flux distribution curve, neutrons below 0.625 eV exhibit a concave curve distribution, neutrons between 0.625 eV and 0.1 MeV show a relatively smooth single-segment curve distribution, while neutrons above 0.1 MeV exhibit a convex curve distribution. The predicted trend and complexity of the neutron flux distribution curve will effectively guide the selection of the number and layout of detection sites. Figure 3 As shown in this embodiment, in order to measure the relative distribution of axial neutron flux in a specified energy range, seven grooves for detection activation plates are provided on the graphite sleeve of the fuel rod.

[0046] For example, when the relative distribution curve of neutron flux in the reactor is a relatively smooth single-segment curve, the number of grooves can be 7, and the positions can be axial position numbers 1, 3, 5, 7, 8, 11, and 14; when the relative distribution curve of neutron flux in the reactor is a concave curve, the number of grooves can be 8, and the positions can be axial position numbers 1, 3, 5, 7, 8, 11, 13, and 14; when the relative distribution curve of neutron flux in the reactor is a convex curve, the number of grooves can be 9, and the positions can be axial position numbers 1, 3, 5, 6, 7, 8, 9, 12, and 14.

[0047] For example, gold is chosen as the material for the detector activation sheet. 197 Au is 100% abundant in nature. When irradiated in a neutron field, it undergoes an (n,γ) reaction to produce... 198 Au. 198 Au has a half-life of 2.7 days and decays to produce gamma rays with an energy of 411.8 keV. The intensity of the gamma rays can be measured by instruments such as high-purity germanium detectors. Figure 5 yes 197 The microscopic cross-sectional curves of the reactions of Au with neutrons of different energies at (n, γ) show that... 197 The microscopic cross-sections of the (n, γ) reactions of Au with neutrons of different energies vary greatly. Figure 5 The horizontal axis represents neutron energy, and the vertical axis represents the microscopic cross-section.

[0048] In some implementations, the size of the detector activation sheet is related to a first perturbation limit, a first neutron flux level, a first nucleon density, a second micro-section, and the range of neutron energies that the detector activation sheet can detect; Among them, the first perturbation limit is the perturbation limit of the activated sheet on the neutron field in the reactor; the first neutron flux level is the neutron flux level corresponding to neutron energy E; and the second microscopic cross section is the microscopic cross section of the material in the activated sheet undergoing a nuclear reaction with neutrons of neutron energy E.

[0049] In this embodiment, limiting the size of the detector activation plate by using a first perturbation limit (i.e., a limit on neutron field perturbation) can reduce perturbation to the neutron field within the reactor and improve the accuracy of the experiment.

[0050] In some examples, the neutron energy range includes an upper limit and a lower limit for neutron energy; The size of the detector activation sheet satisfies formula (2); Formula (2) includes: (2) in, To detect the size of the activated sheet; This is the first disturbance limit; The first neutron flux level; N is the first nucleon density; This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy.

[0051] For example, the unit of d is centimeters (cm); The unit is units / cm 2 / second; N is measured in units of / cm 3 ; It is a measure of the average probability that an incident neutron of a given energy E will undergo a nuclear reaction with a target nucleus (i.e., a nuclide atom in the detector activation sheet), and the unit is cm. 2 ; and The unit is eV or MeV.

[0052] For example, taking gold as the material of the detector activation sheet, the size of the gold detector activation sheet satisfies formula (2). represent 197 The integral value of the reaction rate when Au reacts with neutrons with energies between E1 and E2 (n, γ). The integral value of the neutron flux with energies between E1 and E2 represents the probability of a nuclear reaction occurring per unit distance traveled by a neutron in the medium, which can be theoretically calculated using the RMC program. Therefore, by setting a neutron field perturbation limit to restrict the size of the detector activation plate, perturbation to the neutron field can be reduced. In the embodiments of this application, calculation results show that if the neutron field perturbation limit is set... If the thickness is 1%, then the thickness of the detection activation sheet should not exceed 0.04 mm. Optionally, in this embodiment of the application, the thickness of the Au detection activation sheet is selected to be 0.01 mm, which only disturbs the local neutron field in the reactor by 0.25%, which is beneficial to improving the accuracy of the experiment.

[0053] The derivation process of formula (2) is explained below.

[0054] According to N and Parameter definition, The nuclear neutron is a measure of the average probability that an incident neutron of a given energy E will react with all atomic nuclei within a unit volume. It also characterizes the probability that an incident neutron of a given energy E will travel a unit distance in a medium (i.e., the detector activation sheet material) and react with atomic nuclei. Its unit is cm. -1 .

[0055] therefore, It represents the weighted average of the probability that all neutrons with energies between E1 and E2 will undergo nuclear reactions with atomic nuclei while traveling a unit distance in the detector activated sheet material.

[0056] To ensure that the disturbance of the reactor neutron field caused by the activated detector sheet is less than the limit. That is, the probability of the neutron to be measured undergoing a nuclear reaction with the detector activation sheet material is less than (The nuclear reaction between the activated material and neutrons alters the original neutron distribution, thus perturbing the original neutron field), as shown in the following equation:

[0057] Therefore, the size constraint formula (2) for the detector activation sheet is obtained.

[0058] In some implementations, the energy group of the neutron flux measured by the detector activation sheet is related to a first nucleon density, a first neutron flux level, a second microsection, and a first scale. Wherein, the first neutron flux level is the neutron flux level corresponding to neutron energy E; the second microscopic cross section is the microscopic cross section for detecting nuclear reactions between the material in the activated sheet and neutrons with neutron energy E; and the first ratio is the proportion of the nuclear reaction rate between the activated sheet and neutrons with neutron energy within the neutron energy range that the activated sheet can detect.

[0059] In this embodiment, the measured neutron energy group is determined by the first ratio (i.e., the proportion of activation reaction rate), which enables accurate measurement of the complex neutron energy spectrum of the micro-reactor.

[0060] In some examples, the energy group of the neutron flux measured by the probe activated sheet satisfies formula (3); Formula (3) includes: (3) Where N is the first nucleon density; This represents the first neutron flux level. This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy. The first proportion.

[0061] For example, the energy group of the neutron flux measured by the gold detector activated sheet satisfies formula (3), which represents the range of neutron energies that the gold detector activated sheet can represent. Figure 6 It is a curve of the neutron energy spectrum inside the critical device reactor. Figure 7 yes 197 The Au(n,γ) reaction rate versus neutron energy curve. In this embodiment, the proportion of nuclear reaction rate... Take 99%, combined Figure 6 and Figure 7 The neutron energy range can be determined to be 3 × 10⁻⁶. -3 eV~3×10 3 eV.

[0062] The derivation process of formula (3) is explained below.

[0063] Neutron flux is measured using a detector activation sheet, based on the principle of nuclear reactions between the material of the detector activation sheet and all neutrons in the reactor. However, the neutron flux levels vary at different locations and energies within the reactor, ranging from 10... -12 MeV to 20 MeV, spanning more than ten orders of magnitude. Furthermore, the probability of the same detector activating material undergoing a nuclear reaction with neutrons of different energies varies (e.g., Figure 5 (The curves show that different detector activation sheet materials have different probabilities of undergoing nuclear reactions with neutrons of the same energy).

[0064] To ensure that the neutron flux level measured by the detector activation sheet material used in the experiment is consistent with and accurate with the actual neutron flux level, it is necessary to ensure that the nuclear reaction of neutrons with energies between E1 and E2 has a sufficiently high proportion so as to fully represent the neutron flux level with energies between E1 and E2.

[0065] because This is a measure of the average probability that an incident neutron of a given energy E will react with all the atomic nuclei within a unit volume of the detector-activated plate. This represents the rate at which all neutrons with energies between E1 and E2 undergo nuclear reactions with all atomic nuclei within a unit volume of the detector-activated plate. This represents the rate at which all neutrons in the universal group undergo nuclear reactions with all atomic nuclei within a unit volume of the detector activation sheet. Therefore, the energy group interval E1~E2 corresponding to the neutrons measured by the detector activation sheet material used in the experiment can be determined by formula (3).

[0066] In step S120, in some embodiments, based on the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of first information, the M absolute neutron fluxes corresponding to the M slots are determined, including: Substituting the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information into formula (1), the M absolute neutron fluxes corresponding to the M slots are calculated. Formula (1) includes: (1) in, Let M be the absolute neutron flux at the i-th groove, where i is a positive integer less than or equal to M; The first measurement count corresponds to the i-th groove; For the highest detection efficiency; The first correction factor; This is the second correction factor corresponding to the i-th groove; The third correction factor corresponds to the i-th groove; N is the first nucleon density; This is the i-th first microscopic section; Let i be the first volume.

[0067] In this embodiment, the absolute neutron flux at each groove can be determined quickly and accurately using the above formula (1).

[0068] For example, Figure 7 The results show the relative distribution of neutron flux along the core axis. It can be seen that the deviation between the measured value and the theoretical value is within 2%, indicating high accuracy.

[0069] The derivation process of formula (1) is explained below.

[0070] After determining that the measured neutron energy range was between E1 and E2, The total rate of nuclear reactions between all neutrons within the energy range and atomic nuclei within the entire volume of the detector-activated sheet is represented by R, i.e.: .

[0071] After the activated probe sheet is irradiated in the reactor for a certain period of time, its activity is R. A (defined as above); During the time it takes for the activated probe to be removed from the pile and placed in the instrument to measure its activity, its activity gradually decreases to R. A B (defined as above); the total number of rays emitted by the detector plate during the time period in which its decay radiation is measured in the instrument is R. A B C (defined as above). Not all emitted rays can be measured by the detection instrument; depending on the instrument's performance, its detection efficiency is... At that time, the measurement count obtained by the detection instrument .

[0072] Therefore, based on the above two equations, the neutron flux level (i.e., absolute neutron flux) at the location of the detector activation plate can be obtained as follows:

[0073] In one experiment, the irradiation time of the detector activation plates at different locations is the same (determined by the core startup, operation, and shutdown), but the waiting time before measurement, the measurement time in the instrument, the volume, the measurement count, etc. of each detector activation plate are different. Therefore, the neutron flux level at the location of the i-th detector activation plate is given by formula (1).

[0074] Based on the same inventive concept, embodiments of this application also provide a neutron flux calibration method. This method may include steps S210 to S220.

[0075] S210. Determine M absolute neutron fluxes according to the neutron flux measurement method of any of the above embodiments.

[0076] S220. Based on the M absolute neutron fluxes and the multiple nuclear detection signals corresponding to the multiple detector activation plates, the corresponding curves between the absolute neutron fluxes and the nuclear detection signals are calibrated.

[0077] The neutron flux calibration method provided in this application includes a neutron flux measurement method, which has the beneficial effects and implementation methods of the neutron flux measurement method provided in this application. For details, please refer to the specific description of the neutron flux measurement method in the above embodiments. This embodiment will not repeat the description here.

[0078] To better understand the methods provided in the embodiments of this application, the following description is provided in conjunction with specific implementation methods.

[0079] The neutron flux measurement and calibration method and system for microreactor critical devices provided in this application are beneficial for reducing the number of reactor start-up tests, reducing the interference of the measurement system on the neutron field inside the reactor, and adapting to the compact core arrangement and complex neutron energy spectrum of microreactors to improve measurement accuracy.

[0080] like Figure 2 and Figure 3 As shown, this invention provides a method and system for measuring and calibrating neutron flux in a gas-cooled microreactor critical device. The gas-cooled microreactor critical device uses graphite as a moderator, reflector, and structural material. The fuel consists of cylindrical fuel pellets made of particulate dispersed fuel and a ceramic matrix material. Multiple fuel pellets are placed inside a graphite sleeve to form fuel rods. Grooves for activating detection plates are provided on the graphite sleeve of the fuel rods; the grooves are only 10 mm in diameter and 1 mm deep. For fuel rods that do not contain detection points, the graphite sleeve does not have a groove structure. This limited number of grooves avoids the use of traditional activation plate carriers and supports, meeting the structural requirements of a compact reactor core, and also avoids significant disturbance to the neutron field within the reactor.

[0081] By simultaneously irradiating activation sheets at different detection points during the measurement process, both the relative distribution and absolute flux of neutrons within the reactor can be measured simultaneously. This can reduce the number of reactor start-up tests by at least half.

[0082] The locations of the grooves for arranging the detector activation plates are determined by the neutron flux distribution trend predicted by the core model simulated using a high-fidelity neutronics program. Preferably, a three-dimensional, detailed core model of the critical device is established using the general Monte Carlo program RMC, the curve of the relative distribution of neutron flux within the reactor is theoretically calculated, and the arrangement of the detector points is optimized to measure the relative distribution trend of neutron flux within the reactor with high precision using as few detector points as possible. Figure 4 These are curves showing the relative distribution of neutron flux along the core axis in different energy ranges, calculated by the RMC program. It can be seen that due to differences in the structure of materials such as moderators and fuel, the axial distribution curves of neutron flux in different energy ranges of the gas-cooled microreactor vary significantly. Figure 4 In the neutron flux distribution curve, neutrons below 0.625 eV exhibit a concave curve distribution, neutrons between 0.625 eV and 0.1 MeV show a relatively smooth single-segment curve distribution, while neutrons above 0.1 MeV exhibit a convex curve distribution. The predicted trend and complexity of the neutron flux distribution curve will effectively guide the selection of the number and layout of detection sites. Figure 3 As shown in this embodiment, in order to measure the relative distribution of axial neutron flux in a specified energy range, seven grooves for detection activation plates are provided on the graphite sleeve of the fuel rod.

[0083] In this embodiment, gold is selected as the material for the activation sheet. 197 Au is 100% abundant in nature. When irradiated in a neutron field, it undergoes an (n,γ) reaction to produce...198 Au. 198 Au has a half-life of 2.7 days and decays to produce gamma rays with an energy of 411.8 keV. The intensity of the gamma rays can be measured by instruments such as high-purity germanium detectors. Figure 5 yes 197 The microscopic cross-sectional curves of the reactions of Au with neutrons of different energies at (n, γ) show that... 197 The microscopic cross-sections of the (n, γ) reactions of Au with neutrons of different energies vary greatly.

[0084] The dimensions of the gold detector activation sheet satisfy formula (2). represent 197 The integral value of the reaction rate when Au reacts with neutrons with energies between E1 and E2 (n, γ). The integral value of the neutron flux with energies between E1 and E2 represents the probability of a nuclear reaction occurring per unit distance traveled by a neutron in the medium, which can be theoretically calculated using the RMC program. Therefore, by setting a neutron field perturbation limit to restrict the size of the detector activation plate, perturbation to the neutron field can be reduced. In this embodiment, calculation results show that if the neutron field perturbation limit is set... If the value is 1%, then the thickness of the detection activation sheet should not exceed 0.04 mm. Preferably, in this embodiment, the thickness of the Au detection activation sheet is selected as 0.01 mm, which only disturbs the local neutron field in the reactor by 0.25%, which is beneficial to improving the accuracy of the experiment.

[0085] The energy group of the neutron flux measured by the gold-activated sheet satisfies formula (3), which represents the range of neutron energies that the gold-activated sheet can represent. Figure 6 It is a curve of the neutron energy spectrum inside the critical device reactor. Figure 7 yes 197 The Au(n,γ) reaction rate versus neutron energy curve. In this embodiment, the proportion of nuclear reaction rate... Take 99%, combined Figure 6 and Figure 7 The neutron energy range can be determined to be 3 × 10⁻⁶. -3 eV~3×10 3 eV.

[0086] (2) (3) In formulas (2) and (3), d represents the size of the detector activation plate; To detect the limits of disturbance to the neutron field within the reactor by the activated sheet; This represents the neutron flux level corresponding to the neutron energy E. N represents the nucleon density per unit volume of the probe activation sheet; To detect the microscopic cross-section of the nuclear reaction between the material within the activated sheet and neutrons with energy E; E1 and E2 are the lower and upper limits of the neutron energy detectable by the activated sheet; To detect the proportion of nuclear reaction rates between the activated sheet and neutrons with energies in the E1~E2 range.

[0087] The neutron flux level at each detection point is obtained by formula (1). Figure 8 The results show the relative distribution of neutron flux along the core axis. It can be seen that the deviation between the measured value and the theoretical value is within 2%, indicating high accuracy.

[0088] (1)

[0089] In formula (1), Let be the absolute neutron flux at the i-th detection point; The measurement count of the activated plate at the i-th detection point in the measuring instrument; To improve the detection efficiency of the measuring instrument; A is a correction factor related to the irradiation time of the activated sheet in the pile. e is the natural constant, and T is the irradiation time. To detect the decay constant of the activated sheet after activation; B i This is a correction factor related to the waiting time from when the activated piece at the i-th detection point is removed from the pile to when it is placed into the measuring instrument. , t wi Waiting time; C i This is a correction factor related to the measurement time of the activation plate at the i-th detection point within the measuring instrument. , t ci For measuring time; N represents the nucleon density per unit volume of the probe activation sheet; The average microscopic cross section of the activated sheet material at the i-th detection point undergoes a nuclear reaction with neutrons with energies between E1 and E2; V i Let be the volume of the activated sheet at the i-th detection point.

[0090] In this embodiment, the neutron flux measurement system includes a detector activation sheet, a precision balance, an activity measuring instrument, a timer, and a computer. The precision balance measures the mass of the detector activation sheet and then calculates its volume. The activity measuring instrument measures the activity of the activation sheet after activation. The timer records the irradiation time, the waiting time before measurement, and the measurement time. The computer processes the data.

[0091] The embodiments of this application have at least the following beneficial effects: A groove for arranging the activation plates is set on the fuel rod sleeve of the microreactor criticality device, avoiding the need for activation plate carriers and supports. This satisfies the structural requirements of a compact core while preventing disturbances to the neutron field caused by carriers and supports. Activation plates at different detection points can simultaneously measure the relative and absolute neutron flux distribution, reducing the number of reactor startups. Optimizing detection points using neutron flux distribution trends predicted by a high-fidelity neutronics program allows for measurement with a minimal number of detection points. Limiting the activation plate size by setting limits on neutron field perturbations reduces disturbances to the in-reactor neutron field, improving experimental accuracy. The measured neutron energy group is determined by the activation reaction rate percentage, enabling accurate measurement of the complex neutron spectrum of the microreactor.

[0092] Based on the same inventive concept, this application also provides a neutron flux measurement device, applied to a neutron flux measurement system. The neutron flux measurement system includes a fuel rod sleeve for a microreactor critical device; the fuel rod sleeve has M grooves for accommodating a detection activation plate; M is a positive integer. Figure 9 As shown, the device includes a first acquisition module 310 and a first determination module 320.

[0093] The first acquisition module 310 is used to acquire a first correction factor, a first nucleon density, a first detection efficiency, and M pieces of first information corresponding to the M grooves. The first correction factor is a correction factor related to the radiation time of the activated probe in the reactor; the first nucleon density is the nucleon density per unit volume of the activated probe; the first detection efficiency is the detection efficiency of the measuring instrument; the first information includes a first measurement count, a second correction factor, a third correction factor, a first micro-section, and a first volume; the first measurement count is the measurement count of the activated probe in the groove in the measuring instrument; the second correction factor is a correction factor related to the waiting time of the activated probe in the groove from being removed from the reactor to being placed in the measuring instrument; the third correction factor is a correction factor related to the measurement time of the activated probe in the groove in the measuring instrument; the first micro-section is the average micro-section of the nuclear reaction between the material and energy of the activated probe in the groove and neutrons within the neutron energy range that the activated probe can detect; and the first volume is the volume of the activated probe in the groove.

[0094] The first determining module 320 is used to determine the M absolute neutron fluxes corresponding to the M slots based on the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information.

[0095] According to the neutron flux measurement device provided in the embodiments of this application, the fuel rod sleeve is provided with M grooves for accommodating the detection activation plates. That is, grooves for arranging the detection activation plates are provided on the fuel sleeve of the microreactor criticality device, avoiding the need for a carrier and support for the detection activation plates. This satisfies the structural requirements of a compact reactor core while preventing disturbances to the neutron field within the reactor by the carrier and support. Furthermore, by using M pieces of first information corresponding to each of the M grooves, M absolute neutron fluxes corresponding to each of the M grooves are determined. That is, the absolute neutron fluxes at different detection points are measured simultaneously using the detection activation plates, reducing the number of reactor startups.

[0096] In some implementations, the first determining module 320 is specifically used for: Substituting the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information into formula (1), the M absolute neutron fluxes corresponding to the M slots are calculated. Formula (1) includes: (1) in, Let M be the absolute neutron flux at the i-th groove, where i is a positive integer less than or equal to M; The first measurement count corresponds to the i-th groove; For the highest detection efficiency; The first correction factor; This is the second correction factor corresponding to the i-th groove; The third correction factor corresponds to the i-th groove; N is the first nucleon density; This is the i-th first microscopic section; Let i be the first volume.

[0097] In some embodiments, the device further includes: The second acquisition module is used to acquire the relative distribution curve of neutron flux within the microreactor critical device. The second determining module is used to determine the number and location of the grooves based on the relative distribution curve of neutron flux within the reactor.

[0098] In some implementations, the size of the detector activation sheet is related to a first perturbation limit, a first neutron flux level, a first nucleon density, a second micro-section, and the range of neutron energies that the detector activation sheet can detect; Among them, the first perturbation limit is the perturbation limit of the activated sheet on the neutron field in the reactor; the first neutron flux level is the neutron flux level corresponding to neutron energy E; and the second microscopic cross section is the microscopic cross section of the material in the activated sheet undergoing a nuclear reaction with neutrons of neutron energy E.

[0099] In some implementations, the neutron energy range includes an upper limit and a lower limit for neutron energy; The size of the detector activation sheet satisfies formula (2); Formula (2) includes: (2) in, To detect the size of the activated sheet; This is the first disturbance limit; The first neutron flux level; N is the first nucleon density; This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy.

[0100] In some implementations, the energy group of the neutron flux measured by the detector activation sheet is related to a first nucleon density, a first neutron flux level, a second microsection, and a first scale. Wherein, the first neutron flux level is the neutron flux level corresponding to neutron energy E; the second microscopic cross section is the microscopic cross section of the material in the detection activation sheet undergoing a nuclear reaction with neutrons of energy E; and the first ratio is the proportion of the nuclear reaction rate between the detection activation sheet and neutrons with energies within the neutron energy range that the detection activation sheet can detect.

[0101] In some implementations, the energy group of the neutron flux measured by the detector activated sheet satisfies formula (3); Formula (3) includes: (3) Where N is the first nucleon density; This represents the first neutron flux level. This is the second microscopic section; This is the lower limit of neutron energy. This represents the upper limit of neutron energy. The first proportion.

[0102] The neutron flux measurement system provided in this application embodiment can be used to perform the neutron flux measurement method, that is, it has the beneficial effects and implementation methods of the neutron flux measurement method provided in this application embodiment. For details, please refer to the specific description of the neutron flux measurement method in the above embodiment, which will not be repeated here.

[0103] It is understood that the various method embodiments mentioned above in this application can be combined with each other to form combined embodiments without violating the principle and logic. Due to space limitations, this application will not elaborate further. Those skilled in the art will understand that in the above methods of specific implementation, the specific execution order of each step should be determined by its function and possible internal logic.

[0104] Based on the same inventive concept, embodiments of this application provide a neutron flux calibration system, including: The neutron flux measuring device in the above embodiment is used to determine M absolute neutron fluxes; The calibration device is used to calibrate the correspondence curve between the absolute neutron flux and the nuclear detection signal based on M absolute neutron fluxes and multiple nuclear detection signals corresponding to multiple detector activation plates.

[0105] The neutron flux calibration system provided in this application embodiment can be used to execute the neutron flux calibration method, that is, it has the beneficial effects and implementation methods of the neutron flux calibration method provided in this application embodiment. For details, please refer to the specific description of the neutron flux calibration method in the above embodiment. This embodiment will not repeat the description here.

[0106] Figure 10 This is a block diagram of an electronic device provided in an embodiment of this application.

[0107] Reference Figure 10 This application provides an electronic device, which includes: at least one processor 701; at least one memory 702; and one or more I / O interfaces 703 connected between the processor 701 and the memory 702; wherein the memory 702 stores one or more computer programs that can be executed by the at least one processor 701, and the one or more computer programs are executed by the at least one processor 701 to enable the at least one processor 701 to perform the above-described neutron flux measurement method or neutron flux calibration method.

[0108] This application also provides a computer-readable storage medium storing a computer program thereon, wherein the computer program, when executed by a processor / processor core, implements the above-described neutron flux measurement method or neutron flux calibration method. The computer-readable storage medium may be volatile or non-volatile.

[0109] This application also provides a computer program product, including computer-readable code, or a non-volatile computer-readable storage medium carrying computer-readable code. When the computer-readable code is run in the processor of an electronic device, the processor in the electronic device executes the above-described neutron flux measurement method or neutron flux calibration method.

[0110] Those skilled in the art will understand that all or some of the steps, systems, and apparatuses disclosed above, and their functional modules / units, can be implemented as software, firmware, hardware, or suitable combinations thereof. In hardware implementations, the division between functional modules / units mentioned above does not necessarily correspond to the division of physical components; for example, a physical component may have multiple functions, or a function or step may be performed collaboratively by several physical components. Some or all physical components may be implemented as software executed by a processor, such as a central processing unit, digital signal processor, or microprocessor, or as hardware, or as an integrated circuit, such as an application-specific integrated circuit (ASIC). Such software can be distributed on a computer-readable storage medium, which may include computer storage media (or non-transitory media) and communication media (or transient media).

[0111] As is known to those skilled in the art, the term computer storage medium includes volatile and non-volatile, removable and non-removable media implemented in any method or technology for storing information, such as computer-readable program instructions, data structures, program modules, or other data. Computer storage media includes, but is not limited to, random access memory (RAM), read-only memory (ROM), erasable programmable read-only memory (EPROM), static random access memory (SRAM), flash memory or other memory technologies, portable compact disc read-only memory (CD-ROM), digital versatile disc (DVD) or other optical disc storage, magnetic cartridges, magnetic tape, disk storage or other magnetic storage devices, or any other medium that can be used to store desired information and is accessible to a computer. Furthermore, it is known to those skilled in the art that communication media typically contain computer-readable program instructions, data structures, program modules, or other data in modulated data signals such as carrier waves or other transmission mechanisms, and may include any information delivery medium.

[0112] The computer-readable program instructions described herein can be downloaded from computer-readable storage media to various computing / processing devices, or downloaded via a network, such as the Internet, local area network, wide area network, and / or wireless network, to an external computer or external storage device. The network may include copper transmission cables, fiber optic transmission, wireless transmission, routers, firewalls, switches, gateway computers, and / or edge servers. A network adapter card or network interface in each computing / processing device receives the computer-readable program instructions from the network and forwards them to the computer-readable storage media in the respective computing / processing device.

[0113] The computer program instructions used to perform the operations of this application may be assembly instructions, instruction set architecture (ISA) instructions, machine instructions, machine-dependent instructions, microcode, firmware instructions, status setting data, or source code or object code written in any combination of one or more programming languages, including object-oriented programming languages ​​such as Smalltalk, C++, etc., and conventional procedural programming languages ​​such as the "C" language or similar programming languages. The computer-readable program instructions may be executed entirely on the user's computer, partially on the user's computer, as a standalone software package, partially on the user's computer and partially on a remote computer, or entirely on a remote computer or server. In cases involving a remote computer, the remote computer may be connected to the user's computer via any type of network—including a local area network (LAN) or a wide area network (WAN)—or may be connected to an external computer (e.g., via the Internet using an Internet service provider). In some embodiments, electronic circuits, such as programmable logic circuits, field-programmable gate arrays (FPGAs), or programmable logic arrays (PLAs), are personalized by utilizing the status information of the computer-readable program instructions. These electronic circuits can execute the computer-readable program instructions to implement various aspects of this application.

[0114] The computer program product described herein can be implemented specifically through hardware, software, or a combination thereof. In one alternative embodiment, the computer program product is specifically embodied in a computer storage medium; in another alternative embodiment, the computer program product is specifically embodied in a software product, such as a software development kit (SDK), etc.

[0115] Various aspects of this application are described herein with reference to flowchart illustrations and / or block diagrams of methods, apparatus (systems), and computer program products according to embodiments of this application. It should be understood that each block of the flowchart illustrations and / or block diagrams, and combinations of blocks in the flowchart illustrations and / or block diagrams, can be implemented by computer-readable program instructions.

[0116] These computer-readable program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable data processing apparatus to produce a machine such that, when executed by the processor of the computer or other programmable data processing apparatus, they create means for implementing the functions / actions specified in one or more blocks of the flowchart and / or block diagram. These computer-readable program instructions can also be stored in a computer-readable storage medium that causes a computer, programmable data processing apparatus, and / or other device to operate in a particular manner; thus, the computer-readable medium storing the instructions comprises an article of manufacture that includes instructions for implementing aspects of the functions / actions specified in one or more blocks of the flowchart and / or block diagram.

[0117] Computer-readable program instructions may also be loaded onto a computer, other programmable data processing apparatus, or other device to cause a series of operational steps to be performed on the computer, other programmable data processing apparatus, or other device to produce a computer-implemented process, thereby causing the instructions executed on the computer, other programmable data processing apparatus, or other device to perform the functions / actions specified in one or more boxes of a flowchart and / or block diagram.

[0118] The flowcharts and block diagrams in the accompanying drawings illustrate the architecture, functionality, and operation of possible implementations of systems, methods, and computer program products according to various embodiments of this application. In this regard, each block in a flowchart or block diagram may represent a module, segment, or portion of an instruction containing one or more executable instructions for implementing a specified logical function. In some alternative implementations, the functions marked in the blocks may occur in a different order than those marked in the drawings. For example, two consecutive blocks may actually be executed substantially in parallel, and they may sometimes be executed in reverse order, depending on the functions involved. It should also be noted that each block in the block diagrams and / or flowcharts, and combinations of blocks in the block diagrams and / or flowcharts, can be implemented using a dedicated hardware-based system that performs the specified function or action, or using a combination of dedicated hardware and computer instructions.

[0119] Example embodiments have been disclosed herein, and while specific terminology has been used, it is for general illustrative purposes only and should not be construed as limiting. In some instances, it will be apparent to those skilled in the art that features, characteristics, and / or elements described in conjunction with particular embodiments may be used alone, or in combination with features, characteristics, and / or elements described in conjunction with other embodiments, unless otherwise expressly indicated. Therefore, those skilled in the art will understand that various changes in form and detail may be made without departing from the scope of this application as set forth by the appended claims.

Claims

1. A method for measuring neutron flux, characterized in that, An application is made in a neutron flux measurement system, which includes a fuel rod sleeve for a microreactor critical device; the fuel rod sleeve is provided with M grooves for accommodating a detector activation plate; M is a positive integer; The method includes: Obtain the first correction factor, the first nucleon density, the first detection efficiency, and the M first pieces of information corresponding to each of the M grooves; The first correction factor is a correction factor related to the radiation time of the activated detector within the reactor; the first nucleon density is the nucleon density per unit volume of the activated detector; the first detection efficiency is the detection efficiency of the measuring instrument; the first information includes a first measurement count, a second correction factor, a third correction factor, a first microscopic cross-section, and a first volume; the first measurement count is the measurement count of the activated detector within the groove in the measuring instrument; the second correction factor is a correction factor related to the waiting time between the activated detector being removed from the reactor and being placed into the measuring instrument; the third correction factor is a correction factor related to the measurement time of the activated detector within the groove in the measuring instrument; the first microscopic cross-section is the average microscopic cross-section of the material and energy of the activated detector within the groove undergoing nuclear reactions with neutrons within the neutron energy range detectable by the activated detector; the first volume is the volume of the activated detector within the groove. Based on the first correction factor, the first nucleon density, the first detection efficiency, and M pieces of the first information, determine the M absolute neutron fluxes corresponding to the M grooves.

2. The method according to claim 1, characterized in that, The step of determining the M absolute neutron fluxes corresponding to the M slots one-to-one based on the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of the first information includes: Substituting the first correction factor, the first nucleon density, the first detection efficiency, and M pieces of the first information into formula (1), the M absolute neutron fluxes corresponding to the M grooves are calculated. Formula (1) includes: （1） in, Let be the absolute neutron flux at the i-th groove, where i is a positive integer less than or equal to M; The first measurement count corresponds to the i-th groove; This represents the first detection efficiency; This is the first correction factor; This is the second correction factor corresponding to the i-th groove; The third correction factor corresponds to the i-th groove; N is the first nucleon density; This refers to the i-th first microscopic cross-section; Let i be the i-th volume of the first volume.

3. The method according to claim 1, characterized in that, Before obtaining the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of first information corresponding to the M grooves, the method further includes: Obtain the relative distribution curve of the in-pile neutron flux of the microreactor critical device; The number and location of the grooves are determined based on the relative distribution curve of neutron flux within the stack.

4. The method according to claim 1, characterized in that, The size of the detector activation plate is related to the first perturbation limit, the first neutron flux level, the first nucleon density, the second microscopic cross section, and the range of neutron energies that the detector activation plate can detect; Wherein, the first disturbance limit is the disturbance limit of the detector activation sheet on the neutron field in the reactor; the first neutron flux level is the neutron flux level corresponding to neutron energy E; and the second microscopic cross section is the microscopic cross section of the material in the detector activation sheet undergoing a nuclear reaction with neutrons of neutron energy E.

5. The method according to claim 4, characterized in that, The neutron energy range includes an upper limit and a lower limit for neutron energy; The size of the detector activation sheet satisfies formula (2); Formula (2) includes: （2） in, The size of the detector activation sheet; This is the first disturbance limit; N represents the first neutron flux level; N represents the first nucleon density. This is the second microscopic section; This is the lower limit of the neutron energy; This represents the upper limit of the neutron energy.

6. The method according to claim 1, characterized in that, The energy group of neutron flux measured by the activated detector plate is related to the first nucleon density, the first neutron flux level, the second microscopic cross section, and the first ratio; Wherein, the first neutron flux level is the neutron flux level corresponding to neutron energy E; the second microscopic cross section is the microscopic cross section of the material in the detector activation sheet undergoing a nuclear reaction with neutrons of neutron energy E; the first ratio is the proportion of the nuclear reaction rate between the detector activation sheet and neutrons within the neutron energy range that the detector activation sheet can detect.

7. The method according to claim 6, characterized in that, The energy group of the neutron flux measured by the activated detector plate satisfies formula (3); Formula (3) includes: （3） Wherein, N is the first nucleon density; This represents the first neutron flux level; This is the second microscopic section; This is the lower limit of the neutron energy; This refers to the upper limit of the neutron energy. This is the first ratio.

8. A method for calibrating neutron flux, characterized in that, include: According to any one of claims 1 to 7, the neutron flux measurement method determines M absolute neutron fluxes; Based on the M absolute neutron fluxes and the multiple nuclear detection signals corresponding to the multiple detector activation plates, a corresponding curve between the absolute neutron fluxes and the nuclear detection signals is calibrated.

9. A neutron flux measuring device, characterized in that, An application is made in a neutron flux measurement system, which includes a fuel rod sleeve for a microreactor critical device; the fuel rod sleeve is provided with M grooves for accommodating a detector activation plate; M is a positive integer; The device includes: A first acquisition module is used to acquire a first correction factor, a first nucleon density, a first detection efficiency, and M pieces of first information corresponding to each of the M grooves; the first correction factor is a correction factor related to the radiation time of the activated detector within the reactor; the first nucleon density is the nucleon density per unit volume of the activated detector; the first detection efficiency is the detection efficiency of the measuring instrument; the first information includes a first measurement count, a second correction factor, a third correction factor, a first microscopic cross-section, and a first volume; the first measurement count is the measurement count of the activated detector within the groove in the measuring instrument; the second correction factor is a correction factor related to the waiting time from when the activated detector within the groove is removed from the reactor to when it is placed in the measuring instrument; the third correction factor is a correction factor related to the measurement time of the activated detector within the groove in the measuring instrument; the first microscopic cross-section is the average microscopic cross-section of the material and energy of the activated detector within the groove undergoing nuclear reactions with neutrons within the neutron energy range detectable by the activated detector; the first volume is the volume of the activated detector within the groove. The first determining module is used to determine the M absolute neutron fluxes corresponding to the M grooves based on the first correction factor, the first nucleon density, the first detection efficiency, and the M pieces of the first information.

10. A neutron flux calibration system, characterized in that, include: The neutron flux measuring device according to claim 9 is used to determine M absolute neutron fluxes; A calibration device is used to calibrate the correspondence curve between the absolute neutron flux and the nuclear detection signal based on the M absolute neutron fluxes and the multiple nuclear detection signals corresponding to the multiple detector activation plates.

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