Reactor safety assessment method, device and system

Abstract

The invention relates to a reactor safety assessment method, device and system. The method comprises the following steps: controlling a pre-constructed first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process; performing mesh generation on the first model corresponding to the sub-channel module to obtain a first mesh, and performing mesh generation on the second model corresponding to the neutron dynamics module to obtain a second mesh; mapping the first grid and the second grid to obtain a mapping relation and a corresponding equivalent weight factor; determining the nuclear power of the fuel rod by using the sub-channel module and the neutron dynamics module based on the equivalent weight factor, and obtaining the reactor core power distribution information of the time step length; and based on the reactor core power distribution information, respectively carrying out enthalpy analysis and deviation nucleate boiling ratio DNBR analysis by utilizing a neutron dynamics module, and determining that the reactor is unsafe under the condition that an analysis result does not meet a preset condition. According to the scheme, the accuracy of a reactor safety assessment result is improved.

Description

Technical Field

[0001] The present disclosure relates to the technical field of reactor safety analysis, and in particular to a reactor safety assessment method, device and system. Background Art

[0002] In the related art, if a control rod mechanism pressure shell in the reactor is mechanically damaged, the pressure of the reactor coolant system will cause a bundle of control rod assemblies together with their drive rods to be ejected from the reactor. This mechanical damage will cause the rapid introduction of positive reactivity in the core, causing a sharp increase in the core nuclear power, and produce an unfavorable core power distribution, which may cause local fuel rod damage. Therefore, it is necessary to ensure that after the above accident occurs, the damage caused by the accident needs to be controlled to a certain extent to avoid serious impact on the equipment and the surrounding environment. To this end, it is necessary to evaluate the safety of the reactor by simulating the occurrence process of the above accident. The existing simulation method usually uses the three-dimensional neutron dynamics software and the sub-channel software for external coupling, that is, to display and interact the core data. This method will cause the three-dimensional neutron dynamics software and the sub-channel software to mismatch, resulting in unstable calculation or low evaluation efficiency. Summary of the invention

[0003] In order to overcome the problems existing in the related art, the present disclosure provides a reactor safety assessment method, device and system.

[0004] According to a first aspect of an embodiment of the present disclosure, a reactor safety assessment method is provided, comprising controlling a pre-built first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process; the first process is a process in which a fuel rod assembly is ejected from the reactor; the first model is a three-dimensional model of a reactor core, and the first model is arranged in a sub-channel module;

[0005] Meshing the first model and corresponding to the sub-channel module to obtain a first mesh, and meshing the second model corresponding to the neutron dynamics module to obtain a second mesh;

[0006] Mapping the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; the equivalent weight factor is used to control the reactivity equivalence of the first model and the second model;

[0007] For each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron kinetics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step;

[0008] Based on the core power distribution information, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain analysis results;

[0009] When the analysis result does not satisfy a preset condition, it is determined that the reactor is unsafe.

[0010] In some embodiments of the present disclosure, the fuel rod parameters of the three-dimensional core model in the first model and the second model are consistent;

[0011] The mapping process of the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship, includes:

[0012] According to the position information of the first grid and the position information of the second grid, each first grid is matched with at least one second grid to obtain a plurality of the mapping relationships; the number of fuel rods in the first grid is the same as the total number of fuel rods in the matched at least one second grid;

[0013] For each first grid and at least one second grid corresponding to each mapping relationship, the equivalent weight factor is calculated according to temperature information corresponding to the first grid and temperature information corresponding to the at least one second grid.

[0014] In some embodiments of the present disclosure, calculating the equivalent weight factor according to the temperature information corresponding to the first grid and the temperature information corresponding to the at least one second grid includes:

[0015] Calculating a first temperature average value based on the temperature information corresponding to the first grid, and calculating a second temperature average value based on the temperature information corresponding to the at least one second grid;

[0016] Normalizing the first temperature average value to obtain a first equivalent weight factor, and normalizing the second temperature average value to obtain a second equivalent weight factor;

[0017] After calculating the equivalent weight factor according to the temperature information corresponding to the first grid and the temperature information corresponding to the at least one second grid, the method further includes:

[0018] determining whether the reactivity of the first grid and the at least one second grid is equivalent based on the first equivalent weight factor and the second equivalent weight factor;

[0019] determining a reactivity deviation value in the event that the reactivity of the first grid and the at least one second grid are not equivalent;

[0020] determining a first rate of change of the resonance cross section with temperature based on the first temperature average value, and determining a second rate of change of the resonance cross section with temperature based on the second temperature average value;

[0021] The first equivalent weight factor and the second equivalent weight factor are adjusted based on the reactivity deviation value, the first change rate and the second change rate, and the step of determining whether the reactivity of the first grid and the at least one second grid are equivalent is returned to be executed until the reactivity of the first grid and the at least one second grid are equivalent.

[0022] In some embodiments of the present disclosure, before performing enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis based on the core power distribution information and obtaining analysis results using the neutron kinetics module, the method further includes:

[0023] Preprocess the core power distribution information corresponding to each time step;

[0024] The preprocessed power distribution information of each core is stored in the array variable in chronological order;

[0025] Output all data in the array variable to a hot rod file;

[0026] Based on the core power distribution information, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results, including:

[0027] Based on the hot rod file, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain the analysis results.

[0028] In some embodiments of the present disclosure, when the analysis result does not meet a preset condition, determining that the reactor is unsafe includes:

[0029] Determining a target fuel rod with the highest heat based on the core power distribution information;

[0030] Performing enthalpy analysis on the target fuel rod to obtain a first analysis result;

[0031] Performing DNBR analysis on the target fuel rod to obtain a second analysis result;

[0032] When the second analysis result is greater than a preset threshold, performing DNBR analysis on each assembly to obtain a third analysis result, and determining the number of fuel rods that do not meet the preset condition based on the third analysis result; wherein the assembly includes at least one fuel rod;

[0033] When at least one of the first analysis result and the second analysis result does not satisfy a corresponding preset condition, it is determined that the reactor is unsafe.

[0034] In some embodiments of the present disclosure, before the first process of controlling the pre-built three-dimensional model of the reactor core to simulate the rod ejection accident, the method further includes:

[0035] Acquire model configuration information input by a user; the model configuration information includes parameter information and arrangement information of a target sub-model in the first model;

[0036] When the target sub-model exists in the preset model library, the target sub-model is arranged according to the arrangement information to obtain the first model;

[0037] When the target sub-model does not exist in the preset model library, the target sub-model is generated based on the parameter information, the target sub-model is arranged according to the arrangement information to obtain the first model, and the target sub-model is stored in the model library.

[0038] According to a second aspect of an embodiment of the present disclosure, there is provided a reactor safety assessment device, comprising:

[0039] a simulation unit, used for controlling a pre-built first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process; the first process is a process in which a fuel rod assembly is ejected from the reactor; the first model is a three-dimensional model of a reactor core, and the first model is arranged in a sub-channel module;

[0040] A partitioning unit, configured to perform mesh partitioning on the first model and corresponding to the sub-channel module to obtain a first mesh, and to perform mesh partitioning on the second model corresponding to the neutron dynamics module to obtain a second mesh;

[0041] A mapping unit, configured to map the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; the equivalent weight factor is used to control the reactivity equivalence of the first model and the second model;

[0042] A first determining unit is used for determining, for each time step in the first process, based on the equivalent weight factor, using the sub-channel module and the neutron kinetics module to determine the nuclear power of each fuel rod in the time step, so as to obtain the core power distribution information of the time step;

[0043] An analysis unit, configured to perform enthalpy analysis and deviation from nucleate boiling ratio (DNBR) analysis respectively based on the core power distribution information and using the neutron kinetics module to obtain analysis results;

[0044] The second determination unit is used to determine that the reactor is unsafe if the analysis result does not meet a preset condition.

[0045] According to a third aspect of an embodiment of the present disclosure, an electronic device includes: a memory, a processor, and a computer program stored in the memory and executable on the processor, wherein when the processor executes the computer program, the method described in any one of the first aspects is implemented.

[0046] According to a fourth aspect of an embodiment of the present disclosure, there is provided a computer-readable storage medium on which a computer program is stored. When the computer program is executed by a processor, the method according to any one of the first aspects is implemented.

[0047] According to a fifth aspect of an embodiment of the present disclosure, a computer program product is provided, comprising a computer program, wherein the computer program implements the method as described in any one of the first aspects when executed by a processor.

[0048] The technical solution provided by the embodiments of the present disclosure may include the following beneficial effects: by controlling a pre-constructed first model to simulate a first process of a rod ejection accident, the operating data of the first model in the first process is obtained; the first model and corresponding to the sub-channel module are grid-divided to obtain a first grid, and the second model corresponding to the neutron dynamics module is grid-divided to obtain a second grid; the first grid and the second grid are mapped to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; for each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron dynamics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step; based on the core power distribution information, the neutron dynamics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results; when the analysis results do not meet the preset conditions, the reactor is determined to be unsafe. By internally coupling the subchannel module and the neutron dynamics module, that is, mapping the grids of the subchannel module and the neutron dynamics module, and configuring equivalent weight factors, the consistency of data interaction between the subchannel module and the neutron dynamics module when collaboratively calculating the core power distribution is ensured, avoiding the occurrence of mismatches, and improving the stability of calculations and evaluation efficiency. In addition, by simulating the rod ejection accident and using the operating data in the simulation process to evaluate the safety of the reactor, the accuracy of the reactor safety assessment results is improved, and the damage caused by the carbon rod accident to the equipment and the surrounding environment is reduced.

[0049] It is to be understood that the foregoing general description and the following detailed description are exemplary and explanatory only and are not restrictive of the present disclosure. BRIEF DESCRIPTION OF THE DRAWINGS

[0050] The accompanying drawings, which are incorporated in and constitute a part of this specification, illustrate embodiments consistent with the invention and, together with the description, serve to explain the principles of the invention.

[0051] Figure 1 The present invention is a flowchart of a reactor safety assessment method according to an exemplary embodiment.

[0052] Figure 2 It is a block diagram of a reactor safety assessment device according to an exemplary embodiment.

[0053] Figure 3 It is a block diagram of a device for a reactor safety assessment method according to an exemplary embodiment. DETAILED DESCRIPTION

[0054] Exemplary embodiments will be described in detail herein, examples of which are shown in the accompanying drawings. When the following description refers to the drawings, the same numbers in different drawings represent the same or similar elements unless otherwise indicated. The embodiments described in the following exemplary embodiments do not represent all embodiments consistent with the present invention. Instead, they are merely examples of devices and methods consistent with some aspects of the present invention as detailed in the appended claims.

[0055] The terms used in the disclosed embodiments are only for the purpose of describing specific embodiments, and are not intended to limit the disclosed embodiments. The singular forms "a", "an" and "the" used in the disclosed embodiments and the appended claims are also intended to include plural forms, unless the context clearly indicates other meanings.

[0056] It should be understood that although the terms first, second, third, etc. may be used to describe various information in the disclosed embodiments, these information should not be limited to these terms. These terms are only used to distinguish the same type of information from each other. For example, without departing from the scope of the disclosed embodiments, the first information may also be referred to as the second information, and similarly, the second information may also be referred to as the first information. Depending on the context, the words "if" and "if" as used herein may be interpreted as "at" or "when" or "in response to determination".

[0057] In addition, various forms of processes shown in the embodiments of the present disclosure may be used to reorder, add or delete steps. For example, the steps described in the present application may be performed in parallel, sequentially or in different orders, as long as the desired results of the technical solution disclosed in the present disclosure can be achieved, and this document does not limit this.

[0058] In the related art, if a control rod mechanism pressure shell in the reactor is mechanically damaged, the pressure of the reactor coolant system will cause a bundle of control rod assemblies together with their drive rods to be ejected from the reactor. This mechanical damage will cause the rapid introduction of positive reactivity in the core, causing a sharp increase in the core nuclear power, and produce an unfavorable core power distribution, which may cause local fuel rod damage. Therefore, it is necessary to ensure that after the above accident occurs, the damage caused by the accident needs to be controlled to a certain extent to avoid serious impact on the equipment and the surrounding environment. To this end, it is necessary to evaluate the safety of the reactor by simulating the occurrence process of the above accident. The existing simulation method usually uses the three-dimensional neutron dynamics software and the sub-channel software for external coupling, that is, to display and interact the core data. This method will cause the three-dimensional neutron dynamics software and the sub-channel software to mismatch, resulting in unstable calculation or low evaluation efficiency.

[0059] In order to solve the above problems, the present disclosure provides a reactor safety assessment method, device and system, which control a pre-constructed first model to simulate a first process of a rod ejection accident to obtain the operation data of the first model in the first process; gridding the first model and corresponding to the sub-channel module to obtain a first grid, and gridding the second model corresponding to the neutron dynamics module to obtain a second grid; mapping the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; for each time step in the first process, based on the equivalent weight factor, using the sub-channel module and the neutron dynamics module to determine the nuclear power of each fuel rod in the time step, to obtain the core power distribution information of the time step; based on the core power distribution information, using the neutron dynamics module to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis, respectively, to obtain analysis results; if the analysis results do not meet the preset conditions, it is determined that the reactor is unsafe. By internally coupling the subchannel module and the neutron dynamics module, that is, mapping the grids of the subchannel module and the neutron dynamics module, and configuring equivalent weight factors, the consistency of data interaction between the subchannel module and the neutron dynamics module when collaboratively calculating the core power distribution is ensured, avoiding the occurrence of mismatches, and improving the stability of calculations and evaluation efficiency. In addition, by simulating the rod ejection accident and using the operating data in the simulation process to evaluate the safety of the reactor, the accuracy of the reactor safety assessment results is improved, and the damage caused by the carbon rod accident to the equipment and the surrounding environment is reduced.

[0060] Figure 1 is a flow chart of a reactor safety assessment method according to an exemplary embodiment. Figure 1 As shown, it should be noted that the reactor safety assessment method of the embodiment of the present disclosure is applied to a reactor safety assessment device. Figure 1As shown, the method may include the following steps:

[0061] Step 101 , controlling a pre-built first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process.

[0062] Among them, the first process is the process of ejecting the fuel rod assembly from the reactor; the first model is a three-dimensional model of the reactor core, and the first model is set in the sub-channel module.

[0063] As a possible example, controlling the first process of simulating a rod ejection accident using the pre-built first model may include controlling the ejection of fuel rods, controlling the shutdown time, controlling the insertion of fuel rods, etc. The operating data may include core power, enthalpy rise heat pipe factor FDH, hot rod axial power distribution, and core axial power distribution.

[0064] Step 102, meshing the first model and corresponding to the sub-channel module to obtain a first mesh, and meshing the second model corresponding to the neutron dynamics module to obtain a second mesh.

[0065] It is understandable that there are multiple first grids and multiple second grids. Usually, due to the different functional designs of the sub-channel module and the neutron dynamics module, the number of fuel rods included in the first grid is greater than the number of fuel rods included in the second grid.

[0066] Step 103 : mapping the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship.

[0067] The equivalent weight factor is used to control the reactivity equivalence between the first model and the second model.

[0068] In one embodiment, the first grid and the second grid may be mapped according to a preset mapping rule to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship.

[0069] It should be noted that reactivity is a physical quantity that reflects the state of a nuclear reactor. Since the sub-channel module corresponding to the first model and the neutron dynamics module corresponding to the second model belong to different functional modules and usually use different application software, even if the objects of analysis are the same, there will still be a problem of inequality in the reactivity of the first model and the second model. Therefore, it is necessary to adjust the parameters through equivalent weight factors to ensure the equivalence of the reactivity of the first model and the second model.

[0070] In some embodiments of the present disclosure, the fuel rod parameters of the three-dimensional core model in the first model and the second model are consistent, and step 103 may specifically include the following steps:

[0071] Step a1, matching each first grid with at least one second grid according to the position information of the first grid and the position information of the second grid to obtain a plurality of mapping relationships.

[0072] The number of fuel rods in the first grid is the same as the total number of fuel rods in at least one matching second grid.

[0073] In one embodiment, identifiers may be assigned to the first grid and the second grid in advance. Since the fuel rod parameters of the three-dimensional core model in the first model and the second model are consistent, that is, the quantity and arrangement are the same, the mapping relationship may be determined according to the position information of the first grid and the second grid, that is, the first grid and the second grid with consistent position information are taken as a group of mappings. Since the number of fuel rods in the first grid is usually greater than the number of fuel rods in the second grid, a group of mapping relationships includes an identifier of the first grid and multiple identifiers of the second grids.

[0074] Step a2: for each first grid and at least one second grid corresponding to each mapping relationship, an equivalent weight factor is calculated according to temperature information corresponding to the first grid and temperature information corresponding to the at least one second grid.

[0075] It is understandable that whether the reactivity of the first model and the second model is consistent can be reflected in the temperature information, so the equivalent weight factor can be calculated according to the temperature information corresponding to the first grid and the temperature information corresponding to at least one second grid.

[0076] In some embodiments of the present disclosure, step a2 may specifically include the following steps:

[0077] Step a21, calculating a first temperature average value based on temperature information corresponding to the first grid, and calculating a second temperature average value based on temperature information corresponding to at least one second grid.

[0078] In one embodiment, the current first temperature of each fuel rod in the first grid and the current second temperature of each fuel rod in at least one second grid are obtained, multiple first temperatures are averaged to obtain a first temperature average value, and multiple second temperatures are averaged to obtain a second temperature average value.

[0079] Step a22, normalizing the first temperature average value to obtain a first equivalent weight factor, and normalizing the second temperature average value to obtain a second equivalent weight factor.

[0080] After step 103, the method may further include:

[0081] Step a23, determining whether the reactivity of the first grid and at least one second grid is equivalent based on the first equivalent weight factor and the second equivalent weight factor.

[0082] In one embodiment, the temperature information corresponding to the first grid is weighted using a first equivalent weight factor, and the temperature information corresponding to the second grid is weighted using a second equivalent weight factor, and whether the reactivity of the first grid and at least one second grid is equivalent is determined based on the weighted first equivalent weight factor and the weighted second equivalent weight factor.

[0083] It should be noted that the conventional calculation method is used to determine whether the reactivity is equivalent based on the temperature information, which will not be elaborated here.

[0084] Step a24, determining a reactivity deviation value when the reactivity of the first grid and at least one second grid are not equivalent.

[0085] In one embodiment, a reflection deviation value between the first grid and at least one second grid is determined according to the reactivity calculation result.

[0086] Step a25, determining a first change rate of the resonance cross section with temperature based on the first temperature average value, and determining a second change rate of the resonance cross section with temperature based on the second temperature average value.

[0087] In one embodiment, the first temperature average and the second temperature average may be Doppler equivalent temperatures.

[0088] As an example, the resonance cross section refers to the microscopic absorption cross section of neutrons in a specific energy range. The corresponding first change rate can be found in the cross section database according to the first temperature average value, and the corresponding second change rate can be found in the cross section database according to the second temperature average value.

[0089] Step a26, adjusting the first equivalent weight factor and the second equivalent weight factor based on the reactivity deviation value, the first change rate and the second change rate, and returning to the step of determining whether the reactivity of the first grid and at least one second grid is equivalent until the reactivity of the first grid and at least one second grid is equivalent.

[0090] For example, the initial first temperature average is 100 degrees, and the second temperature average is 300 degrees. Therefore, after normalization, the initial first equivalent weight factor is 0.5, and the second equivalent weight factor is 1.5. The reactivity of the subchannel module and the neutron dynamics module calculated with these two weight factors deviates by 5%. The first change rate corresponding to 100 degrees is 16.0, and the second change rate corresponding to 300 degrees is 53.0. Then the adjustment amount of the first temperature average corresponding to 100 degrees is 5%*16.0=0.8, and the first temperature average becomes 100.8 degrees. The adjustment amount of the second temperature average corresponding to 300 degrees is 5%*53.0=2.65, and the second temperature average becomes 302.65 degrees. Now we get two new Doppler equivalent temperatures, 100.8 degrees and 302.65 degrees. After normalization, the new first equivalent weight factor is 0.4997, and the new second equivalent weight factor is 1.5003.

[0091] Step 104, for each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron dynamics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step.

[0092] In one embodiment, a sub-channel module is used to perform transient coupling calculations, such as solving a three-dimensional two-group neutron diffusion theory kinetic equation set with six groups of delayed neutrons, taking into account the Doppler reactivity feedback of the core fuel and the reactivity feedback of the moderator, and the temperature of the fuel and the moderator and the density of the moderator are calculated by the sub-channel module. Under the thermal feedback provided by the sub-channel module, the neutron kinetic module will calculate the power distribution information of each fuel rod, that is, obtain the core power distribution information of the current time step.

[0093] In some embodiments of the present disclosure, since the calculation steps of the sub-channel module and the neutron dynamics module may be different, it is necessary to select data close to the above-mentioned time step for calculation, that is, for each time step, select the target interaction data corresponding to the time nodes in the sub-channel module and the neutron dynamics module whose time interval with the time step is less than a preset threshold, that is, select the interaction data corresponding to the time node at which the time step converges in time, and use the above-mentioned target interaction data to realize the data interaction between the real sub-channel module and the neutron dynamics module, so as to complete the calculation of the nuclear power of the above-mentioned fuel rod, thereby ensuring the synchronization of the sub-channel module and the neutron dynamics module and solving the mismatch problem.

[0094] Step 105: Based on the core power distribution information, a neutron dynamics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain analysis results.

[0095] In some embodiments of the present disclosure, before step 105, the method may further include the following steps:

[0096] Preprocess the core power distribution information corresponding to each time step;

[0097] The preprocessed power distribution information of each core is stored in the array variable in chronological order;

[0098] Output all data in the array variable to the hot rod file;

[0099] Step 105 may specifically include: based on the hot rod file, using the neutron kinetics module to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results.

[0100] It should be noted that since each time step corresponds to the core power distribution information, that is, the core power distribution information is the three-dimensional fine power distribution data at the transient time point, and the data volume is very large, it cannot be directly used by the sub-channel model and the neutron kinetic model. The core power distribution information corresponding to each time step can be preprocessed, such as data screening of the core power distribution information to obtain target information such as lumped parameters, transient process time series, axial grid parameters, hot rod axial power distribution, core axial power distribution, core heat pipe factor, component heat pipe factor, core thermal power, and FDH of each fuel rod group. Each core power distribution information after preprocessing is stored in the array variable in chronological order. After completing the storage of all information, all data in the array variable is output to the hot rod file. Based on the hot rod file, the neutron kinetic module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain analysis results.

[0101] In one embodiment, the hot stick file may include any one or more of a binary format and a text format.

[0102] In one embodiment, the data can be written to the hot rod file block by block using a streaming output method. This can reduce memory usage and efficiently process various instantaneous parameters during the hot rod accident. Similarly, the hot rod file can be read using streaming reading technology to avoid loading the entire file into the memory at one time. This can process large-scale hot rod files without causing memory overflow. In the process of reading the hot rod file, the file segmentation is performed in accordance with the traversal strategy of the DNBR analysis, and data cleaning, format conversion, screening of specific data and other tasks are completed, thereby reducing the workload of subsequent processing.

[0103] Step 106: When the analysis result does not meet the preset conditions, determine that the reactor is unsafe.

[0104] In one embodiment, the analysis results may include data such as enthalpy rise, peak enthalpy, DNBR, and fuel pellet center peak temperature.

[0105] In some embodiments of the present disclosure, step 106 may include the following steps:

[0106] Step b1, determining the target fuel rod with the highest heat based on the core power distribution information.

[0107] Step b2, performing enthalpy analysis on the target fuel rod to obtain a first analysis result.

[0108] Step b3, performing DNBR analysis on the target fuel rod to obtain a second analysis result.

[0109] Step b4, when the second analysis result is greater than a preset threshold, perform DNBR analysis on each assembly respectively to obtain a third analysis result, and determine the number of fuel rods that do not meet the preset conditions based on the third analysis result; wherein the assembly includes at least one fuel rod.

[0110] Step b5: When at least one of the first analysis result and the second analysis result does not satisfy corresponding preset conditions, determining that the reactor is unsafe.

[0111] In some embodiments of the present disclosure, before step 101, the method may further include the following steps:

[0112] Step c1, obtaining model configuration information input by a user; the model configuration information includes parameter information and arrangement information of a target sub-model in the first model;

[0113] Step c2, when the target sub-model exists in the preset model library, the target sub-model is arranged according to the arrangement information to obtain a first model;

[0114] Step c3, if the target sub-model does not exist in the preset model library, generate the target sub-model based on the parameter information, arrange the target sub-model according to the arrangement information to obtain the first model, and store the target sub-model in the model library.

[0115] It can be understood that by modularizing the model and using pre-stored sub-models to generate the first model, the efficiency of building the model can be effectively improved.

[0116] As an example, the first model can be constructed by the following steps:

[0117] The division of axial and radial nodes in the first model must be consistent with the division of radial nodes in the second model corresponding to the neutron dynamics module used for core physics analysis. The regular arrangement of rod bundles enables the model to more accurately reflect the actual geometric structure of the reactor core, and this regular architecture provides a clear basis for the division of subchannels, providing clear boundary conditions and geometric parameters for the analysis process.

[0118] Based on the regular rod bundle architecture, a basic rod bundle unit is first accurately constructed, which should contain one or several rods and key geometric elements such as the surrounding coolant channels. Appropriate coordinate systems and reference points are set for the basic unit to facilitate subsequent replication and combination operations.

[0119] Then the number and arrangement of channels need to be determined, including: the number of rows and columns of rod bundles; the spacing between rod bundles in the row direction, column direction and the outside; the spacing between rod bundles in the row direction and column direction; the diameter of the rod bundle; the height of the rod bundle; and the number of axial division nodes.

[0120] Next, according to the arrangement of the channels, the connection relationship between the channels and the rods is generated in two different modes: at the intersection of the channels or at the center of the channels, and the gap connection relationship between the rod bundles is generated. Under the architecture of regular rod bundles, there are four types of channels: corner channels, side channels in the row direction, side channels in the column direction, and center channels. There are six types of gaps: corner channel-side channel gap in the row direction, corner channel-side channel gap in the column direction, gap between side channels in the row direction, gap between side channels in the column direction, center channel-side channel gap in the row direction, and center channel-side channel gap in the column direction.

[0121] By setting the direction, spacing, and number of replication parameters, the basic unit can be replicated multiple times according to the arrangement of the regular rod bundle architecture to form a larger-scale local structure. For multi-layer or complex rod bundle structures, replication and arrangement can be performed in layers, first building a layer or a region, and then gradually expanding to the entire geometric model. During the integration process, it is necessary to ensure that the connection and transition between different parts are natural, avoid geometric discontinuities or errors, and ensure that the replicated units accurately reflect the regular arrangement of the actual rod bundle.

[0122] After the first model is established, the corresponding sub-channel information includes: channel number, area, inclination, hot circumference, and wet circumference. The rod information includes: rod number, radius, number of material layers, number of radial layers, number of surface connections, number of surface connection channels, and surface connection share. The gap information includes: gap number, gap direction, gap width, number of channels on both sides of the gap, and distance from the channels on both sides.

[0123] It should be noted that the reactor safety assessment method proposed in the embodiment of the present disclosure can simulate rod ejection accidents according to different operating conditions, so as to determine whether rod ejection accidents occurring in the reactor under different operating conditions can be controlled within a safe range.

[0124] According to the reactor safety assessment method proposed in the embodiment of the present disclosure, the first process of simulating the rod ejection accident by controlling the pre-constructed first model to obtain the operation data of the first model in the first process; the first model and corresponding to the sub-channel module are grid-divided to obtain the first grid, and the second model corresponding to the neutron dynamics module is grid-divided to obtain the second grid; the first grid and the second grid are mapped to obtain the mapping relationship between the first grid and the second grid, and the equivalent weight factor corresponding to each mapping relationship; for each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron dynamics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step; based on the core power distribution information, the neutron dynamics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results; when the analysis results do not meet the preset conditions, it is determined that the reactor is unsafe. By internally coupling the subchannel module and the neutron dynamics module, that is, mapping the grids of the subchannel module and the neutron dynamics module, and configuring equivalent weight factors, the consistency of data interaction between the subchannel module and the neutron dynamics module when collaboratively calculating the core power distribution is ensured, avoiding the occurrence of mismatches, and improving the stability of calculations and evaluation efficiency. In addition, by simulating the rod ejection accident and using the operating data in the simulation process to evaluate the safety of the reactor, the accuracy of the reactor safety assessment results is improved, and the damage caused by the carbon rod accident to the equipment and the surrounding environment is reduced.

[0125] Figure 2 is a block diagram of a reactor safety assessment device according to an exemplary embodiment. Figure 2 The device includes a simulation unit 201, a division unit 202, a mapping unit 203, a first determination unit 204, an analysis unit 205 and a second determination unit 206.

[0126] The simulation unit 201 is used to control the first model constructed in advance to simulate the first process of the rod ejection accident, and obtain the operation data of the first model in the first process; the first process is the process of the fuel rod assembly ejecting the reactor; the first model is a three-dimensional model of the reactor core, and the first model is set in the sub-channel module;

[0127] A division unit 202 is used to perform grid division on the first model and corresponding to the sub-channel module to obtain a first grid, and to perform grid division on the second model corresponding to the neutron dynamics module to obtain a second grid;

[0128] A mapping unit 203 is used to map the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; the equivalent weight factor is used to control the reactivity equivalence of the first model and the second model;

[0129] A first determining unit 204 is used to determine the nuclear power of each fuel rod in the time step based on the equivalent weight factor using the sub-channel module and the neutron dynamics module for each time step in the first process, and obtain the core power distribution information of the time step;

[0130] The analysis unit 205 is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively based on the core power distribution information using a neutron kinetics module to obtain analysis results;

[0131] The second determination unit 206 is used to determine that the reactor is unsafe when the analysis result does not meet the preset condition.

[0132] In some embodiments of the present disclosure, the fuel rod parameters of the three-dimensional core model in the first model and the second model are consistent; the mapping unit 203 can be specifically used for:

[0133] According to the position information of the first grid and the position information of the second grid, each first grid is matched with at least one second grid to obtain a plurality of mapping relationships; the number of fuel rods in the first grid is the same as the total number of fuel rods in at least one second grid that matches;

[0134] For each first grid and at least one second grid corresponding to each mapping relationship, an equivalent weight factor is calculated according to temperature information corresponding to the first grid and temperature information corresponding to the at least one second grid.

[0135] In some embodiments of the present disclosure, the mapping unit 203 may also be used to:

[0136] Calculate a first temperature average value based on temperature information corresponding to the first grid, and calculate a second temperature average value based on temperature information corresponding to at least one second grid;

[0137] Normalizing the first temperature average value to obtain a first equivalent weight factor, and normalizing the second temperature average value to obtain a second equivalent weight factor;

[0138] After calculating the equivalent weight factor according to the temperature information corresponding to the first grid and the temperature information corresponding to at least one second grid, the method further includes:

[0139] determining whether the reactivity of the first grid and the at least one second grid is equivalent based on the first equivalent weight factor and the second equivalent weight factor;

[0140] determining a reactivity deviation value in the event that the reactivity of the first grid and at least one second grid are not equivalent;

[0141] Determining a first rate of change of the resonance cross section with temperature based on the first temperature average, and determining a second rate of change of the resonance cross section with temperature based on the second temperature average;

[0142] The first equivalent weight factor and the second equivalent weight factor are adjusted based on the reactivity deviation value, the first change rate and the second change rate, and the step of determining whether the reactivity of the first grid and the at least one second grid are equivalent is returned to be executed until the reactivity of the first grid and the at least one second grid are equivalent.

[0143] In some embodiments of the present disclosure, the device may further include:

[0144] A preprocessing unit, used for preprocessing the core power distribution information corresponding to each time step;

[0145] A storage unit, used for storing the preprocessed power distribution information of each core in the array variable in chronological order;

[0146] The core power distribution information corresponding to each time step is stored in the array variable in chronological order;

[0147] Output unit, used to output all data in array variables to hot rod file;

[0148] The analysis unit 205 is specifically used for:

[0149] Based on the hot rod file, the neutron dynamics module was used to perform enthalpy analysis and deviation from nucleate boiling ratio (DNBR) analysis, and the analysis results were obtained.

[0150] In some embodiments of the present disclosure, the second determining unit 206 may be specifically configured to:

[0151] Based on the core power distribution information, determine the target fuel rod with the highest heat;

[0152] Performing enthalpy analysis on the target fuel rod to obtain a first analysis result;

[0153] Performing DNBR analysis on the target fuel rod to obtain a second analysis result;

[0154] When the second analysis result is greater than a preset threshold, performing DNBR analysis on each assembly to obtain a third analysis result, and determining the number of fuel rods that do not meet the preset condition based on the third analysis result; wherein the assembly includes at least one fuel rod;

[0155] When at least one of the first analysis result and the second analysis result does not satisfy a corresponding preset condition, it is determined that the reactor is unsafe.

[0156] In some embodiments of the present disclosure, the device may further include:

[0157] An acquisition unit, configured to acquire model configuration information input by a user; the model configuration information includes parameter information and arrangement information of a target sub-model in the first model;

[0158] An arrangement unit, configured to arrange the target sub-model according to the arrangement information to obtain a first model when the target sub-model exists in the preset model library;

[0159] A generating unit is used to generate a target sub-model based on parameter information when the target sub-model does not exist in a preset model library, arrange the target sub-model according to the arrangement information to obtain a first model, and store the target sub-model in the model library.

[0160] Regarding the device in the above embodiment, the specific manner in which each module performs operations has been described in detail in the embodiment of the method, and will not be elaborated here.

[0161] According to the reactor safety assessment device proposed in the embodiment of the present disclosure, the first process of simulating the rod ejection accident by controlling the pre-constructed first model to obtain the operation data of the first model in the first process; the first model and corresponding to the sub-channel module are grid-divided to obtain the first grid, and the second model corresponding to the neutron dynamics module is grid-divided to obtain the second grid; the first grid and the second grid are mapped to obtain the mapping relationship between the first grid and the second grid, and the equivalent weight factor corresponding to each mapping relationship; for each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron dynamics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step; based on the core power distribution information, the neutron dynamics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results; when the analysis results do not meet the preset conditions, it is determined that the reactor is unsafe. By internally coupling the subchannel module and the neutron dynamics module, that is, mapping the grids of the subchannel module and the neutron dynamics module, and configuring equivalent weight factors, the consistency of data interaction between the subchannel module and the neutron dynamics module when collaboratively calculating the core power distribution is ensured, avoiding the occurrence of mismatches, and improving the stability of calculations and evaluation efficiency. In addition, by simulating the rod ejection accident and using the operating data in the simulation process to evaluate the safety of the reactor, the accuracy of the reactor safety assessment results is improved, and the damage caused by the carbon rod accident to the equipment and the surrounding environment is reduced.

[0162] Figure 33 is a block diagram of a device for a reactor safety assessment method according to an exemplary embodiment. For example, the device 300 may be an electronic device, such as a mobile phone, a computer, a digital broadcast terminal, a messaging device, a game console, a tablet device, a medical device, a fitness device, a personal digital assistant, etc.

[0163] Reference Figure 3 , the device 300 may include one or more of the following components: a processing component 302 , a memory 304 , a power component 306 , a multimedia component 308 , an audio component 310 , an input / output (I / O) interface 312 , a sensor component 314 , and a communication component 316 .

[0164] The processing component 302 generally controls the overall operation of the device 300, such as operations associated with display, phone calls, data communications, camera operations, and recording operations. The processing component 302 may include one or more processors 320 to execute instructions to complete all or part of the steps of the above-mentioned method. In addition, the processing component 302 may include one or more modules to facilitate the interaction between the processing component 302 and other components. For example, the processing component 302 may include a multimedia module to facilitate the interaction between the multimedia component 308 and the processing component 302.

[0165] The memory 304 is configured to store various types of data to support operations on the device 300. Examples of such data include instructions for any application or method operating on the device 300, contact data, phone book data, messages, pictures, videos, etc. The memory 304 can be implemented by any type of volatile or non-volatile storage device or a combination thereof, such as static random access memory (SRAM), electrically erasable programmable read-only memory (EEPROM), erasable programmable read-only memory (EPROM), programmable read-only memory (PROM), read-only memory (ROM), magnetic memory, flash memory, magnetic disk or optical disk.

[0166] The power component 306 provides power to the various components of the device 300. The power component 306 may include a power management system, one or more power supplies, and other components associated with generating, managing, and distributing power for the device 300.

[0167] The multimedia component 308 includes a screen that provides an output interface between the device 300 and the user. In some embodiments, the screen may include a liquid crystal display (LCD) and a touch panel (TP). If the screen includes a touch panel, the screen may be implemented as a touch screen to receive input signals from the user. The touch panel includes one or more touch sensors to sense touch, slide, and gestures on the touch panel. The touch sensor may not only sense the boundaries of the touch or slide action, but also detect the duration and pressure associated with the touch or slide operation. In some embodiments, the multimedia component 308 includes a front camera and / or a rear camera. When the device 300 is in an operating mode, such as a shooting mode or a video mode, the front camera and / or the rear camera may receive external multimedia data. Each front camera and rear camera may be a fixed optical lens system or have a focal length and optical zoom capability.

[0168] The audio component 310 is configured to output and / or input audio signals. For example, the audio component 310 includes a microphone (MIC), and when the device 300 is in an operating mode, such as a call mode, a recording mode, and a speech recognition mode, the microphone is configured to receive an external audio signal. The received audio signal can be further stored in the memory 304 or sent via the communication component 316. In some embodiments, the audio component 310 also includes a speaker for outputting audio signals.

[0169] I / O interface 312 provides an interface between processing component 302 and peripheral interface modules, such as keyboards, click wheels, buttons, etc. These buttons may include but are not limited to: a home button, a volume button, a start button, and a lock button.

[0170] The sensor assembly 314 includes one or more sensors for providing various aspects of status assessment for the device 300. For example, the sensor assembly 314 can detect the open / closed state of the device 300, the relative positioning of components, such as the display and keypad of the device 300, the sensor assembly 314 can also detect the position change of the device 300 or a component of the device 300, the presence or absence of user contact with the device 300, the orientation or acceleration / deceleration of the device 300, and the temperature change of the device 300. The sensor assembly 314 may include a proximity sensor configured to detect the presence of nearby objects without any physical contact. The sensor assembly 314 may also include an optical sensor, such as a CMOS or CCD image sensor, for use in imaging applications. In some embodiments, the sensor assembly 314 may also include an accelerometer, a gyroscope sensor, a magnetic sensor, a pressure sensor, or a temperature sensor.

[0171] The communication component 316 is configured to facilitate wired or wireless communication between the device 300 and other devices. The device 300 can access a wireless network based on a communication standard, such as WiFi, 2G or 3G, or a combination thereof. In an exemplary embodiment, the communication component 316 receives a broadcast signal or broadcast-related information from an external broadcast management system via a broadcast channel. In an exemplary embodiment, the communication component 316 also includes a near field communication (NFC) module to facilitate short-range communication. For example, the NFC module can be implemented based on radio frequency identification (RFID) technology, infrared data association (IrDA) technology, ultra-wideband (UWB) technology, Bluetooth (BT) technology and other technologies.

[0172] In an exemplary embodiment, the apparatus 300 may be implemented by one or more application-specific integrated circuits (ASICs), digital signal processors (DSPs), digital signal processing devices (DSPDs), programmable logic devices (PLDs), field programmable gate arrays (FPGAs), controllers, microcontrollers, microprocessors or other electronic components to perform the above method.

[0173] In an exemplary embodiment, a non-transitory computer-readable storage medium including instructions is also provided, such as a memory 304 including instructions, and the instructions can be executed by the processor 320 of the device 300 to perform the above method. For example, the non-transitory computer-readable storage medium can be a ROM, a random access memory (RAM), a CD-ROM, a magnetic tape, a floppy disk, an optical data storage device, etc.

[0174] In an exemplary embodiment, a computer program product is also provided, comprising a computer program, which implements the above method when executed by the processor 320 of the device 300 .

[0175] Those skilled in the art will readily appreciate other embodiments of the present invention after considering the specification and practicing the invention disclosed herein. The present disclosure is intended to cover any variations, uses or adaptations of the present invention that follow the general principles of the present invention and include common knowledge or customary techniques in the art that are not disclosed in the present disclosure. The description and examples are to be considered exemplary only, and the true scope and spirit of the present invention are indicated by the following claims.

[0176] It should be understood that the present invention is not limited to the exact construction that has been described above and shown in the drawings and that various modifications and changes may be made without departing from the scope thereof. The scope of the present invention is limited only by the appended claims.

Claims

1. A reactor safety assessment method, characterized in that: include: Controlling a pre-built first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process; The first process is a process in which a fuel rod assembly is ejected from the reactor; The first model is a three-dimensional model of the reactor core, and the first model is arranged in the sub-channel module; Meshing the first model and corresponding to the sub-channel module to obtain a first mesh, and meshing the second model corresponding to the neutron dynamics module to obtain a second mesh; Mapping the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; The equivalent weight factor is used to control the reactivity equivalence of the first model and the second model; For each time step in the first process, based on the equivalent weight factor, the sub-channel module and the neutron kinetics module are used to determine the nuclear power of each fuel rod in the time step to obtain the core power distribution information of the time step; Based on the core power distribution information, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain analysis results; When the analysis result does not satisfy a preset condition, it is determined that the reactor is unsafe.

2. The reactor safety assessment method according to claim 1, characterized in that: The fuel rod parameters of the three-dimensional core model in the first model and the second model are consistent; The mapping process of the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship, includes: According to the position information of the first grid and the position information of the second grid, each first grid is matched with at least one second grid to obtain a plurality of the mapping relationships; the number of fuel rods in the first grid is the same as the total number of fuel rods in the matched at least one second grid; For each first grid and at least one second grid corresponding to each mapping relationship, the equivalent weight factor is calculated according to temperature information corresponding to the first grid and temperature information corresponding to the at least one second grid.

3. The reactor safety assessment method according to claim 2, characterized in that: The equivalent weight factor includes a first equivalent weight factor and a second equivalent weight factor The calculating the equivalent weight factor according to the temperature information corresponding to the first grid and the temperature information corresponding to the at least one second grid includes: Calculating a first temperature average value based on the temperature information corresponding to the first grid, and calculating a second temperature average value based on the temperature information corresponding to the at least one second grid; Normalizing the first temperature average value to obtain a first equivalent weight factor, and normalizing the second temperature average value to obtain a second equivalent weight factor; After calculating the equivalent weight factor according to the temperature information corresponding to the first grid and the temperature information corresponding to the at least one second grid, the method further includes: determining whether the reactivity of the first grid and the at least one second grid is equivalent based on the first equivalent weight factor and the second equivalent weight factor; determining a reactivity deviation value in the event that the reactivity of the first grid and the at least one second grid are not equivalent; determining a first rate of change of the resonance cross section with temperature based on the first temperature average value, and determining a second rate of change of the resonance cross section with temperature based on the second temperature average value; The first equivalent weight factor and the second equivalent weight factor are adjusted based on the reactivity deviation value, the first change rate and the second change rate, and the step of determining whether the reactivity of the first grid and the at least one second grid are equivalent is returned to be executed until the reactivity of the first grid and the at least one second grid are equivalent.

4. The reactor safety assessment method according to claim 1, characterized in that: Before performing enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis based on the core power distribution information and obtaining analysis results by using the neutron kinetics module, the method further includes: Preprocess the core power distribution information corresponding to each time step; The preprocessed power distribution information of each core is stored in the array variable in chronological order; Output all data in the array variable to a hot rod file; Based on the core power distribution information, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis respectively to obtain analysis results, including: Based on the hot rod file, the neutron kinetics module is used to perform enthalpy analysis and deviation from nucleate boiling ratio DNBR analysis to obtain the analysis results.

5. The reactor safety assessment method according to claim 1, characterized in that: When the analysis result does not meet the preset condition, determining that the reactor is unsafe includes: Determining a target fuel rod with the highest heat based on the core power distribution information; Performing enthalpy analysis on the target fuel rod to obtain a first analysis result; Performing DNBR analysis on the target fuel rod to obtain a second analysis result; When the second analysis result is greater than a preset threshold, performing DNBR analysis on each assembly to obtain a third analysis result, and determining the number of fuel rods that do not meet the preset condition based on the third analysis result; wherein the assembly includes at least one fuel rod; When at least one of the first analysis result and the second analysis result does not satisfy a corresponding preset condition, it is determined that the reactor is unsafe.

6. The reactor safety assessment method according to claim 1, characterized in that: Before the first process of controlling the pre-built reactor core three-dimensional model to simulate the rod ejection accident, the method further includes: Acquire model configuration information input by a user; the model configuration information includes parameter information and arrangement information of a target sub-model in the first model; When the target sub-model exists in the preset model library, the target sub-model is arranged according to the arrangement information to obtain the first model; When the target sub-model does not exist in the preset model library, the target sub-model is generated based on the parameter information, the target sub-model is arranged according to the arrangement information to obtain the first model, and the target sub-model is stored in the model library.

7. A reactor safety assessment device, characterized in that: include: A simulation unit, used for controlling a pre-built first model to simulate a first process of a rod ejection accident, and obtaining operation data of the first model in the first process; The first process is a process in which a fuel rod assembly is ejected from the reactor; The first model is a three-dimensional model of the reactor core, and the first model is arranged in the sub-channel module; A partitioning unit, configured to perform mesh partitioning on the first model and corresponding to the sub-channel module to obtain a first mesh, and to perform mesh partitioning on the second model corresponding to the neutron dynamics module to obtain a second mesh; A mapping unit, configured to perform mapping processing on the first grid and the second grid to obtain a mapping relationship between the first grid and the second grid, and an equivalent weight factor corresponding to each mapping relationship; The equivalent weight factor is used to control the reactivity equivalence of the first model and the second model; A first determining unit is used for determining, for each time step in the first process, based on the equivalent weight factor, using the sub-channel module and the neutron kinetics module to determine the nuclear power of each fuel rod in the time step, so as to obtain the core power distribution information of the time step; An analysis unit, configured to perform enthalpy analysis and deviation from nucleate boiling ratio (DNBR) analysis respectively based on the core power distribution information and using the neutron kinetics module to obtain analysis results; The second determination unit is used to determine that the reactor is unsafe if the analysis result does not meet a preset condition.

8. An electronic device, characterized in that: include: A memory, a processor, and a computer program stored in the memory and executable on the processor, wherein when the processor executes the computer program, the method according to any one of claims 1 to 6 is implemented.

9. A computer-readable storage medium having a computer program stored thereon, characterized in that: When the computer program is executed by a processor, the method according to any one of claims 1 to 6 is implemented.

10. A computer program product, comprising a computer program, characterized in that The computer program implements the method according to any one of claims 1 to 6 when executed by a processor.

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