Method and device for evaluating neutronics parameters of a nuclear reactor core

Abstract

This application discloses a method and apparatus for evaluating neutronics parameters of a nuclear reactor core, belonging to the field of nuclear reactor technology. The method includes: determining the fuel rod insertion positions of a target fuel assembly in the core to be fitted with old fuel rods; constructing a first core model of the core based on information about the target fuel assembly; in response to inserting absorber rods at the fuel rod insertion positions, replacing the original fuel rods at those positions in the target fuel assembly with absorber rods, obtaining information about the replaced target fuel assembly, and constructing a second core model of the core based on the replaced target fuel assembly information; and determining the neutronics parameter evaluation results after inserting old fuel rods into the core based on the first and second core models. This application can avoid the problems of secondary development of nuclear design programs and complex modeling of fuel assemblies for fuel rod replacement, while making the neutronics parameter evaluation results more inclusive, thereby improving the conservatism of the core neutronics parameter evaluation results.

Description

Technical Field

[0001] This application belongs to the field of nuclear reactor technology, and in particular relates to a method and apparatus for evaluating neutronics parameters of a nuclear reactor core. Background Technology

[0002] In the development of nuclear reactor fuel, to verify the fuel's performance under high burnup conditions, fuel rods that have undergone multiple irradiation cycles need to be removed from old fuel assemblies and inserted into new fuel assemblies or fuel assemblies with shallower burnup for continued irradiation to achieve even higher burnup. This process is called fuel assembly replacement. After fuel assembly replacement, its impact on the reactor core neutronics parameters must be evaluated to ensure the safe operation of the nuclear reactor.

[0003] In related technologies, the evaluation of neutronics parameters of nuclear reactor cores widely adopts a two-step procedure. This involves reconstructing a new fuel assembly model containing the nuclide composition of the target fuel rods after the fuel assembly is replaced, and then performing core calculations. Most existing nuclear design programs do not have the function of modeling fuel rod replacement assemblies, requiring secondary development, which is a complex process. Furthermore, after reconstructing the fuel assembly, the calculation of the effective fuel temperature and the burnup of pseudo-nuclides may be inaccurate, resulting in non-conservative neutronics parameter evaluation results. In other words, the evaluation results are biased towards the unsafe side, failing to provide sufficient safety margins and making it difficult to meet the stringent requirements of nuclear safety analysis. Summary of the Invention

[0004] This application provides a method and apparatus for evaluating neutron parameters of a nuclear reactor core. It can improve the conservatism of the evaluation results of neutron parameters by avoiding the complex problems of secondary development of nuclear design programs and modeling of fuel assemblies for rod replacement.

[0005] In a first aspect, embodiments of this application provide a method for evaluating neutronics parameters of a nuclear reactor core, the method comprising: Determine the fuel rod insertion position of the target fuel assembly in the reactor core to be fitted with the old fuel rods; Based on the information of the target fuel assembly, a first core model of the reactor core is constructed; In response to inserting an absorber rod at the fuel rod insertion position, the original fuel rod in the target fuel assembly located at the fuel rod insertion position is replaced with the absorber rod to obtain the target fuel assembly information after replacement, and a second core model of the reactor core is constructed based on the target fuel assembly information after replacement. Based on the first core model and the second core model, the evaluation results of the neutronics parameters after the core is fitted with old fuel rods are determined.

[0006] Secondly, embodiments of this application provide a device for evaluating neutronics parameters of a nuclear reactor core, the device comprising: The information determination module is used to determine the fuel rod insertion position of the target fuel assembly in the reactor core to be inserted with the old fuel rods; The first construction module is used to construct a first core model of the reactor core based on the information of the target fuel assembly; The second construction module is used to replace the original fuel rod located at the fuel rod insertion position in the target fuel assembly with the absorber rod in response to inserting the absorber rod at the fuel rod insertion position, obtain the target fuel assembly information after replacement, and construct the second core model of the reactor core based on the target fuel assembly information after replacement. The evaluation and determination module is used to determine the neutronics parameter evaluation results of the reactor core after inserting old fuel rods, based on the first core model and the second core model.

[0007] Thirdly, embodiments of this application provide a device for evaluating neutronics parameters of a nuclear reactor core, the device comprising: Processor and memory storing computer program instructions; When the processor executes the computer program instructions, it implements the neutronics parameter evaluation method for the nuclear reactor core as described in the first aspect.

[0008] Fourthly, embodiments of this application provide a computer storage medium storing computer program instructions, which, when executed by a processor, implement the neutronics parameter evaluation method for a nuclear reactor core as described in the first aspect.

[0009] Fifthly, embodiments of this application provide a computer program product in which instructions, when executed by a processor of an electronic device, cause the electronic device to perform the neutronics parameter evaluation method for a nuclear reactor core as described in the first aspect.

[0010] In this embodiment, by determining the fuel rod insertion positions of the target fuel assembly to be fitted with old fuel rods in the reactor core, a first core model of the reactor core is constructed based on the information of the target fuel assembly. In response to inserting absorber rods at the fuel rod insertion positions, the original fuel rods at the fuel rod insertion positions in the target fuel assembly are replaced with absorber rods, obtaining information about the replaced target fuel assembly. Based on the replaced target fuel assembly information, a second core model of the reactor core is constructed. The neutronics parameter evaluation results after the old fuel rods are inserted are determined based on the first and second core models. Thus, by constructing a first core model assuming the insertion positions are fitted with more reactive fuel rods and a second core model assuming less reactive absorber rods, the neutronics parameters of the reactor core after the old fuel rods are inserted are analyzed envelopingly using these two models. This avoids the complexities of secondary development of nuclear design procedures and fuel assembly modeling for fuel rod replacement, while making the neutronics parameter evaluation results more comprehensive and improving the conservatism of the neutronics parameter evaluation results. Attached Figure Description

[0011] 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.

[0012] Figure 1 This is a flowchart illustrating the method for evaluating neutronics parameters of a nuclear reactor core provided in an embodiment of this application. Figure 2 This is a schematic diagram of the rod replacement process provided in an embodiment of this application; Figure 3 This is a schematic diagram of the implementation process of the neutronics parameter evaluation method for nuclear reactor cores provided in the embodiments of this application; Figure 4 This is a schematic diagram of the structure of the neutronics parameter evaluation device for a nuclear reactor core provided in the embodiments of this application; Figure 5 This is a schematic diagram of the structure of the neutronics parameter evaluation device for the nuclear reactor core provided in the embodiments of this application. Detailed Implementation

[0013] 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.

[0014] 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.

[0015] In all specific embodiments of this application, when processing data related to user identity or characteristics, such as user information, user behavior data, user historical data, and user location information, user permission or consent is obtained first. Furthermore, the collection, use, and processing of this data comply with relevant laws, regulations, and standards. Additionally, when embodiments of this application require access to sensitive personal information, separate permission or consent from the user is obtained through pop-ups or redirects to confirmation pages. Only after obtaining the user's separate permission or consent is the necessary user-related data required for the proper functioning of these embodiments obtained.

[0016] See Figure 1 , Figure 1 This is one of the flowcharts for the evaluation method of neutronics parameters of a nuclear reactor core provided in the embodiments of this application.

[0017] In reactor fuel development, to improve fuel burnup limits, high burnup testing is required to confirm its performance under high burnup conditions. A widely adopted method involves removing fuel rods that have undergone multiple irradiation cycles from old fuel assemblies and inserting them into new fuel assemblies or fuel assemblies with shallower burnup. The fuel is then irradiated again in the reactor for another cycle to achieve even higher burnup. The rod replacement process is as follows: Figure 2 As shown (A is the old component, B is the new component or a fuel component with shallower burnout).

[0018] Currently, reactor neutronics parameter calculations widely employ a two-step procedure. In assessing the impact of fuel assembly replacement on neutronics parameters, this procedure requires modeling the replaced fuel assembly B before performing core calculations to confirm the impact of the replacement on core neutronics parameters. The assembly modeling process requires reconstructing a new fuel assembly by outputting the nuclide composition of the fuel rods inserted into fuel assembly B (the target fuel rod) and the nuclide composition of fuel rods in other locations within fuel assembly B. This process is complex, as most nuclear design programs lack this functionality and require secondary development. Furthermore, after reconstructing the new fuel assembly, related calculations such as effective fuel temperature and pseudonuclide burnup may be inaccurate.

[0019] Existing testing methods involve removing fuel rods that have already undergone multiple irradiations from old fuel assemblies and inserting them into new fuel assemblies or fuel assemblies with shallower burnout. These assemblies are then irradiated again in the reactor for another cycle to achieve higher burnout. However, some reactor nuclear design programs lack the ability to model fuel rod replacement assemblies or have complex fuel rod replacement calculations, leading to complex neutronics parameter calculations or the inapplicability of existing calculation uncertainties. Consequently, the evaluation results of neutronics parameters may not be conservative.

[0020] To address the problems of the prior art, this application provides a method and apparatus for evaluating neutronics parameters of a nuclear reactor core. The method for evaluating neutronics parameters of a nuclear reactor core provided in this application will be described below.

[0021] The neutronics parameter evaluation method for nuclear reactor cores according to the embodiments of this application can be applied to electronic devices. Specifically, this method can be executed by an electronic device, or by components of the electronic device, such as a processor, chip, or chip system. It can also be implemented by a logic module or software that implements all or part of the functions of the electronic device. In practical applications, the electronic device can be a terminal, server, service platform, cloud, distributed system, Internet of Things (IoT), vehicle network system, etc. Furthermore, the terminal can be a smartphone, tablet computer, laptop computer, desktop computer, etc.

[0022] The following description, in conjunction with the accompanying drawings, details the method for evaluating neutronics parameters of a nuclear reactor core provided in this application through several embodiments and application scenarios.

[0023] See Figure 1 , Figure 1 This is one of the flowcharts for the neutronics parameter evaluation method of a nuclear reactor core provided in the embodiments of this application. For example... Figure 1 As shown, the method for evaluating neutronics parameters of a nuclear reactor core may include the following steps: Step 101: Determine the fuel rod insertion position of the target fuel assembly in the reactor core to be fitted with the old fuel rods.

[0024] In this step, the target fuel assembly can be the fuel assembly in the reactor core that is to be fitted with old fuel rods, or it can be set according to the actual situation, without specific limitations here.

[0025] A new fuel assembly that may contain multiple fuel rod grids is defined as a fuel rod insertion position within the target fuel assembly, specifying the exact location within the assembly where old fuel rods are inserted. This fuel rod insertion position may include grid coordinates and rod position numbers within the target fuel assembly, used to identify the insertion location of the old fuel rods.

[0026] It should be noted that the fuel rod insertion position can also be set according to the actual situation, and no specific limitation is made here.

[0027] Step 102: Construct the first core model of the reactor core based on the information of the target fuel assembly.

[0028] In this step, the information for the target fuel assembly may include fuel type, enrichment level, geometry, etc.

[0029] Furthermore, based on the information of the target fuel assembly, a first core model is constructed. The first core model is a model of the target fuel assembly without any fuel rods inserted, that is, using all the original fuel rods of the target fuel assembly.

[0030] In some embodiments, in response to not inserting fuel rods at the fuel rod insertion positions, a first core model of the reactor core may be constructed based on information about the target fuel assembly.

[0031] It should be noted that the original fuel rods of the target fuel assembly can be new fuel rods if the target fuel assembly is a new assembly; if the target fuel assembly is an old assembly that has already been burned through one or two cycles, the original fuel rods can be old fuel rods that have already been irradiated.

[0032] In some embodiments, the fuel consumption of the original fuel rod is less than the fuel consumption of the old fuel rod to be inserted.

[0033] The first core model characterizes the neutron state when new fuel rods or shallower burnout fuel rods are used at the same fuel rod insertion position, that is, the core neutron state with greater positive reactivity at the insertion position.

[0034] Step 103: In response to inserting absorber rods at the fuel rod insertion positions, replace the original fuel rods in the target fuel assembly located at the fuel rod insertion positions with the absorber rods, obtain the information of the replaced target fuel assembly, and construct the second core model of the reactor core based on the information of the replaced target fuel assembly.

[0035] In this step, the reactivity of the old fuel rods is significantly lower than that of the original fuel rods due to irradiation, thus contributing negative reactivity to the overall reactor core. To simplify calculations and conservatively evaluate this negative reactivity effect, embodiments of this application can use absorber rods to equivalently or encompass the physical effects of the old fuel rods.

[0036] In some implementations, the absorber rod can be made of a material with strong neutron absorption, such as stainless steel or cadmium, to simulate the strong absorption characteristics of old fuel rods due to burnup.

[0037] Furthermore, in response to the insertion of absorber rods at the fuel rod insertion positions, the geometric and material definitions of the original fuel rods at the corresponding insertion positions in the target fuel assembly can be deleted and replaced with the geometric and material definitions of the absorber rods. This updates the assembly information, yields the replaced target fuel assembly information, and a second core model is constructed based on this. The second core model reflects the core physical state after the introduction of strong absorbers.

[0038] The second core model is used to characterize the neutronics state when absorber rods are used at the same fuel rod insertion position, that is, the core neutronics state when there is greater negative reactivity at the insertion position.

[0039] Step 104: Based on the first core model and the second core model, determine the neutronics parameter evaluation results after the core is fitted with old fuel rods.

[0040] In this step, after obtaining the first core model and the second core model, neutronics calculation programs, such as diffusion calculation programs and transport calculation programs, can be used to solve the two models respectively to obtain the neutronics parameter calculation results corresponding to each core model; alternatively, burnup calculations can be performed on the core models separately, and based on the neutronics parameter calculation conditions in the final safety analysis report, the calculation results of each neutronics parameter can be obtained.

[0041] The neutronics parameter calculation results can include parameter values ​​corresponding to at least one category of neutronics parameters, and can include neutronics parameter categories such as moderator density coefficient and maximum boron differential value.

[0042] Furthermore, the neutronics characteristics of the old fuel rods are between those of the original fuel rods (i.e., the new fuel rods / fuel rods with shallower burnout) and the absorber rods in the target fuel assembly. That is, the fission capability of the old fuel rods is weaker than that of the original fuel rods but stronger than that of the absorber rods, and its absorption capability is stronger than that of the original fuel rods but weaker than that of the absorber rods. It can be seen that the actual neutronics parameters of the reactor core after the old fuel rods are inserted will be between the calculation results of the first core model and the second core model.

[0043] In some embodiments, the neutronics parameter evaluation results after the reactor core is fitted with old fuel rods can be determined based on the calculation results of the first core model and the second core model through operations such as comparison, interpolation or equivalent processing.

[0044] For example, to further test fuel consumption, it is necessary to replace one fuel rod b with a fuel consumption of 30,000 MWd / tU in fuel assembly A with a fuel rod a with a fuel consumption of 40,000 MWd / tU, and name the replaced fuel assembly B.

[0045] In fuel assembly A, fuel rod b is replaced with absorber rod, and the resulting fuel assembly is named fuel assembly C.

[0046] The fuel assembly model in the first core model is fuel assembly A; The fuel assembly model in the second core model is fuel assembly C; The results from the first and second core models are used to encapsulate the calculation results of the actual core (fuel assembly B).

[0047] In addition, fuel assembly A may be a new assembly or an old assembly that has already been burned through one or two cycles.

[0048] The neutronics parameter evaluation method for a nuclear reactor core in this embodiment determines the fuel rod insertion position information of the target fuel assembly to be fitted with old fuel rods in the core. Based on the target fuel assembly information, a first core model of the core is constructed. In response to the insertion of absorber rods at the fuel rod insertion positions, the original fuel rods at the fuel rod insertion positions in the target fuel assembly are replaced with absorber rods, obtaining the information of the replaced target fuel assembly. Based on the replaced target fuel assembly information, a second core model of the core is constructed. Based on the first core model and the second core model, the neutronics parameter evaluation results after the old fuel rods are inserted are determined. Thus, by constructing a first core model assuming that more reactive fuel rods are inserted at the insertion positions, and a second core model assuming less reactive absorber rods are inserted, the neutronics parameters after the old fuel rods are inserted are analyzed envelopingly using these two models. This avoids the problems of secondary development of nuclear design programs and complex modeling of fuel assembly replacement, while making the neutronics parameter evaluation results more inclusive and improving the conservatism of the neutronics parameter evaluation results.

[0049] In one embodiment of this application, the target core model can be determined through the following steps: Step 201: Construct a target fuel assembly model based on the target information; Step 202: Construct the target reactor core model based on the target fuel assembly model; Wherein, the target core model is either the first core model or the second core model; In response to the target core model being the first core model, the target information is the information of the target fuel assembly; In response to the target core model being the second core model, the target information is the replaced target fuel assembly information.

[0050] In some embodiments, the target fuel assembly model is a first fuel assembly model or a second fuel assembly model, wherein the fuel rods in the first fuel assembly model are the original fuel rods of the target fuel assembly; and the fuel rods at the intercalation positions in the second fuel assembly model are absorber rods; step 201 may include the following steps: Based on the target fuel assembly model, determine the target fuel assembly library; Based on the target fuel assembly library, construct the target reactor core model; Wherein, in response to the target core model being the first core model, the target fuel assembly model being the first fuel assembly model; In response to the target core model being the second core model, the target fuel assembly model being the second fuel assembly model.

[0051] In this embodiment, the fuel assembly model can refer to a physical model that provides a detailed description of the geometry, material composition, and cell arrangement of the fuel assembly. It may include the arrangement of fuel rods, enrichment information, and the type of filler at specific locations.

[0052] The fuel assembly library is a set of macroscopic cross-sectional parameters obtained by performing neutron transport calculations on the fuel assembly model using a component computation program. This set of macroscopic cross-sectional parameters can include key physical data such as neutron diffusion parameters and reaction rate cross-sections of the fuel assembly under different burn-up depths and temperature conditions. During the subsequent construction of the core model, the core computation program calls the fuel assembly library to assign the physical properties of the fuel assemblies to their corresponding positions in the core mesh, thereby achieving neutronics simulation of the entire core.

[0053] In some embodiments, a target fuel assembly model can be constructed based on the target information, and then a target reactor core model can be constructed based on the target fuel assembly model.

[0054] Specifically, when the target core model is the first core model, the target information is the information of the target fuel assembly. That is, the target fuel assembly model, i.e., the first fuel assembly model, can be established based on the information of the target fuel assembly, such as fuel type, enrichment, geometry, and grid arrangement. In the first fuel assembly model, all fuel rods are the original fuel rods of the target fuel assembly. Based on the first fuel assembly model, a target fuel assembly library corresponding to the first fuel assembly model is generated, and the first core model is constructed based on the target fuel assembly library corresponding to the first fuel assembly model.

[0055] When the target core model is the second core model, the target information is the replaced target fuel assembly information. Therefore, based on the replaced target fuel assembly information, a target fuel assembly model, i.e., the second fuel assembly model, is established. In the second fuel assembly model, the fuel rods at the fuel rod insertion positions of the target fuel assembly are replaced with absorber rods. Then, based on the second fuel assembly model, a target fuel assembly library corresponding to the second fuel assembly model is generated, and the second core model is constructed based on this target fuel assembly library.

[0056] In this embodiment, based on the fact that the neutronics characteristics of the old fuel rods fall between those of the original fuel rods and the absorber rods, the two core models mentioned above are used as boundary conditions to quickly obtain the predicted values ​​of the core neutronics parameters after the old fuel rods are inserted. This avoids the tedious process of recalculating the burnup of each old fuel rod, greatly improving the evaluation efficiency.

[0057] In one embodiment of this application, determining the target fuel assembly library based on the target fuel assembly model may include the following: The ground-state fuel consumption of the target fuel assembly model is calculated to obtain ground-state fuel consumption data; Based on the ground-state fuel consumption data, perturbation calculations are performed on the target fuel assembly model to obtain perturbation parameters; Based on the ground-state fuel consumption data, branch calculations are performed on the target fuel assembly model to obtain branch parameters; The target fuel assembly library is determined based on the ground-state fuel consumption data, the disturbance parameters, and the branch parameters.

[0058] In this embodiment, ground-state burnup calculation refers to the time-varying neutronics simulation calculation of the target fuel assembly model under standard design conditions such as rated power, rated temperature, and rated pressure. This simulation is used to model the burnup evolution process of the fuel assembly from its initial state to its unloading state. Ground-state burnup data refers to the set of fundamental physical parameters obtained through ground-state burnup calculation, including macroscopic and microscopic neutron cross sections and nuclide densities that vary with burnup depth. These parameters serve as a benchmark for characterizing the physical properties of the fuel assembly under standard operating conditions.

[0059] Perturbation calculations refer to the calculations performed on key physical parameters of a target fuel assembly model, such as fuel temperature and moderator density, at specific burnout nodes based on ground-state burnout data. The aim is to obtain information on the sensitivity of these physical parameters to minute changes in operating conditions. Perturbation parameters can refer to feedback coefficients or derivatives determined through perturbation calculations, used to characterize the neutron cross-section as a function of operating parameters, and to describe the sensitivity of physical parameters to minute changes in operating conditions.

[0060] Branching computation refers to the calculation of a target fuel assembly model after discrete state switching at specific burnup nodes, based on ground-state burnup data. Discrete states include control rod insertion and withdrawal state switching, core cold and hot state switching, or significant changes in coolant boron concentration. Branch parameters refer to neutron cross-section data determined through branching computation, corresponding to specific discrete state combinations, used to characterize the physical properties of the fuel assembly in discrete states such as control rod insertion.

[0061] The target fuel assembly library is determined based on ground-state burnup data, perturbation parameters, and branch parameters.

[0062] In this embodiment, by combining ground-state burnup data, perturbation parameters, and branch parameters, a high-fidelity target fuel assembly library covering the entire life cycle and multiple operating conditions can be constructed, thereby ensuring the accuracy of subsequent core model calculations.

[0063] In one embodiment of this application, constructing the target reactor core model based on the target fuel assembly library may include the following: Obtain the fuel assembly loading scheme for the reactor core; Based on the fuel assembly loading scheme and the target fuel assembly library, the target reactor core model is constructed.

[0064] In this embodiment, the fuel assembly loading scheme can refer to the location and arrangement information of all fuel assemblies inside the reactor core, which may include the specific coordinates of each fuel assembly in the reactor core, assembly type, burn-up depth, whether or not old fuel rods are inserted and their insertion positions, etc.

[0065] Using the core computation program, the fuel assembly loading scheme is read, and the physical parameters in the generated target fuel assembly library are mapped to the corresponding positions in the core geometric mesh, thereby establishing a complete core neutronics computation model.

[0066] In some embodiments, a target core model can be constructed by combining relevant information such as cycle length.

[0067] It should be noted that mapping using the fuel assembly library generated based on the first fuel assembly model will result in the construction of the first core model; mapping using the fuel assembly library generated based on the second fuel assembly model will result in the construction of the second core model.

[0068] In this embodiment, the fuel assembly loading scheme of the reactor core is obtained, and the corresponding component model is called from the target fuel assembly library for assembly according to the loading scheme. This approach allows the reactor core modeling to flexibly adapt to different loading configurations. Both the first and second reactor core models can be quickly generated under a unified modeling framework, ensuring the comparability between different models and laying a solid foundation for subsequent comparison and evaluation of neutronics parameters.

[0069] In one embodiment of this application, step 104 may include the following steps: Step 301: Perform neutronics parameter calculations on the first core model and the second core model respectively to obtain the calculation results of the first neutronics parameter corresponding to the first core model and the calculation results of the second neutronics parameter corresponding to the second core model; Step 302: Based on the calculation results of the first neutron parameters and the calculation results of the second neutron parameters, generate the neutron parameter evaluation results after the reactor core is fitted with old fuel rods.

[0070] In this embodiment, neutronics parameters can be calculated for the first core model and the second core model using a core physics calculation program, such as a three-dimensional core diffusion program, to obtain the calculation results of the first neutronics parameters corresponding to the first core model and the calculation results of the second neutronics parameters corresponding to the second core model.

[0071] The calculation results of the first neutronics parameter can represent the core state when a more reactive material, i.e., the original fuel rod, is introduced at the insertion site, thus characterizing the upper limit of reactivity after inserting the old fuel rod. The calculation results of the second neutronics parameter can represent the core state when a less reactive material, i.e., an absorber rod, is introduced at the insertion site, thus characterizing the lower limit of reactivity after inserting the old fuel rod.

[0072] The reactivity value of the old fuel rods to be inserted is between that of the original fuel rods and the absorber rods. In some embodiments, the neutronics parameter evaluation results after the old fuel rods are inserted into the reactor core can be generated based on the calculation results of the first neutronics parameter and the calculation results of the second neutronics parameter.

[0073] For example, the evaluation result of neutron parameters after the core is fitted with old fuel rods can be a conclusion between the calculation result of the first neutron parameter and the calculation result of the second neutron parameter. That is, the calculation result of the first neutron parameter and the calculation result of the second neutron parameter constitute the possible range of values ​​of the neutron parameters after the introduction of old fuel rods.

[0074] The evaluation result of neutron parameters after the reactor core is fitted with old fuel rods can also be the more conservative one between the first and second neutron parameter calculation results, that is, the result that is biased towards the safe side in the neutron parameter calculation results.

[0075] In this embodiment, without directly performing complex modeling of the old fuel rods, the impact of the introduction of old fuel rods on the core neutronics parameters is rapidly and accurately evaluated by using models of both the original fuel rods and absorber rods, through separate calculations and result synthesis. This greatly simplifies the modeling process, improves computational efficiency, and ensures the reliability of the evaluation results, providing strong support for the fuel management design of nuclear reactors.

[0076] In one embodiment of this application, step 301 may include the following: Burnup calculations are performed on the first core model and the second core model respectively to obtain the first burnup calculation result corresponding to the first core model and the second burnup calculation result corresponding to the second core model. The calculation results of the first neutron parameters are obtained based on the first burnup calculation results and the neutron parameter calculation conditions in the final safety analysis report of the reactor core. The second neutron parameter calculation result is obtained based on the second burnup calculation result and the neutron parameter calculation conditions in the final safety analysis report of the reactor core.

[0077] In this embodiment, burnup calculation can simulate the change of fuel nuclide composition with time or burnup depth during the operation of a nuclear reactor. Burnup calculations can generate state libraries, i.e., burnup calculation results, under different burnup conditions, which can then be used for subsequent neutronics parameter calculations.

[0078] The final safety analysis report can refer to the nuclear review report submitted by a nuclear power plant to the relevant regulatory agency when applying for an operating license. The final safety analysis report can predefine several key operating states that need to be verified, namely the neutronics parameter calculation conditions. These neutronics parameter calculation conditions can cover representative or safety-challenging states throughout the reactor's entire lifespan, such as full-power hot state at the beginning of the lifespan or zero-power hot state at the end of the lifespan.

[0079] In some embodiments, data at specific time points matching the operating conditions specified in the final safety analysis report are precisely retrieved and extracted from the first burnup calculation results over the entire lifespan to obtain the first neutron parameter calculation results. For example, if the final safety analysis report specifies the need to verify the initial lifespan, hot full power, and control rod full-lift operating conditions, then the core effective multiplication factor and power distribution data at the corresponding times are extracted from the first burnup calculation results and used as the first neutron parameter calculation results. Similarly, based on the second burnup calculation results and the neutron parameter calculation operating conditions defined in the final safety analysis report, data at the corresponding operating condition nodes are extracted to obtain the second neutron parameter calculation results.

[0080] In this embodiment, by combining the neutronics parameter calculation conditions specified in the final safety analysis report, neutronics parameter calculation results consistent with nuclear safety review requirements can be obtained, ensuring that the final evaluation results are based on legal safety benchmarks and have high engineering reliability and compliance.

[0081] In one embodiment of this application, the neutronics parameter calculation result includes parameter values ​​corresponding to at least one category of neutronics parameters, and step 302 may include the following: Based on the calculation results of the first neutron parameter and the second neutron parameter, the first envelope value corresponding to each category of neutron parameter is determined respectively; Based on the first envelope value corresponding to the neutronics parameters described in each category, the neutronics parameter evaluation results after the reactor core is fitted with old fuel rods are generated.

[0082] In this embodiment, the envelope value can refer to a specific value or an envelope interval.

[0083] In some embodiments, for each category of neutron parameters, the parameter value corresponding to the neutron parameter of that category in the calculation result of the first neutron parameter can be compared with the parameter value corresponding to the neutron parameter of that category in the calculation result of the second neutron parameter to determine the first envelope value corresponding to the neutron parameter of that category.

[0084] In some embodiments, for each category of neutron parameters, the smaller parameter value between the first neutron parameter calculation result and the second neutron parameter calculation result can be taken as the first envelope value corresponding to that category of neutron parameter.

[0085] In some embodiments, for each category of neutron parameters, the larger parameter value of that category of neutron parameters between the calculation results of the first neutron parameter and the calculation results of the second neutron parameter can be taken as the first envelope value corresponding to that category of neutron parameters.

[0086] In some embodiments, for each category of neutronics parameter, the first envelope value corresponding to that category of neutronics parameter can be determined based on the impact of the value of that category of neutronics parameter on core safety.

[0087] For example, if the larger the parameter value of the neutronics parameter of this category is, the more detrimental it is to the safety of the reactor core, that is, the greater its impact on the safety of the reactor core, then the larger value between the first neutronics parameter calculation result and the second neutronics parameter calculation result is taken as the first envelope value corresponding to the neutronics parameter of this category.

[0088] In other embodiments, for each category of neutron parameters, the parameter value corresponding to that category of neutron parameter in the calculation results of the first and second neutron parameters represents the upper and lower limits of that category of neutron parameter, and the extreme value range of the two can be directly used as the envelope interval.

[0089] Furthermore, the first envelope value corresponding to each category of neutronics parameter can be used as the evaluation result of the neutronics parameters after the old fuel rods are inserted into the reactor core.

[0090] In this embodiment, through the envelope strategy, the parameter values ​​of each neutronics parameter under real conditions are better than or equal to the envelope value, thereby ensuring the conservatism and reliability of the neutronics parameter evaluation results after the core is fitted with old fuel rods.

[0091] In one embodiment of this application, the first envelope value corresponding to each category of neutron parameters is determined based on the calculation results of the first neutron parameter and the second neutron parameter, which may include the following: For neutron parameters of the same category, the parameter value of the neutron parameter corresponding to the category in the calculation result of the first neutron parameter and the parameter value of the neutron parameter corresponding to the category in the calculation result of the second neutron parameter are compared, and the parameter value with the greater impact on the core safety is determined as the first envelope value corresponding to the neutron parameter of that category.

[0092] In this embodiment, the parameter value that has a greater impact on core safety can refer to a value that is closer to the safety limit, or is less conducive to safe core operation, or requires more safety margin in nuclear safety analysis.

[0093] In this embodiment, for neutron parameters of the same category, the parameter value of the neutron parameter corresponding to the category in the calculation result of the first neutron parameter and the parameter value of the neutron parameter corresponding to the category in the calculation result of the second neutron parameter are determined as the envelope value of the neutron parameter. This ensures that the finally determined envelope value represents a more unfavorable physical state, thereby safely encompassing the actual core state after the old fuel rods are inserted.

[0094] For example, neutronics parameters such as the moderator density coefficient are used to characterize changes in reactivity caused by variations in the core moderator density. For a reactor, to ensure the inherent safety of negative feedback in the core, the moderator density coefficient must be negative. The more positive the moderator density coefficient (closer to zero or even becoming positive), the weaker the core's negative feedback capability, the greater the risk of runaway reactivity in the event of coolant loss or temperature rise transients, and the greater the impact on core safety.

[0095] For the same category of moderator density coefficient parameters, the larger parameter value between the calculated result of the first neutronics parameter and the calculated result of the second neutronics parameter is determined as the first envelope value corresponding to the neutronics parameter of that category.

[0096] For example, neutronics parameters such as the maximum boron differential value, used to characterize the change in reactivity caused by a unit change in core boron concentration (ppm), are typically negative (adding boron reduces reactivity). In the event of a design-baseline accident such as accidental boron dilution, the larger the absolute value of the boron differential value (i.e., the more negative the value), the more positive reactivity is introduced for the same amount of misoperational dilution, the more dramatic the power increase, the more severe the accident consequences, and the greater the impact on core safety.

[0097] For the maximum boron differential value parameter of the same category, the smaller parameter value between the calculated result of the first neutron parameter and the calculated result of the second neutron parameter is determined as the first envelope value corresponding to the neutron parameter of that category.

[0098] In this embodiment, since different types of neutronics parameters have different constraints on core safety (e.g., higher peak power is more dangerous, while lower reactivity margin is more dangerous), this embodiment does not adopt a simple "maximum or minimum" logic. Instead, it uses the parameter with the greater impact on core safety as a unified benchmark to accurately extract the extreme values ​​of parameters under the operating conditions that have the greater impact on core safety. This avoids the waste of design margins caused by excessive conservatism while ensuring that the absolute bottom line of safety evaluation is not breached, thus improving the engineering practicality and scientific rigor of the neutronics parameter evaluation results.

[0099] In one embodiment of this application, the evaluation results of neutron parameters after the core is fitted with old fuel rods are generated based on the first envelope value corresponding to each category of neutron parameters. This may include the following: Obtain the limit values ​​corresponding to each category of neutronics parameters used in the accident analysis in the final safety analysis report of the reactor core; The evaluation results of the neutronics parameters are generated based on the first envelope value and limit value corresponding to each category of neutronics parameters.

[0100] In some embodiments, generating neutron parameter evaluation results based on the first envelope value and limit value corresponding to each category of neutron parameters may include the following: For the same category of neutron parameters, the second envelope value corresponding to the neutron parameter of that category is selected from the first envelope value and the limit value corresponding to the neutron parameter of that category, and has a greater impact on the core safety. The neutronics parameter evaluation results are generated based on the second envelope value corresponding to each category of neutronics parameter.

[0101] In this embodiment, after determining the first envelope value corresponding to each category of neutronics parameters, it can be further determined whether they meet nuclear safety requirements. In the field of nuclear engineering, core design needs to meet various safety limits in the final safety analysis report.

[0102] In some embodiments, the final safety analysis report conducts an extremely conservative accident analysis for various design baseline accidents such as main steam pipe rupture and accidental boron dilution, and based on this, provides thresholds, i.e. limits, for various neutronics parameters to ensure core safety.

[0103] For example, the final security analysis report may specify: The limit for the density coefficient of the moderator is: it must be less than or equal to 0 pcm / (kg / m³) under all operating conditions (i.e., negative feedback must be guaranteed); The maximum boron differential value is limited to an absolute value of no more than 15 pcm / ppm (i.e., limiting the rate of positive reactivity introduction under dilution accidents).

[0104] Furthermore, for neutron parameters of the same category, the second envelope value corresponding to the neutron parameter of the same category is selected from the first envelope value and the limit value corresponding to the neutron parameter of the same category, and is determined to have a greater impact on core safety.

[0105] In some embodiments, for neutronics parameters where larger values ​​indicate greater danger, if the first envelope value is greater than the limit, it indicates that the calculated state has exceeded the safety boundary. The first envelope value represents the actual dangerous state, and is taken as the second envelope value to reflect the true risk. If the first envelope value is less than the limit, it indicates that the state is safe. In this case, the limit represents the potential risk boundary, and is taken as the second envelope value to maintain a conservative margin. That is, the second envelope value is the larger of the two.

[0106] For neutronics parameters where smaller values ​​indicate greater risk, if the envelope value is less than the limit, it indicates that the safety requirements are not met, and the envelope value is taken as the target value; if the envelope value is greater than the limit, it indicates that the requirements are met, and the limit value is taken as the target value to ensure conservative constraints. That is, the target value is the smaller of the two.

[0107] For example, for neutronics parameters such as the moderator density coefficient, where a larger value indicates greater danger, suppose the corresponding first envelope value is -2 pcm / (kg / m³), while the limit value in the final safety analysis report is 0 pcm / (kg / m³). Since 0 is more positive than -2, has a weaker negative feedback capability, and has a greater impact on safety, 0 is selected as the second envelope value for this category.

[0108] For neutronics parameters like the maximum boron differential value, where smaller values ​​are more dangerous, we assume the corresponding first envelope value is -12 pcm / ppm, while the final safety analysis report limits it to -15 pcm / ppm. Since -15 is smaller, it introduces a more vigorous positive reactivity in boron dilution accidents, having a greater impact on safety; therefore, -15 is chosen as the second envelope value for this category.

[0109] It should be noted that, in some embodiments, considering the uncertainties in nuclear design calculations, such as calculation model errors, nuclear data errors, fuel manufacturing tolerances, etc., for each neutronics parameter, a certain design margin can be taken based on the first envelope value to obtain a corrected first envelope value. Then, the corrected first envelope value is compared with the limit value to determine the second envelope value, so as to ensure that the neutronics parameter evaluation results have sufficient safety margin.

[0110] It should be noted that whether smaller or larger values ​​of neutron parameters in each category are more dangerous can be determined based on actual circumstances or historical experience, and no specific limit is made here.

[0111] Furthermore, neutron parameter evaluation results are generated based on the second envelope values ​​corresponding to each category of neutron parameters.

[0112] In some embodiments, the second envelope value corresponding to each category of neutron parameters can be used as the neutron parameter evaluation result. The neutron parameter evaluation result can be a report containing the second envelope value corresponding to each category of neutron parameters, or it can be a safety margin chart drawn based on the second envelope value corresponding to each category of neutron parameters.

[0113] In this embodiment, the target value represents a more unfavorable situation, that is, either the actual extreme calculated value or the legal safety boundary, so that the obtained neutronics parameter evaluation results can provide conservative and reliable safety guidance for core operators and ensure the absolute safety of core operation after the old fuel rods are inserted.

[0114] It should be noted that the various embodiments described in this application can be combined with each other or implemented individually without conflict, and this application does not limit this.

[0115] It should be noted that, unless otherwise specified, the embodiments and features described in this application can be combined with each other. This application will now be described in detail with reference to the accompanying drawings and embodiments.

[0116] For ease of understanding, a specific embodiment will be used as an example. Please refer to [link / reference]. Figure 3 , Figure 3 This is a schematic diagram illustrating the implementation process of evaluating neutronics parameters in a nuclear reactor core, as provided in an embodiment of this application. Specifically: S1: Initial conditions.

[0117] Information on the fuel rods to be irradiated has been determined; The information on the components to be fitted and the corresponding fitting positions of the fuel rods to be irradiated have been determined. The irradiation cycle core loading scheme (i.e., the fuel assembly loading scheme of the core) and related information on the cycle length have been determined.

[0118] S2: Component modeling calculation.

[0119] 1) Based on the information of the fuel assembly to be interpolated (i.e., the information of the target fuel assembly), establish a first fuel assembly model, perform ground state fuel consumption calculation, disturbance calculation, and branch calculation, and form a new fuel assembly library; 2) Based on the information of the fuel assembly to be inserted and the insertion position information of the fuel rod to be irradiated (i.e., the information of the target fuel assembly after replacement), a second fuel assembly model with stainless steel rods inserted is established, and ground state burnup calculation, disturbance calculation, and branch calculation are carried out to form a new fuel assembly library with stainless steel rods inserted.

[0120] S3: Core neutronics parameter calculation.

[0121] Based on the information of the new fuel assembly library and the core loading scheme (fuel assembly loading scheme of the core), as well as the relevant information of the cycle length, core model 1 (i.e. the first core model) is established, and burnup calculation is carried out. Based on the neutronics parameter calculation conditions in the final safety analysis report, the calculation of various neutronics parameters (such as moderator density coefficient, maximum boron differential value, etc.) is carried out. Replace the component library information of the component to be irradiated with the component library information of the inserted stainless steel rod (i.e. absorber rod) and the relevant information of the cycle length, establish core model 2 (i.e. the second core model), carry out burnup calculation, and calculate various neutron parameters (such as moderator density coefficient, maximum boron differential value, etc.) based on the neutron parameters calculation conditions in the final safety analysis report.

[0122] S4: Neutron spectral parameter envelope value determined.

[0123] 1) Determine the envelope value of each neutronics parameter based on the calculation results of neutronics parameters of core model 1 and core model 2; 2) Compare the neutronics parameter envelope value with the neutronics parameter limit value used in the accident analysis in the final safety analysis report, and pass the envelope value with a certain design margin or the neutronics parameter limit value used in the accident analysis in the final safety analysis report (if the limit value is more conservative than the calculated values ​​of core model 1 and core model 2) to downstream professionals to carry out safety analysis and evaluation.

[0124] In this embodiment, a squeeze-and-response evaluation method is adopted. By calculating the neutronics parameters of two computational models, namely the model without replacing the fuel rods in fuel assembly B (core model 1) and the model with replacing the fuel rods in fuel assembly B with absorber rods (such as stainless steel rods) (core model 2), the influence of fuel assembly rod replacement on core neutronics parameters is analyzed envelopingly. (The fuel rods in fuel assembly B have lower burnup and higher reactivity than the target fuel rods, while absorber rods such as stainless steel rods have lower reactivity than the target fuel rods. Therefore, the neutronics parameters of the core with rod replacement are between those of core model 1 and core model 2.) On the one hand, this avoids the complex modeling process of replacing fuel assemblies and the secondary development of nuclear design programs. On the other hand, the analysis results are more comprehensive and more beneficial to core safety.

[0125] Based on the neutronics parameter evaluation method for nuclear reactor cores provided in the above embodiments, this application also provides specific implementation methods for a neutronics parameter evaluation device for nuclear reactor cores. Please refer to the following embodiments.

[0126] Based on the neutronics parameter evaluation method for nuclear reactor cores provided in the above embodiments, this application also provides specific implementation methods for a neutronics parameter evaluation device for nuclear reactor cores. Please refer to the following embodiments.

[0127] See Figure 4 The neutronics parameter evaluation device 400 for a nuclear reactor core provided in this application embodiment may include: Information determination module 401 is used to determine the fuel rod insertion position of the target fuel assembly to be inserted with the old fuel rod in the reactor core; The first construction module 402 is used to construct a first core model of the reactor core based on the information of the target fuel assembly; The second construction module 403 is used to replace the original fuel rod in the target fuel assembly located at the fuel rod insertion position with the absorber rod in response to inserting the absorber rod at the fuel rod insertion position, obtain the target fuel assembly information after replacement, and construct the second core model of the reactor core based on the target fuel assembly information after replacement. The evaluation and determination module 404 is used to determine the neutronics parameter evaluation results after the old fuel rods are inserted into the core, based on the first core model and the second core model.

[0128] The neutronics parameter evaluation device for the nuclear reactor core provided in this application embodiment can achieve... Figure 4 The various processes implemented by the neutronics parameter evaluation device in the nuclear reactor core in the method embodiment will not be described again here to avoid repetition.

[0129] Figure 5 A schematic diagram of the hardware structure of the neutronics parameter evaluation device for a nuclear reactor core provided in an embodiment of this application is shown.

[0130] The neutronics parameter evaluation device for a nuclear reactor core may include a processor 501 and a memory 502 storing computer program instructions.

[0131] Specifically, the processor 501 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.

[0132] Memory 502 may include mass storage for data or instructions. For example, and not limitingly, memory 502 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 502 may include removable or non-removable (or fixed) media. Where appropriate, memory 502 may be internal or external to the integrated gateway disaster recovery device. In a particular embodiment, memory 502 is non-volatile solid-state memory.

[0133] Memory may include read-only memory (ROM), random access memory (RAM), disk storage media devices, optical storage media devices, flash memory devices, and electrical, optical, or other physical / tangible memory storage devices. Therefore, typically, memory includes one or more tangible (non-transitory) computer-readable storage media (e.g., memory devices) encoded with software including computer-executable instructions, and when the software is executed (e.g., by one or more processors), it is operable to perform the operations described with reference to the methods according to one aspect of this disclosure.

[0134] The processor 501 reads and executes computer program instructions stored in the memory 502 to implement any of the neutronics parameter evaluation methods for nuclear reactor cores in the above embodiments.

[0135] In one example, the neutronics parameter evaluation device for a nuclear reactor core may also include a communication interface 503 and a bus 510. For example, Figure 5 As shown, the processor 501, memory 502, and communication interface 503 are connected through bus 510 and complete communication with each other.

[0136] The communication interface 503 is mainly used to realize communication between various modules, devices, units and / or equipment in the embodiments of this application.

[0137] Bus 510 includes hardware, software, or both, that couples together components of a neutronics parameter evaluation device for a nuclear reactor core. This is an example, not a limitation. The bus may include an Accelerated Graphics Port (AGP) or other graphics bus, an Extended Industry Standard Architecture (EISA) bus, a Front Side Bus (FSB), a Hyper Transport (HT) interconnect, an Industry Standard Architecture (ISA) bus, an Infinite Bandwidth Interconnect, a Low Pin Count (LPC) bus, a memory bus, a Micro Channel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a Serial Advanced Technology Attachment (SATA) bus, a Video Electronics Standards Association Local Bus (VLB) bus, or other suitable buses, or a combination of two or more of these. Where appropriate, bus 510 may include one or more buses. Although specific buses are described and illustrated in embodiments of this application, this application contemplates any suitable bus or interconnect.

[0138] Furthermore, in conjunction with the neutronics parameter evaluation method for nuclear reactor cores 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 neutronics parameter evaluation methods for nuclear reactor cores in the above embodiments.

[0139] This application embodiment may also provide a computer program product, wherein when the instructions in the computer program product are executed by the processor of the neutronics parameter evaluation device of the nuclear reactor core, the neutronics parameter evaluation device of the nuclear reactor core performs any of the neutronics parameter evaluation methods of the nuclear reactor core in the above embodiments.

[0140] 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.

[0141] 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. The 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, floppy disks, CD-ROMs, optical disks, hard disks, fiber optic media, radio frequency links, etc. Code segments can be downloaded via computer networks such as the Internet, intranets, etc.

[0142] 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.

[0143] 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.

[0144] 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 for evaluating neutronics parameters of a nuclear reactor core, characterized in that, include: Determine the fuel rod insertion position of the target fuel assembly in the reactor core to be fitted with the old fuel rods; Based on the information of the target fuel assembly, a first core model of the reactor core is constructed; In response to inserting an absorber rod at the fuel rod insertion position, the original fuel rod in the target fuel assembly located at the fuel rod insertion position is replaced with the absorber rod to obtain the target fuel assembly information after replacement, and a second core model of the reactor core is constructed based on the target fuel assembly information after replacement. Based on the first core model and the second core model, the evaluation results of the neutronics parameters after the core is fitted with old fuel rods are determined.

2. The method according to claim 1, characterized in that, The target core model is determined through the following steps: Based on the target information, construct a target fuel component model; Based on the target fuel assembly model, construct the target reactor core model; Wherein, the target core model is either the first core model or the second core model; In response to the target core model being the first core model, the target information is the information of the target fuel assembly; In response to the target core model being the second core model, the target information is the replaced target fuel assembly information.

3. The method according to claim 2, characterized in that, The target fuel assembly model is either a first fuel assembly model or a second fuel assembly model. The fuel rods in the first fuel assembly model are the original fuel rods of the target fuel assembly. In the second fuel assembly model, the fuel rods at the insertion positions are absorber rods. The construction of the target reactor core model based on the target fuel assembly model includes: Based on the target fuel assembly model, determine the target fuel assembly library; Based on the target fuel assembly library, construct the target reactor core model; Wherein, in response to the target core model being the first core model, the target fuel assembly model being the first fuel assembly model; In response to the target core model being the second core model, the target fuel assembly model being the second fuel assembly model.

4. The method according to claim 3, characterized in that, The step of determining the target fuel assembly library based on the target fuel assembly model includes: The ground-state fuel consumption of the target fuel assembly model is calculated to obtain ground-state fuel consumption data; Based on the ground-state fuel consumption data, perturbation calculations are performed on the target fuel assembly model to obtain perturbation parameters; Based on the ground-state fuel consumption data, branch calculations are performed on the target fuel assembly model to obtain branch parameters; The target fuel assembly library is determined based on the ground-state fuel consumption data, the disturbance parameters, and the branch parameters.

5. The method according to claim 3, characterized in that, The step of constructing the target reactor core model based on the target fuel assembly library includes: Obtain the fuel assembly loading scheme for the reactor core; Based on the fuel assembly loading scheme and the target fuel assembly library, the target reactor core model is constructed.

6. The method according to claim 1, characterized in that, The determination of neutronics parameter evaluation results after inserting old fuel rods into the reactor core based on the first core model and the second core model includes: Neutronics parameters are calculated for the first core model and the second core model respectively to obtain the calculation results of the first neutronics parameter corresponding to the first core model and the calculation results of the second neutronics parameter corresponding to the second core model. Based on the calculation results of the first neutron parameters and the calculation results of the second neutron parameters, the neutron parameter evaluation results after the reactor core is fitted with old fuel rods are generated.

7. The method according to claim 6, characterized in that, The step of calculating neutron parameters for the first core model and the second core model respectively, to obtain the first neutron parameter calculation result corresponding to the first core model and the second neutron parameter calculation result corresponding to the second core model, includes: Burnup calculations are performed on the first core model and the second core model respectively to obtain the first burnup calculation result corresponding to the first core model and the second burnup calculation result corresponding to the second core model. Based on the first burnup calculation result and the neutronics parameter calculation conditions in the final safety analysis report of the reactor core, the first neutronics parameter calculation result is obtained; The second neutron parameter calculation result is obtained based on the second burnup calculation result and the neutron parameter calculation conditions in the final safety analysis report of the reactor core.

8. The method according to claim 6, characterized in that, The neutronics parameter calculation results include parameter values ​​corresponding to at least one category of neutronics parameters; The step of generating neutron parameter evaluation results after inserting old fuel rods into the reactor core based on the calculation results of the first and second neutron parameters includes: Based on the calculation results of the first neutron parameter and the second neutron parameter, the first envelope value corresponding to each category of neutron parameter is determined respectively; Based on the first envelope value corresponding to the neutronics parameters described in each category, the neutronics parameter evaluation results after the reactor core is fitted with old fuel rods are generated.

9. The method according to claim 8, characterized in that, The step of determining the first envelope value corresponding to each category of neutron parameters based on the calculation results of the first neutron parameter and the second neutron parameter includes: For neutron parameters of the same category, the parameter value of the neutron parameter corresponding to the category in the calculation result of the first neutron parameter and the parameter value of the neutron parameter corresponding to the category in the calculation result of the second neutron parameter are compared, and the parameter value with the greater impact on the core safety is determined as the first envelope value corresponding to the neutron parameter of that category.

10. The method according to claim 8, characterized in that, The step of generating neutron parameter evaluation results after inserting old fuel rods into the reactor core based on the first envelope value corresponding to each category of neutron parameters includes: Obtain the limit values ​​corresponding to each category of neutronics parameters used in the accident analysis in the final safety analysis report of the reactor core; The evaluation results of the neutronics parameters are generated based on the first envelope value and limit value corresponding to each category of neutronics parameters.

11. The method according to claim 10, characterized in that, The step of generating neutronics parameter evaluation results based on the first envelope value and limit value corresponding to each category of neutronics parameter includes: For the same category of neutron parameters, the second envelope value corresponding to the neutron parameter of that category is selected from the first envelope value and the limit value corresponding to the neutron parameter of that category, and has a greater impact on the core safety. The neutronics parameter evaluation results are generated based on the second envelope value corresponding to each category of neutronics parameter.

12. A device for evaluating neutronics parameters of a nuclear reactor core, characterized in that, include: The information determination module is used to determine the fuel rod insertion position of the target fuel assembly in the reactor core to be inserted with the old fuel rods; The first construction module is used to construct a first core model of the reactor core based on the information of the target fuel assembly; The second construction module is used to replace the original fuel rod located at the fuel rod insertion position in the target fuel assembly with the absorber rod in response to inserting the absorber rod at the fuel rod insertion position, obtain the target fuel assembly information after replacement, and construct the second core model of the reactor core based on the target fuel assembly information after replacement. The evaluation and determination module is used to determine the neutronics parameter evaluation results of the reactor core after inserting old fuel rods, based on the first core model and the second core model.

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