Method and device for determining neutron fluence of nuclear reactor

Abstract

The invention discloses a nuclear reactor neutron fluence determination method and device, and relates to the technical field of nuclear reactor core design and safety. Comprising the following steps: acquiring a first neutron fluence rate of a first position in an axial active section of a fuel assembly in a nuclear reactor; acquiring a correction coefficient of a second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient comprises at least one of a space correction coefficient and an energy group correction coefficient, the space correction coefficient is used for correcting attenuation of the neutron fluence rate from the first position to the second position, and the energy group correction coefficient is used for correcting energy group structure deviation between a first energy group structure of the first neutron fluence rate and a second energy group structure of the second position; based on the correction coefficient, correcting the first neutron fluence rate to obtain a second neutron fluence rate of the second position; and determining the neutron fluence of the second position based on the second neutron fluence rate of the second position. According to the invention, the accuracy of neutron fluence prediction outside the axial active section of the fuel assembly can be improved.

Description

Technical Field

[0001] This application belongs to the field of nuclear reactor core design and safety technology, and in particular relates to a method and apparatus for determining the neutron flux in a nuclear reactor. Background Technology

[0002] In the design and operation of nuclear reactors, neutron flux rate is a critical parameter, referring to the number of neutrons passing through a unit area per unit time; its time integral is the neutron flux. The neutron flux outside the axial active section of the fuel assemblies in a nuclear reactor is particularly important. It not only relates to the assessment of irradiation damage to reactor internals and the analysis of heat release from reactor components, but also plays a decisive role in assessing spring irradiation relaxation and accelerated oxidation.

[0003] Currently, the determination of neutron flux mainly relies on numerical simulations using deterministic methods. This method, based on specific theoretical models and assumptions, simulates neutron behavior by numerically solving approximate solutions to the neutron transport equations, thereby estimating the neutron flux and providing data support for reactor design.

[0004] However, the diffusion theory assumptions used in deterministic methods differ significantly from the actual conditions outside the axial active section of the fuel assembly, resulting in a significant difference in neutron flux rates and thus poor accuracy in predicting neutron flux outside the axial active section of the fuel assembly. Summary of the Invention

[0005] This application provides a method and apparatus for determining the neutron flux in a nuclear reactor, which can improve the accuracy of neutron flux prediction outside the axial active section of the fuel assembly.

[0006] One aspect of this application provides a method for determining the neutron flux in a nuclear reactor, including: Obtain the first neutron flux rate at the first location within the axial active section of the fuel assembly in the nuclear reactor; Obtain the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient includes at least one of spatial correction coefficient and energy group correction coefficient, the spatial correction coefficient is used to correct the attenuation of neutron flux rate from the first position to the second position, and the energy group correction coefficient is used to correct the energy group structure deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. Based on the correction factor, the first neutron fluence rate is corrected to obtain the second neutron fluence rate at the second position; The neutron flux at the second position is determined based on the second neutron flux rate at the second position.

[0007] Furthermore, this application also proposes that, before obtaining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, the method further includes: Neutron transport simulation was performed on the axial active section of fuel assemblies in nuclear reactors to construct a neutron behavior model; Based on the neutron behavior model, the simulated neutron flux rate at each position of the axial active section of the fuel assembly under various energy group structures was determined. A first database was constructed based on the simulated neutron fluence at various locations of the axial active section of the fuel assembly under various energy group structures. Obtain the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, including: Based on the first database, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0008] Furthermore, this application also proposes that the correction factor includes a spatial correction factor; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the second energy group structure corresponding to the second position; Obtain the first simulated neutron fluence rate corresponding to the first position and the second energy group structure from the first database, and the second simulated neutron fluence rate corresponding to the second position and the second energy group structure. The spatial correction coefficient for the second position is determined based on the deviation between the second simulated neutron flux rate and the first simulated neutron flux rate.

[0009] Furthermore, this application also proposes that the correction coefficients include energy group correction coefficients; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the first energy group structure corresponding to the first neutron fluence rate, and the second energy group structure corresponding to the second position; Obtain the second simulated neutron fluence rate corresponding to the second position and the second energy group structure from the first database, and the third simulated neutron fluence rate corresponding to the second position and the first energy group structure. The energy group correction coefficients for the second position are determined based on the deviation between the second and third simulated neutron flux rates.

[0010] Furthermore, this application also proposes that the first database includes a sub-database corresponding to the axial active section of the fuel assembly under different moderator densities; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the target moderator density in the axial active section of the fuel assembly; Based on the target sub-database corresponding to the target moderator density, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0011] Furthermore, this application also proposes that the correction coefficients include spatial correction coefficients and energy group correction coefficients; Based on the correction factor, the first neutron fluence rate is corrected to obtain the second neutron fluence rate at the second position, including: The second neutron fluence at the second position is obtained by multiplying the first neutron fluence rate, the space correction coefficient, and the energy group correction coefficient.

[0012] Furthermore, this application also proposes obtaining the first neutron flux rate at a first location within the axial active section of a fuel assembly in a nuclear reactor, including: Obtain the first burnup information of the axial active section of the fuel assembly in the nuclear reactor; Based on the first burnup information, the first neutron flux rate at the first position is obtained from the second database; the second database includes the actual neutron flux rate at each position within the axial active section of the fuel assembly under various burnup information.

[0013] Furthermore, this application also proposes that, before obtaining the first neutron flux rate at the first location from the second database based on the first burnup information, the method further includes: Based on the target strategy, construct a second database; The target strategy is at least one of a deterministic strategy based on diffusion theory assumptions and a flux map experimental strategy.

[0014] Furthermore, this application also proposes that the target strategy is a flux mapping experiment strategy; Based on the target strategy, a second database is constructed, including: Flux map experiments were conducted on the axial active section of the fuel assembly under the second burnup information to obtain the flux map experiment results; the second burnup information can be any type of burnup information, and the flux map experiment results are used to characterize the distribution of neutron flux density in the axial active section of the fuel assembly under the second burnup information. Based on the flux map test results, the actual neutron flux rate at each position in the axial active section of the fuel assembly was determined under the second burnup information. A second database is constructed based on the actual neutron flux rate at each position within the axial active segment of the fuel assembly under each second burnup information.

[0015] In two aspects, this application provides an apparatus for determining the neutron flux in a nuclear reactor, comprising: The flux rate acquisition module is used to acquire the first neutron flux rate at the first position within the axial active section of the fuel assembly in the nuclear reactor. The coefficient acquisition module is used to acquire the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor. The correction coefficient includes at least one of spatial correction coefficient and energy group correction coefficient. The spatial correction coefficient is used to correct the attenuation of neutron flux rate from the first position to the second position, and the energy group correction coefficient is used to correct the energy group structure deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. The flux rate correction module is used to correct the first neutron flux rate based on the correction coefficient to obtain the second neutron flux rate at the second position. The flux determination module is used to determine the neutron flux at the second position based on the second neutron flux rate at the second position.

[0016] A third aspect of the embodiments of this application provides an electronic device, the device comprising: a memory and a program or instructions stored in the memory and executable on a processor, wherein when the program or instructions are executed by the processor, they implement the method for determining the neutron flux in a nuclear reactor as provided in any of the embodiments of the present application described above.

[0017] A fourth aspect of the embodiments of this application provides a readable storage medium on which a program or instructions are stored, and when the program or instructions are executed by a processor, implement the method for determining the neutron flux in a nuclear reactor as provided in any of the embodiments of this application above.

[0018] A fifth aspect of the embodiments of this application provides a computer program product in which instructions, when executed by a processor of an electronic device, cause the electronic device to perform a method for determining the nucleon flux in a nuclear reactor as provided in any of the embodiments of this application described above.

[0019] The method for determining the neutron flux in a nuclear reactor provided in this application first obtains the first neutron flux rate at a first position within the axial active section of the fuel assembly in the nuclear reactor, and a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly. The correction coefficient includes at least one of a spatial correction coefficient and an energy group correction coefficient. The spatial correction coefficient corrects for the decay of the neutron flux rate from the first position to the second position, while the energy group correction coefficient corrects for the deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. Next, the first neutron flux rate is corrected based on the correction coefficient to obtain the second neutron flux rate at the second position. Finally, the neutron flux at the second position is determined based on the second neutron flux rate at the second position. Thus, by introducing correction coefficients, the decay of the neutron flux rate inside and outside the axial active section of the fuel assembly, as well as the error caused by the energy group structure of the first neutron flux rate not meeting the actual requirements of the second position, are fully considered. This makes the prediction of the neutron flux outside the axial active section of the fuel assembly more realistic and improves the accuracy of the neutron flux prediction outside the axial active section of the fuel assembly. Attached Figure Description

[0020] To more clearly illustrate the technical solutions of the embodiments of this application, the accompanying drawings used in the embodiments of this application will be briefly introduced below. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0021] Figure 1 This is a flowchart illustrating a method for determining the neutron flux in a nuclear reactor according to an embodiment of this application; Figure 2 This is a schematic diagram of the structure of an axial active section of a fuel assembly according to an embodiment of this application; Figure 3 This is a schematic diagram illustrating a change in neutron flux according to an embodiment of this application; Figure 4 This is a schematic flowchart illustrating the calculation of neutron flux outside the axial active segment according to an embodiment of this application; Figure 5 This is a schematic diagram of a device for determining the neutron flux in a nuclear reactor according to one embodiment of this application; Figure 6 This is a schematic diagram of a device for determining the neutron flux in a nuclear reactor, provided in one embodiment of this application. Detailed Implementation

[0022] The features and exemplary embodiments of various aspects of this application will be described in detail below. To make the objectives, technical solutions, and advantages of this application clearer, the application will be further described in detail below with reference to the accompanying drawings and specific embodiments. It should be understood that the specific embodiments described herein are only intended to explain this application and not to limit it. For those skilled in the art, this application can be implemented without some of these specific details. The following description of the embodiments is merely to provide a better understanding of this application by illustrating examples.

[0023] It should be noted that, in this document, relational terms such as "first" and "second" are used merely to distinguish one entity or operation from another, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Furthermore, the terms "comprising," "including," or any other variations thereof are intended to cover non-exclusive inclusion, such that a process, method, article, or apparatus that comprises a list of elements includes not only those elements but also other elements not expressly listed, or elements inherent to such a process, method, article, or apparatus. Without further limitations, an element defined by the phrase "comprising..." does not exclude the presence of additional identical elements in the process, method, article, or apparatus that includes said element.

[0024] It should be noted that the acquisition, storage, use, and processing of data in the technical solution of this application all comply with the relevant provisions of national laws and regulations.

[0025] It should be noted that in the embodiments of this application, certain software, components, models and other existing solutions in the industry may be mentioned. These should be regarded as exemplary and are only intended to illustrate the feasibility of implementing the technical solution of this application. However, it does not mean that the applicant has used or necessarily used the solution.

[0026] In traditional methods for determining neutron flux, deterministic methods based on diffusion theory assumptions simulate neutron behavior by numerically solving approximate solutions to the neutron transport equations. However, in the region outside the axial active section of the fuel assembly, due to the increased geometric complexity and significant changes in the neutron energy spectrum, the predicted neutron flux rate outside the active section exhibits a systematic deviation from the actual physical process. This deviation stems from the fact that diffusion theory assumptions cannot accurately describe the anisotropic distribution characteristics of neutron flux outside the active section, and the differences in energy group structure lead to a mismatch in the neutron energy spectrum response function, thus resulting in a lack of physical basis for the neutron flux rate correction model.

[0027] To address the aforementioned issues, this application first identifies that the core root cause of prediction bias in the axial active section region of the fuel assembly using traditional methods lies in the ineffective characterization of spatial decay effects and energy group structure differences. Specifically, abrupt changes in the geometry outside the active section alter neutron transport paths, leading to increased errors in the calculation of the spatial distribution of the flux rate. Simultaneously, changes in the neutron energy group structure and mismatches in the energy group response function cause errors in the flux rate energy spectrum. To address this, this application considers two correction paths: one is to compensate for neutron flux rate decay at locations outside the axial active section using spatial correction coefficients; the other is to eliminate biases between different energy group structures using energy group correction coefficients. Further analysis reveals that using only a single correction method, such as correcting only spatial decay while ignoring energy group differences, still results in distorted energy spectrum integration results; conversely, correcting only the energy group structure without considering spatial decay fails to reflect the impact of changes in the neutron transport path. Therefore, this application proposes combining spatial correction coefficients and energy group correction coefficients, using a product relationship to perform dual correction of the initial neutron flux rate within the active section, thereby simultaneously resolving the issues of spatial decay and energy group structure biases. Furthermore, to ensure the physical accuracy of the correction coefficients, this application selects a probabilistic calculation method to construct a neutron behavior model, thereby determining the spatial correction coefficients and the energy group correction coefficients, avoiding the uncertainty caused by relying on empirical parameters.

[0028] In this regard, such as Figure 1 As shown in the figure, this application provides a flowchart illustrating a method for determining the neutron flux in a nuclear reactor. This method can be applied to electronic devices and may include the following steps S100 to S400: S100, obtain the first neutron flux rate at the first position within the axial active section of the fuel assembly in the nuclear reactor.

[0029] In this step, a nuclear reactor is a device capable of achieving a controlled chain reaction of nuclear fission, releasing enormous amounts of energy through the fission of nuclear fuel. This energy can be used for purposes such as power generation and propulsion. In a nuclear reactor, the fuel assembly is one of the key components; it contains nuclear fuel and is the primary site where the nuclear fission reaction occurs.

[0030] The active section is the region within a fuel assembly where efficient nuclear fission reactions can occur, producing neutrons and energy. Not all parts of the entire fuel assembly can participate efficiently in nuclear fission reactions; only a portion within a specific length range possesses high reactivity, and this range is called the axial active section.

[0031] The first position is a specific point or region located within the axial active section of the fuel assembly, and is the measurement or calculation location used to obtain the first neutron flux rate. For example, the first position could be at the end of the axial active section of the fuel assembly.

[0032] The first neutron flux rate refers to the number of neutrons passing through a unit area per unit time at the first location. It is an important physical quantity for measuring the neutron flux intensity at that location, reflecting the density and flow of neutrons at the first location.

[0033] like Figure 2 The diagram shows a structural schematic of an axial active section of a fuel assembly. The first position is a specific location within the axial active section 210.

[0034] The first neutron flux rate is a crucial physical quantity for measuring the neutron flux intensity at the first location, reflecting the neutron density and flow at that position. Obtaining this data provides a fundamental reference for subsequent derivation and analysis of neutron conditions at other locations within the nuclear reactor, as the distribution and variation of neutrons at different locations within the reactor exhibit certain patterns. The data from the first location can serve as a starting point for studying the neutron characteristics at other locations.

[0035] Specifically, deterministic methods are more accurate in calculating the neutron flux rate at locations within the axial active section of a fuel assembly. Therefore, the neutron flux rate at these locations can be calculated using deterministic methods. Specifically, first, the active section within the fuel assembly capable of undergoing effective nuclear fission reactions and producing neutrons and energy is identified, and its axial range is determined. Then, a specific point or region within this axial active section is selected as the first location. Using deterministic methods, this location is measured or calculated to obtain the number of neutrons passing through a unit area per unit time, i.e., the first neutron flux rate.

[0036] S200, obtain the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient includes at least one of spatial correction coefficient and energy group correction coefficient, the spatial correction coefficient is used to correct the attenuation of neutron flux rate from the first position to the second position, and the energy group correction coefficient is used to correct the energy group structure deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position.

[0037] In this step, the second position is a specific point or region outside the axial active section of the fuel assembly. By obtaining the correction coefficient corresponding to this position and combining it with the first neutron flux rate of the first position, the second neutron flux rate at that position is calculated, thereby determining the neutron flux. For example... Figure 2 As shown, the first position is a specific position in the outer 220 of the axial active segment.

[0038] The correction factor is used to adjust the first neutron fluence rate to account for various influencing factors from the first position to the second position, thereby obtaining a more accurate second neutron fluence rate at the second position. The correction factor includes at least one of the spatial correction factor and the energy group correction factor.

[0039] The spatial correction factor is used to correct for the attenuation of the neutron fluence rate from the first position to the second position. As neutrons propagate from the first position to the second position, they interact with surrounding matter, such as through scattering and absorption, causing the neutron fluence rate to decrease. The spatial correction factor takes into account this spatial attenuation factor and adjusts the initial neutron fluence rate accordingly.

[0040] Energy group correction coefficients are used to correct for energy group structure deviations between the first energy group structure corresponding to the first neutron fluence rate and the second energy group structure corresponding to the second position. Neutrons have different energies, and in nuclear engineering, the energy range of neutrons is typically divided into several energy group structures. Neutrons in different energy groups have different interaction characteristics with matter. The energy group structures of the first and second positions may differ due to their different physical environments, which can affect the measurement and calculation of neutron fluence rates. Energy group correction coefficients are used to eliminate the influence of such energy group structure deviations on neutron fluence rates, so that the corrected second neutron fluence rate more accurately reflects the actual neutron situation at the second position.

[0041] Because the second position differs from the first position in terms of spatial location and energy group structure, the first neutron fluence rate cannot be directly used to represent the neutron situation at the second position. The correction coefficients are used to adjust the first neutron fluence rate to account for various influencing factors from the first to the second position. The spatial correction coefficient considers the spatial attenuation factor during neutron propagation, while the energy group correction coefficient eliminates the influence of energy group structure deviations on the neutron fluence rate, thus obtaining more accurate data that can be used to calculate the second neutron fluence rate at the second position.

[0042] Specifically, a specific point or region outside the axial active section of the fuel assembly is first identified as the second position. Considering the various influencing factors during neutron propagation from the first position to the second position, corresponding correction coefficients are obtained. These correction coefficients include at least one of a spatial correction coefficient and an energy group correction coefficient. For the spatial correction coefficient, the reduction in the number of neutrons (i.e., neutron fluence rate decay) due to scattering and absorption of surrounding matter during neutron propagation from the first position to the second position is analyzed, and the spatial correction coefficient is determined accordingly. For the energy group correction coefficient, due to the different neutron energies, energy group structures are defined in nuclear engineering, and the first and second positions may have different energy group structures due to different physical environments. The impact of this difference on neutron fluence rate measurement and calculation is studied to determine the energy group correction coefficient.

[0043] S300, based on the correction coefficient, corrects the first neutron fluence rate to obtain the second neutron fluence rate at the second position.

[0044] In this step, the second neutron fluence rate is the number of neutrons passing through a unit area per unit time at the second position, obtained after correction. Since the second position may differ from the first position in terms of spatial location and energy group structure, the first neutron fluence rate cannot be directly used to represent the neutron situation at the second position. Therefore, a correction factor is needed to obtain the second neutron fluence rate.

[0045] The differences in spatial location and energy group structure between the second and first positions mean that the first neutron fluence rate cannot directly represent the neutron situation at the second position. Correcting the first neutron fluence rate using correction factors is based on our understanding of the propagation and interaction of neutrons within a nuclear reactor. Spatial correction factors and energy group correction factors adjust the first neutron fluence rate from the perspectives of spatial attenuation and energy group structure differences, respectively, so that the corrected second neutron fluence rate more accurately reflects the actual neutron flux intensity at the second position.

[0046] Specifically, based on the correction coefficients (space correction coefficients and / or energy group correction coefficients) obtained above, the first neutron fluence rate is used as an input parameter and calculated in conjunction with the correction coefficients using the corresponding mathematical model or algorithm. This calculation adjusts the first neutron fluence rate, eliminating the effects of spatial attenuation and energy group structure deviations during the transition from the first to the second position. The final result is the number of neutrons passing through a unit area per unit time at the second position, i.e., the second neutron fluence rate.

[0047] S400, based on the second neutron fluence rate at the second position, determines the neutron fluence at the second position.

[0048] In this step, neutron flux is a physical quantity that measures the total amount of neutron irradiation received at a location over a period of time, and it is related to the neutron flux rate. After determining the second neutron flux rate at the second location, the neutron flux at the second location can be determined by integrating over time, etc. It is of great significance for assessing the degree of neutron irradiation of materials at different locations in a nuclear reactor and predicting changes in material properties.

[0049] Neutron flux rate reflects the number of neutrons passing through a unit area per unit time, while neutron flux takes into account the time factor, providing a more comprehensive description of the neutron irradiation of a location over a period of time. By determining neutron flux based on the second neutron flux rate, accurate data can be provided to assess the degree of neutron irradiation on materials at different locations in a nuclear reactor. Since materials undergo performance changes under neutron irradiation, accurate neutron flux data helps predict these performance changes, thereby ensuring the safe operation and effective management of the nuclear reactor.

[0050] Specifically, after determining the second neutron fluence rate at the second position, mathematical methods such as integration are used to calculate the second neutron fluence rate over the corresponding time period, based on specific needs and time range, thereby obtaining the total amount of neutron irradiation received by the second position over a period of time, i.e., the neutron fluence.

[0051] like Figure 3 The diagram illustrates the variation of neutron flux. The horizontal axis represents time duration, and the vertical axis represents neutron flux. Neutron flux increases with increasing time.

[0052] This application introduces a dual compensation mechanism of spatial correction coefficient and energy group correction coefficient to dynamically adjust the neutron flux data inside the fuel assembly in combination with the transport characteristics and energy group structure differences in the external region. This effectively solves the problem of insufficient neutron flux prediction accuracy outside the active section caused by theoretical assumption deviations in traditional deterministic methods.

[0053] As an example, firstly, during reactor operation, the neutron flux rate at the axial active section end of the fuel assembly is measured using deterministic methods and denoted as the first neutron flux rate. Next, the target location for calculating the neutron flux is determined, for example, a position outside the axial active section of the fuel assembly 150 cm from the core center, designated as the second location. Then, the spatial correction coefficient and energy group correction coefficient corresponding to the second location are extracted from a pre-established correction coefficient database. The spatial correction coefficient reflects the decay of the neutron flux rate over the distance from the first to the second location, while the energy group correction coefficient reflects the change in the neutron energy spectrum between the two locations. Multiplying the obtained first neutron flux rate by the spatial correction coefficient and the energy group correction coefficient yields the second neutron flux rate at the second location. Finally, multiplying the second neutron flux rate by a specific time period gives the neutron flux at the second location.

[0054] This embodiment first obtains the first neutron flux rate at a first position within the axial active section of a fuel assembly in a nuclear reactor, and a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly. The correction coefficient includes at least one of a spatial correction coefficient and an energy group correction coefficient. The spatial correction coefficient corrects for the decay of the neutron flux rate from the first position to the second position, while the energy group correction coefficient corrects for the deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. Next, the first neutron flux rate is corrected based on the correction coefficient to obtain the second neutron flux rate at the second position. Finally, the neutron flux at the second position is determined based on the second neutron flux rate at the second position. Thus, by introducing correction coefficients, the decay of the neutron flux rate inside and outside the axial active section of the fuel assembly, as well as the error caused by the energy group structure of the first neutron flux rate not meeting the actual requirements of the second position, are fully considered. This makes the prediction of the neutron flux outside the axial active section of the fuel assembly more realistic and improves the accuracy of the neutron flux prediction outside the axial active section of the fuel assembly.

[0055] In some of the schemes described above in this application, the determination of the correction coefficient lacks precise data support, resulting in deviations in the corrected neutron flux rate and affecting the accuracy of neutron flux calculation.

[0056] In this regard, this application further proposes that, prior to S200, the method for determining the neutron flux in the nuclear reactor may also include: Neutron transport simulation was performed on the axial active section of fuel assemblies in nuclear reactors to construct a neutron behavior model; Based on the neutron behavior model, the simulated neutron flux rate at each position of the axial active section of the fuel assembly under various energy group structures was determined. A first database was constructed based on the simulated neutron fluence at various locations of the axial active section of the fuel assembly under various energy group structures. S200 includes: Based on the first database, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0057] In this embodiment, the construction of the first database employs a probabilistic deterministic method. The neutron transport simulation uses the Monte Carlo method to accurately track the neutron trajectory. The neutron behavior model includes geometric structure, material composition, and boundary condition parameters. The simulated neutron fluence rate covers all spatial coordinate points within the axial active segment and preset energy group structures (such as 8-group or 2-group). The first database adopts a hierarchical storage structure, establishing a three-dimensional grid index based on spatial coordinates. Each grid node is associated with a fluence rate numerical matrix of different energy group structures. Spatial correction coefficients are calculated by comparing the fluence rate differences of grid nodes at different locations under the same energy group structure, while energy group correction coefficients are determined by comparing the fluence rate differences of grid nodes at the same location under different energy group structures.

[0058] Specifically, a high-fidelity neutron transport simulation is first performed to generate a full-domain dataset containing neutron spatial distribution and energy spectrum characteristics. During the simulation, the fuel rod diameter error is controlled within ±0.1 mm, and the moderator density fluctuation range is limited to ±5 mm. The calculation results were then divided into grid nodes every 2 cm along the axial height, with each grid node storing the fluence rate value for each energy group at its corresponding location. When correction coefficients were needed, spatial correction coefficients were obtained by comparing the fluence rate differences of grid nodes at different locations under the same energy group structure, and energy group correction coefficients were obtained by comparing the fluence rate differences of grid nodes at the same location under different energy group structures.

[0059] As an example, neutron transport simulations are performed on the axial active sections of fuel assemblies in nuclear reactors to construct a neutron behavior model. Specifically, any Monte Carlo software can be used for three-dimensional geometric modeling and neutron transport simulation calculations to obtain the neutron behavior model. The simulation process considers the interactions of neutrons of different energies in materials such as fuel, cladding, and moderators, including scattering, absorption, and fission.

[0060] Based on the neutron behavior model, the simulated neutron flux rate at various locations along the axial active section of the fuel assembly under different energy group structures is determined. For example, 10 axial locations can be selected, and the neutron flux rate at each location can be calculated under different energy group structures, such as those greater than 0.625 eV and those greater than 1 MeV.

[0061] Based on the simulated neutron flux rates at various locations along the axial active section of the fuel assembly under different energy group structures, a first database was constructed. This first database contains information such as location, energy group structure, and neutron flux rate.

[0062] Finally, when obtaining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, the correction coefficient is determined based on the first database. Specifically, the difference analysis can be performed by obtaining the neutron fluence rate corresponding to the required position and energy group structure from the database through interpolation or fitting methods.

[0063] This embodiment achieves accurate prediction of neutron flux at locations outside the axial active section of fuel assemblies in nuclear reactors. By constructing a neutron behavior model and a first database, a reliable data foundation is provided for determining subsequent correction coefficients. This simulation- and database-based approach overcomes the prediction bias of traditional deterministic methods outside the axial active section of fuel assemblies, improving the accuracy and reliability of neutron flux prediction.

[0064] In some of the schemes described above in this application, although the simulated neutron flux rates at various positions of the axial active section of the fuel assembly under various energy group structures are obtained by constructing a neutron behavior model and a first database, the influence of spatial decay of the neutron flux rate is not clearly determined when determining the spatial correction coefficient, resulting in insufficient accuracy of the spatial correction coefficient.

[0065] In this regard, this application further proposes correction factors including spatial correction factors; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the second energy group structure corresponding to the second position; Obtain the first simulated neutron fluence rate corresponding to the first position and the second energy group structure from the first database, and the second simulated neutron fluence rate corresponding to the second position and the second energy group structure. The spatial correction coefficient for the second position is determined based on the deviation between the second simulated neutron flux rate and the first simulated neutron flux rate.

[0066] In this embodiment, obtaining the second energy group structure ensures that the correction process is performed on the neutron fluence rate under the same energy group structure, avoiding additional errors introduced by energy group differences. Both the first and second simulated neutron fluence rates are derived from simulation data under the same energy group structure, directly reflecting the spatial decay law of the neutron fluence rate. The deviation can be calculated as a ratio or difference, for example, dividing the second simulated neutron fluence rate by the first simulated neutron fluence rate to obtain the spatial correction coefficient.

[0067] Specifically, when determining the spatial correction coefficient, the first step is to match the corresponding second energy group structure in the first database with the actual energy group structure at the second position. Then, the first simulated neutron fluence rate at the first position under the second energy group structure and the second simulated neutron fluence rate at the second position under the same energy group structure are extracted from the database. The ratio of these two values ​​is calculated; for example, the ratio of the second simulated neutron fluence rate to the first simulated neutron fluence rate is 0.85. This value directly characterizes the degree of spatial attenuation of neutrons from the first position to the second position. This ratio is used as the spatial correction coefficient and applied to the actual neutron fluence rate at the first position. For example, multiplying the first neutron fluence rate by 0.85 eliminates the error caused by spatial attenuation, yielding the neutron fluence rate at the second position. Because the energy group structure remains consistent, the correction process only reflects the impact of spatial position changes on the neutron fluence rate, avoiding correction bias caused by differences in energy group structure.

[0068] For example, the spatial correction factor can be determined by the following formula (1): (1) In formula (1), Used to characterize spatial correction coefficients The second simulated neutron fluence rate is used to characterize the second position and the second energy group structure. The first simulated neutron fluence rate is used to characterize the first position and the second energy group structure.

[0069] As an example, when determining the spatial correction factor corresponding to the second position outside the axial active section of a fuel assembly in a nuclear reactor based on a first database, the second energy group structure corresponding to the second position is first obtained. For example, the second energy group structure can be the structure of three energy groups: thermal neutron, fast neutron, and moderated neutron.

[0070] Next, the first simulated neutron fluence rate corresponding to the first position and the second energy group structure, and the second simulated neutron fluence rate corresponding to the second position and the second energy group structure, are obtained from the first database. Specifically, the first database stores simulated neutron fluence rate data for different positions and energy group structures. By querying this database, the first simulated neutron fluence rate of the first position under the second energy group structure, and the second simulated neutron fluence rate of the second position under the second energy group structure, can be obtained.

[0071] Furthermore, based on the deviation between the second simulated neutron fluence rate and the first simulated neutron fluence rate, a spatial correction coefficient for the second position is determined. Thus, by comparing the simulated neutron fluence rates at the two positions under the same energy group structure, a correction coefficient reflecting the spatial decay effect can be obtained. For example, the spatial correction coefficient can be obtained by dividing the second simulated neutron fluence rate by the first simulated neutron fluence rate.

[0072] This embodiment enables the accurate acquisition of spatial correction coefficients at specific locations outside the axial active section of a fuel assembly in a nuclear reactor. These spatial correction coefficients reflect the neutron flux rate decay from within the axial active section to its outer edge, thus improving the prediction accuracy of the neutron flux outside the axial active section of the fuel assembly. Furthermore, by comparing simulation data under the same energy group structure, the influence of energy group structure differences is eliminated, resulting in more accurate spatial correction coefficients.

[0073] In some of the above-mentioned schemes in this application, although the simulated neutron flux rates at various positions of the axial active section of the fuel assembly under various energy group structures were obtained by constructing a neutron behavior model and a first database, the deviation of the neutron flux rate between different energy group structures was not clearly determined when determining the energy group correction coefficient, resulting in insufficient accuracy of the energy group correction coefficient.

[0074] In response, this application further proposes that correction coefficients include energy group correction coefficients; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the first energy group structure corresponding to the first neutron fluence rate, and the second energy group structure corresponding to the second position; Obtain the second simulated neutron fluence rate corresponding to the second position and the second energy group structure from the first database, and the third simulated neutron fluence rate corresponding to the second position and the first energy group structure. The energy group correction coefficients for the second position are determined based on the deviation between the second and third simulated neutron flux rates.

[0075] In this embodiment, both the second and third simulated neutron fluence rates correspond to the second position, avoiding additional errors introduced by position differences and directly reflecting the degree of difference between the second position and different energy group structures. The deviation can be calculated as a ratio or difference, for example, dividing the second simulated neutron fluence rate by the third simulated neutron fluence rate to obtain the energy group correction coefficient.

[0076] Specifically, when determining the energy group correction coefficient, firstly, based on the first neutron fluence rate and the actual energy group structure at the second position, the corresponding first and second energy group structures in the first database are matched. Then, the third simulated neutron fluence rate at the second position under the first energy group structure and the second simulated neutron fluence rate at the second position under the second energy group structure are extracted from the database. The ratio of these two values ​​is calculated; for example, the ratio of the second simulated neutron fluence rate to the third simulated neutron fluence rate is 0.85. This value directly characterizes the degree of difference between the first and second energy group structures at the second position. This ratio is used as the energy group correction coefficient and applied to the actual neutron fluence rate at the first position. For example, multiplying the first neutron fluence rate by 0.85 eliminates the error caused by the difference between the energy group structures, thus obtaining the neutron fluence rate at the second position.

[0077] For example, the energy group correction coefficient can be determined by the following formula (2): (2) In formula (2), Used to characterize energy group correction coefficients The second simulated neutron fluence rate is used to characterize the second position and the second energy group structure. The third simulated neutron fluence rate is used to characterize the second position corresponding to the first energy group structure.

[0078] As an example, when determining the energy group correction coefficient corresponding to the second position outside the axial active section of a fuel assembly in a nuclear reactor based on a first database, the first energy group structure corresponding to the first neutron fluence rate and the second energy group structure corresponding to the second position are first obtained. For example, the first and second energy group structures can be the structures of three energy groups: thermal neutrons, fast neutrons, and moderated neutrons.

[0079] Next, the third simulated neutron fluence rate corresponding to the second position and the first energy group structure, and the second simulated neutron fluence rate corresponding to the second position and the second energy group structure are obtained from the first database. Specifically, the first database stores simulated neutron fluence rate data for different positions and energy group structures. By querying this database, the third simulated neutron fluence rate of the second position under the first energy group structure, and the second simulated neutron fluence rate of the second position under the second energy group structure can be obtained.

[0080] Furthermore, based on the deviation between the second and third simulated neutron fluence rates, the energy group correction coefficient for the second position is determined. Thus, by comparing the simulated neutron fluence rates at the same position under different energy group structures, the energy group correction coefficient reflecting the energy group deviation can be obtained. For example, the energy group correction coefficient can be obtained by dividing the second simulated neutron fluence rate by the third simulated neutron fluence rate.

[0081] This embodiment enables the accurate acquisition of energy group correction coefficients at specific locations outside the axial active section of a fuel assembly in a nuclear reactor. These energy group correction coefficients reflect the neutron flux rate deviation at these specific locations outside the axial active section of the fuel assembly under different energy group structures, thus helping to improve the prediction accuracy of the neutron flux outside the axial active section of the fuel assembly.

[0082] In some of the schemes described above in this application, the correction process for neutron flux rate does not take into account the influence of moderator density changes on neutron behavior, which limits the applicability of the correction coefficient under different moderator density conditions and affects the accuracy of the neutron flux rate correction results.

[0083] In this regard, this application further proposes a first database including a sub-database corresponding to the axial active section of the fuel assembly under different moderator densities; Based on the first database, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined, including: Obtain the target moderator density in the axial active section of the fuel assembly; Based on the target sub-database corresponding to the target moderator density, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0084] In this embodiment, the neutron behavior of the axial active section of the fuel assembly is significantly affected by the moderator density, and the neutron flux rate exhibits a nonlinear change with varying moderator density. Sub-databases are established by pre-dividing different moderator density ranges, with each sub-database storing simulated neutron flux rate data within its corresponding density range. The target moderator density is obtained through real-time monitoring or by retrieving operating parameters, and the matching process for the target sub-databases is achieved through density range threshold comparison.

[0085] Specifically, when constructing the neutron behavior model, neutron flux distributions under multiple preset moderator density conditions are calculated through neutron transport simulations, forming a set of sub-databases grouped by density. When the moderator density of the fuel assembly reaches a specific value during actual operation, the sub-database corresponding to that density range is automatically selected. For example, when the moderator density is detected to be between 850-900... When calculating the range, the spatial correction coefficient and energy group correction coefficient are calculated by calling the sub-database data corresponding to the density interval. This process effectively eliminates the simulation data deviation caused by the difference in moderator density, enabling the correction coefficients to accurately reflect the spatial decay law of neutron fluence and the energy group structure change characteristics under the current operating conditions, thereby improving the calculation accuracy of neutron fluence at the second position.

[0086] As an example, the first database includes a sub-database corresponding to the axial active section of the fuel assembly under different moderator densities. When determining the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor based on the first database, the target moderator density of the axial active section of the fuel assembly is first obtained. For example, the target moderator density can be obtained by measurement or estimation as 0.7. Then, based on the target sub-database corresponding to the target moderator density, the correction factor corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined. Specifically, a correction factor corresponding to 0.7 can be found in the first database. The sub-database with the closest moderator density, such as 0.69. The corresponding sub-database is used as the target sub-database. Further, correction coefficient data corresponding to the second position, such as spatial correction coefficients and energy group correction coefficients, are determined based on the target sub-database. This allows for the acquisition of correction coefficients that match the actual moderator density conditions, improving the accuracy of neutron flux determination.

[0087] This embodiment allows for the selection of appropriate sub-databases based on actual moderator density conditions, yielding correction coefficients that more closely reflect actual operating conditions. This database subdivision and selection method based on moderator density effectively improves the accuracy of neutron flux determination, particularly for predicting neutron flux in the region outside the axial active section of the fuel assembly. Furthermore, by considering the influence of moderator density, this method can adapt to the neutron flux determination requirements under different operating conditions, enhancing its applicability and reliability.

[0088] In some of the schemes described above in this application, the correction coefficients include spatial correction coefficients and energy group correction coefficients, but the combination of the two correction coefficients is not specified. This may lead to the neglect of the superposition effect of spatial decay and energy group structure deviation during the correction process, affecting the accuracy of neutron fluence calculation.

[0089] In response, this application further proposes that the correction coefficients include spatial correction coefficients and energy group correction coefficients; The S300 includes: The second neutron fluence at the second position is obtained by multiplying the first neutron fluence rate, the space correction coefficient, and the energy group correction coefficient.

[0090] In this embodiment, the spatial correction coefficient is used to quantify the spatial decay of the neutron fluence rate from the first position to the second position, and the energy group correction coefficient is used to quantify the difference in neutron energy spectrum between the first and second energy group structures. The multiplication operation superimposes the two correction effects to form a comprehensive correction factor. The product of the two can simultaneously reflect the combined influence of spatial position changes and energy group structure differences during neutron transport.

[0091] Specifically, the neutron flux at the first position is calculated based on diffusion theory using experimental or deterministic methods, and its energy group structure is constrained by the measurement method or theoretical assumptions. The neutron energy group structure at the second position changes due to its distance from the active section, and the neutron flux decays with spatial distance. By multiplying the spatial correction coefficient with the energy group correction coefficient, the flux error caused by spatial decay and the energy spectrum deviation caused by differences in energy group structure can be eliminated simultaneously. For example, when the spatial correction coefficient is 0.8 and the energy group correction coefficient is 0.9, the combined correction factor is 0.72, indicating that the neutron flux at the second position needs to be reduced by 0.28 from that at the first position. This product ensures that the corrected second neutron flux simultaneously meets the requirements of spatial distribution and energy group structure matching, thereby improving the accuracy of neutron flux calculations outside the axial active section of the fuel assembly.

[0092] As an example, suppose the first neutron fluence rate is The spatial correction factor is 0.8, and the energy group correction factor is 1.2. When making corrections, these three values ​​are multiplied together: × 0.8 × 1.2 = Therefore, the second neutron fluence rate at the second position is obtained. .

[0093] This embodiment simultaneously considers the effects of spatial location changes and energy group structure changes on neutron flux rate, improving the calculation accuracy of neutron flux rate at the second location. Furthermore, the use of simple multiplication operations for correction makes the calculation process simple and efficient, facilitating the rapid acquisition of corrected neutron flux rate values ​​in practical applications.

[0094] In some of the schemes described above in this application, the influence of changes in burnup state on the neutron flux rate at the first position within the axial active section of the fuel assembly was not considered when obtaining the neutron flux rate. Since the fuel assembly continuously undergoes burnup changes during nuclear reactor operation, the neutron flux rate distribution varies significantly under different burnup states. Directly using fixed burnup parameters would cause the obtained neutron flux rate to deviate from the actual operating conditions.

[0095] In this regard, this application further proposes that S100 may include: Obtain the first burnup information of the axial active section of the fuel assembly in the nuclear reactor; Based on the first burnup information, the first neutron flux rate at the first position is obtained from the second database; the second database includes the actual neutron flux rate at each position within the axial active section of the fuel assembly under various burnup information.

[0096] In this embodiment, the second database establishes a mapping relationship between burnup information and neutron flux rate by pre-storing measured neutron flux rate data at various locations of the axial active segment under different burnup states. The first burnup information includes at least one parameter among burnup depth, fuel assembly running time, or equivalent burnup index. The construction process of the second database can be based on a deterministic strategy of diffusion theory assumptions or a flux map experimental strategy, for example, by arranging a neutron detector array to measure the neutron flux density distribution at different burnup stages and storing the measured data according to burnup parameters.

[0097] Specifically, during reactor operation, the burnup status of the fuel assemblies is monitored and recorded in real time by sensors, serving as the first burnup information. This first burnup information is then input into a second database, which automatically matches historical measured data corresponding to that burnup status and extracts the neutron flux value at a first location. For example, when the fuel assembly burnup depth is detected to reach 30,000... At that time, the measured values ​​of neutron flux rate at all positions within the axial active segment at the burn-out depth are retrieved from the second database, and the value corresponding to the first position is output as the current first neutron flux rate.

[0098] As an example, first obtain the initial burnup information of the axial active section of the fuel assembly in the nuclear reactor. For instance, the cumulative burnup data of the fuel assembly can be obtained through a reactor operation monitoring system, such as... .

[0099] Based on the first burnup information, the first neutron flux rate at the first location is obtained from the second database. Specifically, the second database contains actual neutron flux rate data at various locations within the axial active section of the fuel assembly under various burnup information. By querying this second database, data related to 30... The neutron flux rate corresponding to the burnup, located at the center of the axial active section of the fuel assembly.

[0100] This embodiment enables the rapid and accurate acquisition of neutron flux rates at specific locations within the axial active section of a nuclear reactor fuel assembly. This avoids the need for complex theoretical calculations or experimental measurements each time, thus improving the efficiency of neutron flux acquisition. Furthermore, by utilizing a pre-established database, this approach can obtain neutron flux rate data consistent with actual operating conditions, improving the accuracy and reliability of the data. This provides reliable foundational data for subsequent neutron flux calculations and reactor-related analyses.

[0101] In some of the schemes described above in this application, the second database is constructed using only a deterministic strategy based on diffusion theory assumptions. This results in a discrepancy between the theoretical model and the real physical environment in obtaining the actual neutron flux rate in the axial active section of the fuel assembly, affecting the accuracy of subsequent correction calculations.

[0102] In this regard, this application further proposes a method for determining the neutron flux of a nuclear reactor before obtaining the first neutron flux rate at the first location from the second database based on the first burnup information, which further includes: Based on the target strategy, construct a second database; The target strategy is at least one of a deterministic strategy based on diffusion theory assumptions and a flux map experimental strategy.

[0103] In this embodiment, the target strategy includes two parallel or complementary implementation methods. In the flux mapping experimental strategy, the neutron flux density distribution of the axial active section of the fuel assembly under different burnup information is directly measured through physical experiments, and the experimental results are converted into actual neutron flux rate data and stored in a database. In the deterministic strategy, a mathematical model is established based on diffusion theory assumptions, and the neutron flux rate distribution data under different burnup information is calculated through numerical simulation. The two strategies can be executed independently or jointly, and the experimental data and simulation data are fused through weighted averaging or interpolation methods to form a multidimensional dataset.

[0104] Specifically, when employing a flux mapping experimental strategy, a neutron detector is positioned within the axial active section of the fuel assembly during specific burnout stages. Real-time measurements are used to obtain the spatial distribution data of the neutron flux density. After normalization, the measured data is converted into actual neutron flux rates by incorporating detector efficiency parameters. For different burnout depths, experimental measurements are repeated, and corresponding operating parameters are recorded, ultimately forming an experimental database covering the entire burnout cycle. When both strategies are used in combination, the experimental data is used to calibrate the boundary condition parameters of the deterministic model, while numerical simulation results are used to supplement operating points not covered by the experiments. By establishing a mapping table between burnout information and neutron flux rates, rapid querying of the neutron flux rate at the first position under any burnout state is achieved.

[0105] As an example, when the target strategy is a deterministic strategy based on diffusion theory assumptions, the specific implementation steps are as follows: First, detailed neutron transport calculations are performed on individual fuel assemblies in the reactor core, fully considering factors such as the geometry, material composition, and boundary conditions of the fuel assemblies. Based on this, the minority group constants of the assemblies at different burnout points are output (covering absorption, scattering, fission cross-section, etc., which reflect the neutronics characteristics of the fuel assemblies). Then, these minority group constants characterizing the assemblies are input into the reactor core program, and the diffusion equations are solved using a deterministic method based on diffusion theory assumptions. Combined with the input power information, the neutron flux density distribution of the entire reactor core is calculated. Simultaneously, based on the input burnout information, the burnout equations are continuously solved to update the minority group constant information, simulating the changes in fuel assembly characteristics with burnout, thereby updating the neutron flux distribution and power distribution of the reactor core. Finally, the axial flux rate of the assemblies is reconstructed, thus constructing a second database containing relevant neutron flux rate information.

[0106] This embodiment effectively solves the problem of neutron flux prediction deviation outside the axial active section caused by diffusion theory assumptions. The database built based on measured data can accurately reflect the true distribution characteristics of neutron flux under different burnup conditions, avoiding the error accumulation caused by the simplification of theoretical models. The actual neutron flux data obtained in this way can be directly used for subsequent correction calculations, significantly improving the engineering applicability of the neutron flux prediction results outside the axial active section of the fuel assembly.

[0107] In some of the schemes described above in this application, when constructing the second database based on the deterministic strategy of diffusion theory assumptions, the accuracy of neutron flux prediction is insufficient due to the difference between the diffusion theory assumptions and the actual conditions outside the axial active section of the fuel assembly. In response, this application further proposes a flux mapping experiment strategy as the target strategy; Based on the target strategy, a second database is constructed, including: Flux map experiments were conducted on the axial active section of the fuel assembly under the second burnup information to obtain the flux map experiment results; the second burnup information can be any type of burnup information, and the flux map experiment results are used to characterize the distribution of neutron flux density in the axial active section of the fuel assembly under the second burnup information. Based on the flux diagram test results, the actual neutron flux rate at each position in the axial active section of the fuel assembly under the second burnup information was determined. A second database is constructed based on the actual neutron flux rate at each position within the axial active segment of the fuel assembly under each second burnup information.

[0108] In this embodiment, the flux map experiment obtains the spatial distribution of neutron flux density through actual measurement or simulation, and the specific experimental conditions need to match the target burnup information. The distribution of neutron flux density is recorded by a sensor array or a high-precision neutron detector, and the experimental data is normalized to form the flux map experiment results. The determination of the actual neutron flux rate needs to combine the flux map experiment results with a time integration algorithm, and the flux rate at different locations is supplemented by interpolation or fitting methods. The second database is constructed using a hierarchical storage structure, with burnup information as the index key, and neutron flux rate data stored according to axial position and energy group structure.

[0109] Specifically, the flux mapping experiment was initiated under preset burnup conditions. The axial active section of the fuel assembly was divided into multiple axial regions, each equipped with a neutron detector array. During the experiment, real-time neutron flux density data was collected, and the data underwent noise filtering and normalization to generate two-dimensional or three-dimensional flux distribution maps. Based on the flux distribution maps, the neutron fluence rate at each axial position was calculated through integration, and the results were stored in association with burnup information. When constructing the second database, fluence rate data under different burnup conditions were mapped according to axial coordinates and energy group structure, supporting rapid retrieval via burnup parameters. Thus, the actual neutron flux distribution was directly obtained through the experiment, avoiding errors introduced by theoretical assumptions and improving the accuracy of neutron fluence rate data in the second database.

[0110] As an example, during the operating cycle of a nuclear reactor fuel assembly, a flux mapping test procedure is initiated when the axial active section of the fuel assembly reaches a specific burnup state. Neutron detectors are arranged axially at equal intervals along the axial profile of the active section of the fuel assembly, with the detector array covering the upper and lower boundary extension regions of the active section. During the test, the reactor maintains steady-state operation, and neutron flux density distribution data is acquired in real time through the detector array. The raw measurement signals are converted into a neutron flux density spatial distribution map by an energy spectrum analyzer. After normalization, the neutron fluence rate at the second position outside the active section is obtained by integrating the time function of the neutron flux density, and this data is recorded as the actual neutron fluence rate corresponding to the second burnup information. Multiple tests are performed at different burnup stages of the fuel assembly, ultimately establishing a full-cycle neutron fluence rate database encompassing multiple burnup states.

[0111] This embodiment effectively overcomes the prediction bias of neutron flux rate in the region outside the active section of the deterministic method. The flux map experiment directly obtains neutron flux density distribution data under real operating conditions, avoiding systematic errors introduced by diffusion theory assumptions. In particular, it can accurately reflect the influence of the complex geometry outside the axial active section of the fuel assembly on the neutron transport process, significantly improving the reliability of neutron flux calculation in engineering applications.

[0112] As an example, such as Figure 4The diagram illustrates a flowchart for calculating the neutron flux at the second position outside the axial active zone. First, a deterministic strategy combined with a flux map experimental strategy is used to calculate the neutron flux rate at the first position within the active zone. Then, spatial and energy group correction coefficients are obtained through probabilistic core calculations to correct the neutron flux rate at the first position within the active zone, resulting in the spatially and energy group corrected flux rate at the second position outside the active zone. Finally, the corrected flux rate at the second position is used to calculate the neutron flux at the second position outside the axial active zone. This process integrates deterministic and probabilistic methods, ensuring the comprehensiveness and accuracy of the calculation.

[0113] Based on the method for determining the neutron flux in a nuclear reactor provided in this application, correspondingly, this application also provides specific embodiments of a device for determining the neutron flux in a nuclear reactor.

[0114] Figure 5 A schematic diagram of the structure of a nuclear reactor neutron flux determination device 500 provided in an embodiment of this application is shown. The nuclear reactor neutron flux determination device 500 includes a flux rate acquisition module 510, a coefficient acquisition module 520, a flux rate correction module 530, and a flux determination module 540.

[0115] Flux rate acquisition module 510 is used to acquire the first neutron flux rate at the first position within the axial active section of the fuel assembly in the nuclear reactor. The coefficient acquisition module 520 is used to acquire the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor. The correction coefficient includes at least one of spatial correction coefficient and energy group correction coefficient. The spatial correction coefficient is used to correct the attenuation of neutron flux rate from the first position to the second position, and the energy group correction coefficient is used to correct the energy group structure deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. Flux rate correction module 530 is used to correct the first neutron flux rate based on the correction coefficient to obtain the second neutron flux rate at the second position; Flux determination module 540 is used to determine the neutron fluence at the second position based on the second neutron fluence rate at the second position.

[0116] The reactor neutron flux determination apparatus provided in this application first obtains a first neutron flux rate at a first position within the axial active section of a fuel assembly in the nuclear reactor, and a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly. The correction coefficient includes at least one of a spatial correction coefficient and an energy group correction coefficient. The spatial correction coefficient corrects for the decay of the neutron flux rate from the first position to the second position, while the energy group correction coefficient corrects for the deviation between the first energy group structure corresponding to the first neutron flux rate and the second energy group structure corresponding to the second position. Next, the first neutron flux rate is corrected based on the correction coefficient to obtain the second neutron flux rate at the second position. Finally, the neutron flux at the second position is determined based on the second neutron flux rate at the second position. Thus, by introducing correction coefficients, the decay of the neutron flux rate inside and outside the axial active section of the fuel assembly, as well as the error caused by the energy group structure of the first neutron flux rate not meeting the actual requirements of the second position, are fully considered. This makes the prediction of the neutron flux outside the axial active section of the fuel assembly more realistic and improves the accuracy of the neutron flux prediction outside the axial active section of the fuel assembly.

[0117] Furthermore, this application also proposes that, before obtaining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, the device 500 for determining the quantum flux in the nuclear reactor further includes: The model building module is used to simulate neutron transport in the axial active sections of fuel assemblies in nuclear reactors and build neutron behavior models. The flux rate calculation module is used to determine the simulated neutron flux rate at various positions of the axial active section of the fuel assembly under various energy group structures based on the neutron behavior model. The database construction module is used to build the first database based on the simulated neutron fluence at various positions of the axial active section of the fuel assembly under various energy group structures. Coefficient acquisition module 520 is used for: Based on the first database, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0118] Furthermore, this application also proposes that the correction factor includes a spatial correction factor; Coefficient acquisition module 520 is used for: Obtain the second energy group structure corresponding to the second position; Obtain the first simulated neutron fluence rate corresponding to the first position and the second energy group structure from the first database, and the second simulated neutron fluence rate corresponding to the second position and the second energy group structure. The spatial correction coefficient for the second position is determined based on the deviation between the second simulated neutron flux rate and the first simulated neutron flux rate.

[0119] Furthermore, this application also proposes that the correction coefficients include energy group correction coefficients; Coefficient acquisition module 520 is used for: Obtain the first energy group structure corresponding to the first neutron fluence rate, and the second energy group structure corresponding to the second position; Obtain the second simulated neutron fluence rate corresponding to the second position and the second energy group structure from the first database, and the third simulated neutron fluence rate corresponding to the second position and the first energy group structure. The energy group correction coefficients for the second position are determined based on the deviation between the second and third simulated neutron flux rates.

[0120] Furthermore, this application also proposes that the first database includes a sub-database corresponding to the axial active section of the fuel assembly under different moderator densities; Coefficient acquisition module 520 is used for: Obtain the target moderator density in the axial active section of the fuel assembly; Based on the target sub-database corresponding to the target moderator density, the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is determined.

[0121] Furthermore, this application also proposes that the correction coefficients include spatial correction coefficients and energy group correction coefficients; Injection rate correction module 530, used for: The second neutron fluence at the second position is obtained by multiplying the first neutron fluence rate, the space correction coefficient, and the energy group correction coefficient.

[0122] Furthermore, this application also proposes a volume rate acquisition module 510, comprising the following units: The burnup acquisition unit is used to acquire the first burnup information of the axial active section of the fuel assembly in the nuclear reactor. The flux rate acquisition unit is used to obtain the first neutron flux rate at the first position from the second database based on the first burnup information; the second database includes the actual neutron flux rate at each position in the axial active section of the fuel assembly under various burnup information.

[0123] Furthermore, this application also proposes that, before obtaining the first neutron flux rate of the first location from the second database based on the first burnup information, the neutron flux determination device 500 of the nuclear reactor further includes: The database building module is also used to build a second database based on the target strategy; The target strategy is at least one of a deterministic strategy based on diffusion theory assumptions and a flux map experimental strategy.

[0124] Furthermore, this application also proposes that the target strategy is a flux mapping experiment strategy; Database building modules are used for: Flux map experiments were conducted on the axial active section of the fuel assembly under the second burnup information to obtain the flux map experiment results; the second burnup information can be any type of burnup information, and the flux map experiment results are used to characterize the distribution of neutron flux density in the axial active section of the fuel assembly under the second burnup information. Based on the flux map test results, the actual neutron flux rate at each position in the axial active section of the fuel assembly was determined under the second burnup information. A second database is constructed based on the actual neutron flux rate at each position within the axial active segment of the fuel assembly under each second burnup information.

[0125] Based on the method for determining the neutron flux in a nuclear reactor provided in this application, correspondingly, this application also provides specific embodiments of a device for determining the neutron flux in a nuclear reactor.

[0126] Figure 6 A schematic diagram of the structure of the device for determining the neutron flux in a nuclear reactor provided in an embodiment of this application is shown.

[0127] The device for determining the neutron flux in a nuclear reactor may include a processor 601 and a memory 602 storing computer program instructions.

[0128] Specifically, the processor 601 may include a central processing unit (CPU), an application-specific integrated circuit (ASIC), or one or more integrated circuits that can be configured to implement the embodiments of this application.

[0129] Memory 602 may include mass storage for data or instructions. For example, and not limitingly, memory 602 may include a hard disk drive (HDD), floppy disk drive, flash memory, optical disk, magneto-optical disk, magnetic tape, or Universal Serial Bus (USB) drive, or a combination of two or more of these. Where appropriate, memory 602 may include removable or non-removable (or fixed) media. Where appropriate, memory 602 may be internal to an integrated gateway disaster recovery device. In a particular embodiment, memory 602 is non-volatile solid-state memory.

[0130] The processor 601 reads and executes computer program instructions stored in the memory 602 to implement any of the methods for determining the neutron flux in a nuclear reactor in the above embodiments.

[0131] In one example, the device for determining the neutron flux in a nuclear reactor may further include a communication interface 603 and a bus 610. Wherein, as Figure 6As shown, the processor 601, memory 602, and communication interface 603 are connected through bus 610 and complete communication with each other.

[0132] The communication interface 603 is mainly used to realize communication between various modules, devices, units and / or equipment in the embodiments of this application.

[0133] Bus 610 includes hardware, software, or both, that couples components of a device for determining the quantum flux in a nuclear reactor together. For example, and not as a limitation, the bus may include an Accelerated Graphics Port (AGP) or other graphics bus, an Enhanced Industry Standard Architecture (EISA) bus, a Front Side Bus (FSB), HyperTransport (HT) interconnect, an Industry Standard Architecture (ISA) bus, an Infinite Bandwidth Interconnect, a Low Pin Count (LPC) bus, a memory bus, a Microchannel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a PCI-Express (PCI-X) bus, a Serial Advanced Technology Attachment (SATA) bus, a Video Electronics Standards Association Local (VLB) bus, or other suitable buses, or combinations of two or more of these. Where appropriate, bus 610 may include one or more buses. Although specific buses are described and illustrated in embodiments of this application, any suitable bus or interconnect is contemplated herein.

[0134] Furthermore, in conjunction with the method for determining the neutron flux in a nuclear reactor in the above embodiments, this application embodiment can provide a computer storage medium for implementation. This computer storage medium stores computer program instructions; when these computer program instructions are executed by a processor, they implement any of the methods for determining the neutron flux in a nuclear reactor in the above embodiments.

[0135] In addition, in conjunction with the method for determining the neutron flux in a nuclear reactor in the above embodiments, this application embodiment can provide a computer program product for implementation. When the instructions in the computer program product are executed by the processor of an electronic device, the electronic device performs the method for determining the neutron flux in a nuclear reactor as provided in any aspect of the above embodiments of this application.

[0136] It should be clarified that this application is not limited to the specific configurations and processes described above and shown in the figures. For the sake of brevity, detailed descriptions of known methods are omitted here. In the above embodiments, several specific steps are described and shown as examples. However, the method process of this application is not limited to the specific steps described and shown. Those skilled in the art can make various changes, modifications, and additions, or change the order of steps, after understanding the spirit of this application.

[0137] The functional blocks shown in the above-described structural diagram can be implemented as hardware, software, firmware, or a combination thereof. When implemented in hardware, they can be, for example, electronic circuits, application-specific integrated circuits (ASICs), appropriate firmware, plug-ins, function cards, etc. When implemented in software, the elements of this application are programs or code segments used to perform the required tasks. Programs or code segments can be stored on a machine-readable medium or transmitted over a transmission medium or communication link via data signals carried on a carrier wave. "Machine-readable medium" can include any medium capable of storing or transmitting information. Examples of machine-readable media include electronic circuits, semiconductor memory devices, ROM, flash memory, erasable ROM (EROM), floppy disks, CD-ROMs, optical disks, hard disks, fiber optic media, radio frequency (RF) links, etc. Code segments can be downloaded via computer networks such as the Internet, intranets, etc.

[0138] It should also be noted that the exemplary embodiments mentioned in this application describe methods or systems based on a series of steps or apparatus. However, this application is not limited to the order of the above steps; that is, the steps can be performed in the order mentioned in the embodiments, or in a different order, or several steps can be performed simultaneously.

[0139] The aspects of this disclosure have been described above with reference to flowchart illustrations and / or block diagrams of methods, apparatus (systems), and computer program products according to embodiments of this disclosure. It should be understood that each block in the flowchart illustrations and / or block diagrams, and combinations of blocks in the flowchart illustrations and / or block diagrams, can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable data processing apparatus to produce a machine such that these instructions, executable via the processor of the computer or other programmable data processing apparatus, enable the implementation of the functions / actions specified in one or more blocks of the flowchart illustrations and / or block diagrams. Such a processor can be, but is not limited to, a general-purpose processor, a special-purpose processor, a special application processor, or a field-programmable logic circuit. It is also understood that each block in the block diagrams and / or flowcharts, and combinations of blocks in the block diagrams and / or flowcharts, can also be implemented by special-purpose hardware performing the specified functions or actions, or can be implemented by a combination of special-purpose hardware and computer instructions.

[0140] The above description is merely a specific implementation of this application. Those skilled in the art will clearly understand that, for the sake of convenience and brevity, the specific working processes of the systems, modules, and units described above can be referred to the corresponding processes in the foregoing method embodiments, and will not be repeated here. It should be understood that the protection scope of this application is not limited thereto. Any person skilled in the art can easily conceive of various equivalent modifications or substitutions within the technical scope disclosed in this application, and these modifications or substitutions should all be covered within the protection scope of this application.

Claims

1. A method of determining a neutron fluence in a nuclear reactor, characterized in that, The method comprises: obtaining a first neutron fluence rate at a first position in an axial active section of a fuel assembly in a nuclear reactor; obtaining a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient comprises at least one of a spatial correction coefficient and a group correction coefficient, the spatial correction coefficient is used to correct the attenuation of the neutron fluence rate from the first position to the second position, and the group correction coefficient is used to correct the group structure deviation between a first group structure corresponding to the first neutron fluence rate and a second group structure corresponding to the second position; based on the correction coefficient, correcting the first neutron fluence rate to obtain a second neutron fluence rate at the second position; based on the second neutron fluence rate at the second position, determining the neutron fluence at the second position.

2. The method of claim 1, wherein, Before the method comprises: performing neutron transport simulation on the axial active section of the fuel assembly in the nuclear reactor to construct a neutron behavior model; based on the neutron behavior model, determining simulated neutron fluence rates of each position in the axial active section of the fuel assembly under various group structures; based on the simulated neutron fluence rates of each position in the axial active section of the fuel assembly under various group structures, constructing a first database; the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor is obtained, comprising: based on the first database, determining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor.

3. The method of claim 2, wherein, The correction coefficient comprises the spatial correction coefficient; based on the first database, determining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, comprising: obtaining a second group structure corresponding to the second position; obtaining a first simulated neutron fluence rate corresponding to the first position and the second group structure, and a second simulated neutron fluence rate corresponding to the second position and the second group structure from the first database; based on the deviation between the second simulated neutron fluence rate and the first simulated neutron fluence rate, determining the spatial correction coefficient of the second position.

4. The method of claim 2, wherein, The correction coefficient comprises the group correction coefficient; based on the first database, determining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, comprising: obtaining a first group structure corresponding to the first neutron fluence rate, and a second group structure corresponding to the second position; obtaining a second simulated neutron fluence rate corresponding to the second position and the second group structure, and a third simulated neutron fluence rate corresponding to the second position and the first group structure from the first database; based on the deviation between the second simulated neutron fluence rate and the third simulated neutron fluence rate, determining the group correction coefficient of the second position.

5. The method of claim 2, wherein, The first database comprises a sub-database corresponding to the axial active section of the fuel assembly under different moderator densities; based on the first database, determining the correction coefficient corresponding to the second position outside the axial active section of the fuel assembly in the nuclear reactor, comprising: obtaining a target moderator density of the axial active section of the fuel assembly; Determine a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly in the nuclear reactor based on a target sub-database corresponding to the target moderator density.

6. The method of claim 1, wherein, The correction coefficient includes the spatial correction coefficient and the energy group correction coefficient; The method further includes: Multiply the first neutron fluence rate, the spatial correction coefficient, and the energy group correction coefficient to obtain the second neutron fluence rate at the second position.

7. The method according to any one of claims 1 to 6, characterized in that, The method further includes: Obtain first burnup information of the axial active section of the fuel assembly in the nuclear reactor; Based on the first burnup information, obtain the first neutron fluence rate at the first position from a second database; the second database includes actual neutron fluence rates of various positions in the axial active section of the fuel assembly under various burnup information.

8. The method of claim 7, wherein, The method further includes: Based on a target strategy, construct the second database; The target strategy is at least one of a deterministic strategy based on diffusion theory assumptions and a flux map test strategy.

9. The method of claim 8, wherein, The target strategy is the flux map test strategy. Based on the target strategy, constructing the second database includes: Perform a flux map test on the axial active section of the fuel assembly under second burnup information to obtain a flux map test result; the second burnup information is any kind of burnup information, and the flux map test result is used to represent the distribution of the neutron flux density of the axial active section of the fuel assembly under the second burnup information; Based on the flux map test result, determine actual neutron fluence rates of various positions in the axial active section of the fuel assembly under the second burnup information; Based on the actual neutron fluence rates of various positions in the axial active section of the fuel assembly under each of the second burnup information, construct the second database.

10. A device for determining the neutron fluence in a nuclear reactor, characterized in that The device includes: A fluence rate obtaining module configured to obtain a first neutron fluence rate at a first position in an axial active section of a fuel assembly in a nuclear reactor; A coefficient obtaining module configured to obtain a correction coefficient corresponding to a second position outside the axial active section of the fuel assembly in the nuclear reactor; the correction coefficient includes at least one of a spatial correction coefficient and an energy group correction coefficient, the spatial correction coefficient is used to correct the attenuation of the neutron fluence rate from the first position to the second position, and the energy group correction coefficient is used to correct the energy group structure deviation between a first energy group structure corresponding to the first neutron fluence rate and a second energy group structure corresponding to the second position; A fluence rate correction module configured to correct the first neutron fluence rate based on the correction coefficient to obtain a second neutron fluence rate at the second position; A fluence determining module configured to determine the neutron fluence at the second position based on the second neutron fluence rate at the second position.

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