Data processing method, device and medium for primary circuit water pressure test of nuclear power plant

Abstract

The present application relates to a data processing method, apparatus, computer equipment, storage medium, and computer program product for a nuclear power plant primary circuit hydrostatic test. The method comprises: determining, based on a final safety analysis report, a target accident that causes the pressure of the primary circuit system to exceed a preset design pressure; determining, from a set of initiating event classes, a target initiating event class that matches the target accident; determining the frequency of occurrence of the target initiating event class based on preset historical event information of nuclear power units; determining the rupture probability level corresponding to pipelines in the primary circuit system and in support systems connected to the primary circuit system; determining an assessment result based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and a probabilistic safety analysis model; and determining whether the test results of the sealing test performed on the primary circuit system during each refueling cycle can be used as a substitute for the test results of the primary circuit hydrostatic test. This method can improve the accuracy of the assessment results.

Description

Technical Field

[0001] The present application relates to the field of nuclear power technology, and in particular to a data processing method, device, computer equipment, storage medium and computer program product for a primary circuit water pressure test of a nuclear power plant. Background Art

[0002] In the nuclear power sector, the primary circuit hydrostatic test is a strength test of the high-pressure portion of a nuclear power unit's primary circuit system and related auxiliary systems at an appropriate test pressure. This test verifies the seals and weld quality of the primary circuit system's equipment and piping, and verifies its tightness and safety during pressurized operation. This demonstrates that the reactor's primary circuit system is safe under normal operation and designed accident conditions, and complies with nuclear safety regulations, from the end of the test until the next test. During the primary circuit hydrostatic test, the system pressure is increased to 1.2 times the design pressure, increasing the impact on the integrity of the primary circuit system's boundaries. Therefore, nuclear power plants intend to use the results of other tests instead of the primary circuit hydrostatic test to verify the tightness and safety of the primary circuit system.

[0003] However, when using the test results of other tests to replace the test results of the primary-loop water pressure test, the test results of other tests are usually inspected by manual visual inspection to evaluate whether the test results of other tests can replace the test results of the primary-loop water pressure test, which has the problem of low accuracy of the evaluation results. Summary of the Invention

[0004] Based on this, it is necessary to provide a data processing method, device, computer equipment, computer-readable storage medium and computer program product for the primary circuit water pressure test of a nuclear power plant, which can improve the accuracy of the evaluation results in response to the above technical problems.

[0005] In a first aspect, the present application provides a data processing method for a primary circuit hydraulic test of a nuclear power plant, comprising:

[0006] Determining, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure;

[0007] Determining a target initiating event class that matches the target accident from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model;

[0008] Determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit according to preset nuclear power unit historical event information that matches the type of the nuclear power unit;

[0009] Determining the rupture possibility levels corresponding to the pipelines in the primary loop system and the support system connected to the primary loop system according to a preset pipeline degradation mechanism judgment criterion table and a preset pipeline rupture possibility level table;

[0010] An assessment result is determined based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test result of the sealing test performed on the primary circuit system during each material change cycle can be used as an alternative test result for the primary circuit water pressure test.

[0011] In a second aspect, the present application further provides a data processing device for a primary circuit hydraulic test of a nuclear power plant, comprising:

[0012] An initiating event class management module is configured to determine, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure; determine, from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model, a target initiating event class that matches the target accident; and determine, based on preset nuclear power unit historical event information that matches the type of the nuclear power unit, an occurrence frequency of the target initiating event class corresponding to the nuclear power unit;

[0013] a pipeline information management module, configured to determine, based on a preset pipeline degradation mechanism determination criterion table and a preset pipeline rupture probability level table, corresponding rupture probability levels of pipelines in the primary loop system and in a support system connected to the primary loop system;

[0014] An evaluation module is configured to determine an evaluation result based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the evaluation result indicates whether the test result of the sealing test performed on the primary circuit system during each refueling cycle can be used as an alternative test result for the primary circuit water pressure test.

[0015] In a third aspect, the present application further provides a computer device comprising a memory and a processor, wherein the memory stores a computer program, and when the processor executes the computer program, the following steps are implemented:

[0016] Determining, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure;

[0017] Determining a target initiating event class that matches the target accident from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model;

[0018] Determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit according to preset nuclear power unit historical event information that matches the type of the nuclear power unit;

[0019] Determining the rupture possibility levels corresponding to the pipelines in the primary loop system and the support system connected to the primary loop system according to a preset pipeline degradation mechanism judgment criterion table and a preset pipeline rupture possibility level table;

[0020] An assessment result is determined based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test result of the sealing test performed on the primary circuit system during each material change cycle can be used as an alternative test result for the primary circuit water pressure test.

[0021] In a fourth aspect, the present application further provides a computer-readable storage medium having a computer program stored thereon, wherein when the computer program is executed by a processor, the following steps are implemented:

[0022] Determining, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure;

[0023] Determining a target initiating event class that matches the target accident from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model;

[0024] Determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit according to preset nuclear power unit historical event information that matches the type of the nuclear power unit;

[0025] Determining the rupture possibility levels corresponding to the pipelines in the primary loop system and the support system connected to the primary loop system according to a preset pipeline degradation mechanism judgment criterion table and a preset pipeline rupture possibility level table;

[0026] An assessment result is determined based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test result of the sealing test performed on the primary circuit system during each material change cycle can be used as an alternative test result for the primary circuit water pressure test.

[0027] In a fifth aspect, the present application further provides a computer program product, comprising a computer program, which, when executed by a processor, implements the following steps:

[0028] Determining, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure;

[0029] Determining a target initiating event class that matches the target accident from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model;

[0030] Determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit according to preset nuclear power unit historical event information that matches the type of the nuclear power unit;

[0031] Determining the rupture possibility levels corresponding to the pipelines in the primary loop system and the support system connected to the primary loop system according to a preset pipeline degradation mechanism judgment criterion table and a preset pipeline rupture possibility level table;

[0032] An assessment result is determined based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test result of the sealing test performed on the primary circuit system during each material change cycle can be used as an alternative test result for the primary circuit water pressure test.

[0033] The above-mentioned data processing method, apparatus, computer equipment, storage medium, and computer program product for the primary circuit hydrostatic test of a nuclear power plant, compared to the primary circuit hydrostatic test, cannot verify the leaktightness of the primary circuit system when the pressure exceeds the preset design pressure during operation, nor can it verify the leaktightness of the pipe between two isolation valves and the second valve in the support system connected to the primary circuit system. Therefore, by identifying a target accident that causes the pressure of the primary circuit system to exceed the preset design pressure, and then determining a target initiating event class that matches the target accident and the occurrence frequency of the target initiating event class, and further determining the corresponding rupture probability level of the pipes in the primary circuit system and the support system, it is possible to evaluate the portions that cannot be verified by the leaktightness test during each refueling cycle. Furthermore, based on the corresponding rupture probability level of the pipes, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model, compared to manual visual inspection, it is possible to more accurately determine the evaluation results indicating whether the test results of the leaktightness test conducted during each refueling cycle can be used as a substitute for the test results of the primary circuit hydrostatic test. BRIEF DESCRIPTION OF THE DRAWINGS

[0034] In order to more clearly illustrate the technical solutions in the embodiments of the present application or related technologies, the following briefly introduces the drawings required for use in the embodiments or related technical descriptions. Obviously, the drawings described below are only some embodiments of the present application. For ordinary technicians in this field, other drawings can be obtained based on these drawings without paying any creative work.

[0035] Figure 11 is a flow chart of a data processing method for a primary circuit water pressure test of a nuclear power plant in one embodiment;

[0036] Figure 2 A schematic diagram of risk assessment guidance under the core damage indicator dimension in one embodiment;

[0037] Figure 3 This is a schematic diagram of risk assessment guidance under the radioactive release indicator dimension in one embodiment;

[0038] Figure 4 A schematic flow chart of the steps for evaluating changes to a primary circuit water pressure test in one embodiment;

[0039] Figure 5 is a structural block diagram of a data processing device for a primary circuit hydraulic test of a nuclear power plant in one embodiment;

[0040] Figure 6 FIG. 1 is a diagram showing the internal structure of a computer device in one embodiment. DETAILED DESCRIPTION

[0041] In order to make the purpose, technical solutions and advantages of this application more clear, the following further describes this application in detail with reference to the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain this application and are not intended to limit this application.

[0042] In one embodiment, Figure 1 As shown, a data processing method for the primary circuit water pressure test of a nuclear power plant is provided. This embodiment uses the method applied to a computer device as an example for illustration. It is understandable that the computer device can be a terminal or a server. The method can also be applied to a system including a terminal and a server and implemented through the interaction between the terminal and the server. The terminal can be a personal computer, a laptop, a smartphone, or a tablet computer. The server can be implemented as an independent server or a server cluster consisting of multiple servers. In this embodiment, the method includes the following steps:

[0043] Step 102 : determining, based on a final safety analysis report pre-generated for the nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure.

[0044] A nuclear power plant is a mechanism for converting nuclear power into electrical energy to provide electricity. A nuclear power plant may include one or more nuclear power units. A nuclear power unit is a basic power generation unit consisting of a reactor, an associated steam turbine generator unit, and the systems and facilities required to maintain their normal operation and ensure safety. The primary coolant system in a nuclear power unit includes the cooling system, pressure regulation system, and overpressure protection system. The primary coolant system is located within the containment vessel and can be connected to support systems within the containment vessel that support the safe operation of the primary coolant system. The containment vessel is an enclosed container that houses the primary coolant system and is used to prevent and control the leakage of radioactive materials from the reactor. Support systems may include secondary coolant systems, safety systems, or other systems. The secondary coolant system is a secondary coolant circulation system used to remove heat from the primary coolant. Safety systems include safety injection systems, containment spray systems, or containment isolation systems. The preset design pressure is the maximum pressure set for the primary coolant system.

[0045] The Final Safety Analysis Report (FSAR) is used to describe information about the nuclear power plant and its operating conditions, so that regulatory agencies can evaluate the safety of the nuclear power plant. The Final Safety Analysis Report may include the safety requirements, design basis, system equipment design, operating limits, safety analysis and other information of the nuclear power plant. The design basis section of the Final Safety Analysis Report may describe accidents that may be expected to occur in the nuclear power plant, including design basis accidents and beyond-design basis accidents. Design basis accidents are accidents for which the nuclear power plant is pre-configured with solutions. Beyond-design basis accidents are accidents whose severity exceeds that of design basis accidents. When the impact described for the accident includes causing a change in the pressure of a primary circuit system, the Final Safety Analysis Report also records the pressure of the primary circuit system caused by the accident; the pressure of the primary circuit system caused by the accident refers to the maximum pressure that the primary circuit system withstands after the pressure of the primary circuit system changes after the accident occurs.

[0046] For example, the preset design pressure could be 17 MPa (megapascals). The accident described in the final safety analysis report could be a steam turbine trip, which could result in a primary-circuit system pressure of 18.25 MPa; a main feedwater system pipe rupture, which could result in a primary-circuit system pressure of 18.24 MPa; a main pump shaft jam, which could result in a primary-circuit system pressure of 17.82 MPa; or a rod popping accident, which could result in a primary-circuit system pressure of 16.40 MPa. Based on the above examples, the target accidents could be a steam turbine trip, a main feedwater system pipe rupture, and a main pump shaft jam.

[0047] In one embodiment, a computer device may obtain an accident impact data table generated based on a pre-generated final safety analysis report for a nuclear power plant, and determine, based on the accident impact data table, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure. The accident impact data table may record a correspondence between accidents that cause a change in the pressure of the primary circuit system and the pressure of the primary circuit system caused by the accident.

[0048] In one embodiment, the computer device can extract design basis accidents and beyond design basis accidents from the final safety analysis report as accidents, and extract the impact of each accident from the final safety analysis report. From the extracted accidents, the computer device can screen out accidents whose impact characteristics cause changes in the pressure of the primary circuit system, obtain the pressure of the primary circuit system caused by the screened accidents, establish a correspondence between the screened accidents and the pressure of the primary circuit system caused by the accidents, and generate an accident impact data table based on the correspondence.

[0049] Step 104 : Determine a target initiating event class that matches the target accident from the initiating event class set configured for the pre-generated probabilistic safety analysis model.

[0050] Probabilistic safety analysis models are used to conduct probabilistic safety analysis of nuclear power plants, obtaining quantitative results and assessing the risks of these plants based on these quantitative results. Probabilistic safety analysis models are based on event tree analysis and fault tree analysis. Probabilistic safety analysis can be abbreviated as PSA (Probabilistic Safety Analysis). In some scenarios, probabilistic safety analysis can also be referred to as probabilistic risk assessment (PRA).

[0051] The risks of a nuclear power plant can include the risk of core damage to the reactor and the risk of radioactive material release from the plant's containment vessel. The core damage risk can be quantified by the core damage indicator, while the risk of radioactive material release from the plant's containment vessel can be quantified by the radioactive release indicator.

[0052] The pre-generated probabilistic safety analysis model can be a probabilistic safety analysis model pre-generated in a preset probabilistic safety analysis software, specifically a probabilistic safety analysis model configured by default in the preset probabilistic safety analysis software. The preset probabilistic safety analysis software can be Risk Spectrum software (computer software for probabilistic safety analysis developed by the Swedish company Relcom), or RiskA software (computer software for probabilistic safety analysis developed by the FDS team of the Chinese Academy of Sciences). The set of initiating event classes is pre-configured for the probabilistic safety analysis model, and the set of initiating event classes may include multiple preset initiating event classes. Multiple initiating event classes may include a secondary circuit transient event class that causes a temperature rise in the primary circuit, an off-site power event class, a main water supply loss event class, a water supply pipe rupture accident, or others.

[0053] An initiating event is a disturbance occurring in a nuclear power plant that may cause damage to the reactor core. It is the initial event leading to an accident. If remedial measures are not taken after the initiating event, it will lead to further accidents. An initiating event can also be called the primary cause.

[0054] In one embodiment, a computer device may obtain an accident initiating event relationship table that records the correspondence between preset accidents and initiating event classes in a set of initiating event classes, determine a preset accident that matches a target accident from the accident initiating event relationship table, and determine the initiating event class that has a corresponding relationship with the preset accident in the accident initiating event relationship table as the target initiating event class that matches the target accident in the set of initiating event classes. The preset accident that matches the target accident may be a preset accident with the same textual content as the target accident, or may be a preset accident that is semantically closest to the target accident among the preset accidents recorded in the accident initiating event relationship table.

[0055] For example, if the target accident is a turbine trip, the target initiating event class could be a secondary circuit system transient event class that causes a temperature rise in the primary circuit system. In this case, the target initiating event class could be ST2A. If the target accident is a loss of normal feedwater flow, the target initiating event class could be a loss of main feedwater event class. In this case, the target initiating event class could be SW1A.

[0056] Step 106 : determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit based on preset nuclear power unit historical event information that matches the type of the nuclear power unit.

[0057] Among them, the type of nuclear power unit may include pressurized water reactor type, heavy water reactor type, boiling water reactor type or others. The preset nuclear power unit historical event information may include the correspondence between the preset nuclear power unit historical events and the preset initiating event class. The preset nuclear power unit historical events are historical events of multiple sample nuclear power units that are pre-set in the preset nuclear power unit historical event information and match the type of the nuclear power unit. The type of the sample nuclear power unit may be the same as the type of the nuclear power unit. The pre-set initiating event class is an initiating event class that is pre-set in the preset nuclear power unit historical event information and exists in the initiating event class set.

[0058] In one embodiment, a computer device can determine the length of the time span between the earliest historical event and the latest historical event in the preset nuclear power unit historical event information that matches the type of nuclear power unit, and determine the historical events that match the target initiating event class from the preset nuclear power unit historical event information, count the number of first events that match the target initiating event class, and count the number of units of the sample nuclear power units recorded in the preset nuclear power unit historical event information, and determine the occurrence frequency of the target initiating event class corresponding to the nuclear power unit based on the counted number of first events, the number of units, and the length of the time span. The length of the time span is the length of the interval between the occurrence time of the earliest historical event and the occurrence time of the latest historical event. The length of the time span can be in years.

[0059] In one embodiment, the computer device can determine a first ratio of the number of first events to the number of units, and determine a second ratio between the first ratio and the length of the time span, and determine the second ratio as the occurrence frequency of the target initiating event class corresponding to the nuclear power unit.

[0060] Step 108 : Determine the corresponding rupture possibility levels of the pipelines in the primary loop system and the support system connected to the primary loop system according to the preset pipeline degradation mechanism judgment criterion table and the preset pipeline rupture possibility level table.

[0061] Among them, the pipeline is a channel connecting the various devices in the primary circuit system and the supporting system, or a channel set inside each device. The preset pipeline degradation mechanism judgment criterion table is a pre-set table for evaluating the types of degradation mechanisms that may exist in the pipeline. The preset pipeline degradation mechanism judgment criterion table may include the correspondence between the degradation mechanism type, the judgment criterion, and the area in the pipeline where the degradation mechanism is prone to occur. The degradation mechanism type is the type of mechanism that the pipeline may suffer that causes the pipeline quality to age. The judgment criterion is the criterion that the system meets when determining that a degradation mechanism of the corresponding degradation mechanism type exists in the pipeline. The preset pipeline rupture possibility level table is used to determine the rupture possibility level of the pipeline. The rupture possibility level may include levels representing high rupture possibility, low rupture possibility, and medium rupture possibility.

[0062] In one embodiment, the computer device can determine the degradation mechanism determination results corresponding to the pipelines in the primary loop system and the pipelines in the support system connected to the primary loop system based on a preset pipeline degradation mechanism determination criteria table; and determine the rupture probability levels corresponding to the pipelines in the primary loop system and the support system based on a preset pipeline rupture probability level table and the corresponding degradation mechanism determination results of the pipelines. The degradation mechanism determination results include the absence of a degradation mechanism or the presence of a degradation mechanism type.

[0063] In one embodiment, a preset pipeline rupture possibility level table records the correspondence between preset rupture possibility levels and preset degradation mechanism determination results. In this embodiment, the computer device can determine the degradation mechanism determination results corresponding to pipelines in the primary loop system and the support system connected to the primary loop system based on the preset pipeline degradation mechanism determination criteria table; and use the preset rupture possibility level corresponding to the degradation mechanism determination result corresponding to the pipeline in the preset pipeline rupture possibility level table as the rupture possibility level corresponding to the pipeline.

[0064] Step 110 , determining an assessment result based on the corresponding rupture probability level of the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test results of the sealing test performed on the primary circuit system during each refueling cycle can be used as an alternative test result for the primary circuit water pressure test.

[0065] Among them, in the single-circuit water pressure test, water is injected into the single-circuit system during the test, causing the pressure of the single-circuit system to rise to 1.2 times the preset design pressure. This can test the sealing of the single-circuit system when the pressure of the single-circuit system rises to 1.2 times the preset design pressure. Moreover, the pressure boundary that can be tested by the single-circuit water pressure test includes all equipment in the single-circuit system, as well as the connecting pipes at the connection between the single-circuit system and the support system, and the valves set at both ends of the connecting pipes in the single-circuit system and the support system respectively.

[0066] The refueling cycle is the period for replacing equipment and materials in the primary-circuit system. After each refueling is completed and the primary-circuit system is in a cold shutdown state, a sealing test can be performed. The sealing test performed during the refueling cycle does not require the addition of water to the primary-circuit system, and cannot increase the pressure of the primary-circuit system to 1.2 times the preset design pressure. The pressure boundary that can be tested by the sealing test performed during the refueling cycle only includes all equipment in the primary-circuit system. Compared with the primary-circuit water pressure test, the sealing test performed during the refueling cycle cannot test the sealing condition when the pressure of the primary-circuit system exceeds 1 times the preset design pressure, nor can it test the connecting pipes at the connection between the primary-circuit system and the support system, nor the valves connecting the connecting pipes and the support system.

[0067] Since the target initiating event class is determined based on the target accident that causes the pressure of the primary circuit system to exceed the preset design pressure, the evaluation result is determined according to the corresponding rupture possibility level of the pipeline, the occurrence frequency of the target initiating event class and the probabilistic safety analysis model. To a certain extent, it can reflect the evaluation results of the sealing test conducted during the refueling cycle compared to the part that cannot be tested by the primary circuit water pressure test, thereby determining whether the test results of the sealing test conducted in each refueling cycle should be used as an alternative test result of the primary circuit water pressure test.

[0068] In one embodiment, the computer device can determine the occurrence frequency of the breach initiating event class that matches the probabilistic safety analysis model based on the corresponding rupture probability level of the pipeline; and determine the output evaluation result based on the occurrence frequency of the breach initiating event class, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model.

[0069] In the above-mentioned data processing method for the primary circuit water pressure test of a nuclear power plant, compared with the primary circuit water pressure test, since the sealing test conducted during each refueling cycle cannot verify the sealing when the pressure exceeds the preset design pressure during the operation of the primary circuit system, and cannot verify the pipeline between the two isolation valves and the second valve in the support system connected to the primary circuit system, therefore, by determining the target accident that causes the pressure of the primary circuit system to exceed the preset design pressure, and then determining the target initiating event class that matches the target accident, and determining the occurrence frequency of the target initiating event class, and further determining the corresponding rupture probability level of the pipelines in the primary circuit system and the support system, it is possible to evaluate the parts that cannot be verified by the sealing test conducted during each refueling cycle. Then, based on the corresponding rupture probability level of the pipeline, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model, compared with the manual visual method, it is possible to more accurately determine the evaluation results that indicate whether the test results of the sealing test conducted during each refueling cycle can be used as a substitute test result for the primary circuit water pressure test.

[0070] In one embodiment, step 102 includes: obtaining a final safety analysis report pre-generated for the nuclear power plant; determining, from the final safety analysis report, a set of accidents that cause a change in the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant; and determining, from the set of accidents, a target accident that causes the pressure of the primary circuit system to exceed a preset design pressure.

[0071] The accident set includes multiple accidents that cause pressure changes in the primary circuit system of the nuclear power units in the nuclear power plant.

[0072] In this embodiment, by determining the set of accidents that cause changes in the pressure of the primary circuit system from the final safety analysis report, the set of accidents that cause disturbances to the pressure of the primary circuit system can be determined, and then the target accident that causes the pressure of the primary circuit system to exceed the preset design pressure can be determined from the accident set. By combining the evaluation results obtained in subsequent steps, it is possible to evaluate the situation where the pressure that cannot be verified by the sealing test performed on the primary circuit system in each refueling cycle exceeds the preset design pressure, and then it is possible to more accurately evaluate whether the test results of the sealing test performed on the primary circuit system in each refueling cycle can be used as an alternative test result of the primary circuit water pressure test, thereby improving the accuracy of the evaluation results.

[0073] In one embodiment, a computer device can obtain a final safety analysis report pre-generated for a nuclear power plant, determine from the final safety analysis report a set of accidents that cause changes in the pressure of a primary-loop system of a nuclear power unit in the nuclear power plant, determine the pressure of the primary-loop system caused by each accident in the accident set, and obtain a preset design pressure. Based on the preset design pressure and the pressure of the primary-loop system caused by each accident, select from the accident set candidate accidents that cause the pressure of the primary-loop system to exceed the preset design pressure as target accidents.

[0074] In one embodiment, step 106 includes: determining multiple sample nuclear power groups recorded in the preset nuclear power group historical event information that matches the type of the nuclear power group; for each sample nuclear power group, determining from the preset nuclear power group historical event information a target historical event that occurred within multiple preset historical statistical time periods for the sample nuclear power group and matches the target initiating event class; determining a reference occurrence frequency of the target initiating event class corresponding to the sample nuclear power group based on multiple preset historical statistical time periods and the number of events of the target historical events that occurred within multiple preset historical statistical time periods for the sample nuclear power group; and determining an occurrence frequency of the target initiating event class corresponding to the nuclear power group based on the respective reference occurrence frequencies of multiple sample nuclear power groups.

[0075] The preset historical statistical period is a pre-set historical statistical time period. Multiple preset historical statistical periods are statistical time periods with the same preset time length. The preset historical statistical period can be determined based on the time span recorded by the preset nuclear power unit historical event information. For example, the preset nuclear power unit historical event information may record events occurring within three years from 2020 to 2022. The preset time length may be one year, and the multiple historical statistical periods may be 2020, 2021, and 2022, respectively.

[0076] The target historical event is a historical event that matches the target initiating event class within the preset nuclear power plant historical event information. The occurrence frequency of the target initiating event class corresponding to a nuclear power plant can be the average of the reference occurrence frequencies of multiple sample nuclear power plants, or the minimum or maximum of the reference occurrence frequencies of multiple sample nuclear power plants.

[0077] In this embodiment, by determining the reference occurrence frequency of the target initiating event class corresponding to multiple sample nuclear power units that match the type of nuclear power unit, and then determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit based on the reference occurrence frequency, a more accurate occurrence frequency can be obtained to a certain extent, creating conditions for subsequently accurately obtaining the evaluation results.

[0078] In one embodiment, the computer device may calculate the ratio between the number of target historical events occurring in each preset historical period and the preset duration for the sample nuclear power plant, obtain the ratios corresponding to multiple preset historical periods, and determine a reference occurrence frequency for the sample nuclear power plant based on the ratios corresponding to the multiple preset historical periods. The reference occurrence frequency for the sample nuclear power plant may be the average of the ratios corresponding to the multiple preset historical periods, or a weighted average of the ratios corresponding to the multiple preset historical periods, with the ratios corresponding to the preset historical periods closer to the current time being assigned a greater weight.

[0079] In one embodiment, a preset pipeline degradation mechanism determination criterion table includes multiple degradation mechanism types and determination criteria corresponding to each of the multiple degradation mechanism types; step 108 includes: determining, for the pipelines in the primary loop system and the support system connected to the primary loop system, whether a determination criterion is met for the targeted pipeline according to each determination criterion in the preset pipeline degradation mechanism determination criterion table; when it is determined that a determination criterion is met for the targeted pipeline, determining the degradation mechanism type corresponding to the existing determination criterion as the corresponding degradation mechanism determination result for the targeted pipeline; when it is determined that no determination criterion is met for the targeted pipeline, determining a degradation mechanism determination result corresponding to the targeted pipeline that indicates the absence of a degradation mechanism; and determining a rupture possibility level corresponding to the pipelines in the primary loop system and the support system according to the preset pipeline rupture possibility level table and the degradation mechanism results corresponding to the pipelines.

[0080] The preset pipeline degradation mechanism judgment criteria table may include the correspondence between degradation mechanism types, judgment criteria, and areas in the pipeline where degradation mechanisms are prone to occur. The preset pipeline degradation mechanism judgment criteria table may be shown in Table 1 below.

[0081] Table 1 Preset criteria for determining pipeline degradation mechanisms

[0082]

[0083]

[0084] Among them, in the degradation mechanism type column of Table 1, the degradation mechanism type TF (Thermal Fatigue) includes two degradation mechanism types: TASCS (Thermal Stratification Cycling and Striping) and TT (Thermal Transients).

[0085] The degradation mechanism type of SCC (Stress Corrosion Cracking) includes four types of degradation mechanisms: IGSCC (Intergranular Stress Corrosion Cracking), TGSCC (Transgranular Stress Corrosion Cracking), ECSCC (External Chloride Stress Corrosion Cracking), and PWSCC (Primary Water Stress Corrosion Cracking).

[0086] The degradation mechanism type of LC (Localized Corrosion) includes three types of degradation mechanisms: MIC (Microbiologically Influenced Corrosion), PIT (Pitting Corrosion), and CC (Crevice Corrosion).

[0087] The FS (Flow Sensitive) degradation mechanism type includes two types of degradation mechanisms: EC (Erosion-Cavitation) and FAC (Flow-Accelerated Corrosion).

[0088] The preset pipeline rupture probability level table can record the correspondence between the preset rupture probability levels and the preset degradation mechanism determination results, and can also record the expected leakage conditions of the pipeline and the rupture frequency of the pipeline. The preset pipeline rupture probability level table can be shown in Table 2 below.

[0089] Table 2 Preset pipeline rupture probability level table

[0090]

[0091] In this embodiment, each judgment criterion in the preset pipeline degradation mechanism judgment criterion table can determine whether a specific degradation mechanism type exists in each pipeline in the primary circuit system and each pipeline in the support system. Furthermore, combined with the preset pipeline rupture possibility level table, the rupture possibility level of the pipeline can be quickly determined, thereby creating conditions for subsequent determination of the assessment results.

[0092] In one embodiment, the computer device may determine, for pipelines in a primary-loop system and a support system connected to the primary-loop system, whether a determination criterion is met for the targeted pipeline according to each determination criterion in a preset pipeline degradation mechanism determination criterion table as shown in Table 1; if a determination criterion is met for the targeted pipeline, the degradation mechanism type corresponding to the existing determination criterion is determined as the corresponding degradation mechanism determination result for the targeted pipeline; if it is determined that no determination criterion is met for the targeted pipeline, the degradation mechanism determination result corresponding to the targeted pipeline indicating the absence of a degradation mechanism is determined; and the preset rupture possibility level corresponding to the degradation mechanism determination result corresponding to the targeted pipeline in the preset pipeline rupture possibility level table as shown in Table 2 is used as the rupture possibility level corresponding to the targeted pipeline.

[0093] In one embodiment, step 110 includes: determining the occurrence frequency of the breach initiating event class in the initiating event class set based on the rupture possibility level corresponding to the pipeline; determining the index value of the core damage index and the index value of the radioactive release index according to the occurrence frequency of the breach initiating event class, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model; and determining the evaluation result according to the index value of the core damage index and the index value of the radioactive release index.

[0094] The core damage indicator values ​​may include the core damage frequency (CDF) and the core damage frequency increment. The radioactive release indicator values ​​include the large early radioactivity release frequency (LERF) and the large early radioactivity release frequency increment.

[0095] The breach-initiating event class is an event class for the loss of coolant in a primary circuit system due to a pipeline rupture. This class can be referred to as the LOCA (Loss of Charged Air) initiating event class or a LOCA accident. The breach-initiating event class can overlap with the target-initiating event class; that is, the breach-initiating event class can be a subset of the target-initiating event class, or the breach-initiating event class and the target-initiating event class can be the same. Alternatively, the breach-initiating event class and the target-initiating event class can be different.

[0096] In this embodiment, the frequency of occurrence of the rupture initiating event class is determined by the rupture possibility level corresponding to the pipeline, and then the index value of the core damage index and the index value of the radioactive release index can be determined based on the occurrence frequency of the rupture initiating event class, the occurrence frequency of the target initiating event class and the probabilistic safety analysis model. The evaluation results are determined based on the quantified index values, which can improve the evaluation accuracy.

[0097] In one embodiment, a computer device may determine, for a breach initiating event class in a set of initiating event classes, the pipelines in the primary system and the support system that are associated with the targeted breach initiating event class, and determine the frequency of occurrence of the targeted breach initiating event class based on the corresponding rupture probability levels of the associated pipelines. The breach initiating event class may be an event class for a specific system, and the pipelines associated with the breach initiating event class may be pipelines in the specific system targeted by the breach initiating event class. For example, if the breach initiating event class is a water supply pipeline rupture accident, and the breach initiating event class is a breach initiating event class for the main water supply system of the secondary system in the support system, then the pipelines associated with the breach initiating event class may be pipelines in the main water supply system.

[0098] In one embodiment, a computer device may determine a rupture probability value corresponding to a rupture probability level for each pipeline in the relevant pipelines, and determine the occurrence frequency of a targeted rupture initiating event class based on the rupture probability values ​​corresponding to each of the relevant pipelines. When the rupture probability levels represent high, medium, and low rupture probability, respectively, the corresponding rupture probability values ​​may be a first preset probability value, a second preset probability value, and a third preset probability value, respectively. The first preset probability value is greater than the second preset probability value, and the second preset probability value is greater than the third preset probability value. Specifically, the first preset probability value may be 50%, the second preset probability value may be 20%, and the third preset probability value may be 0. The occurrence frequency of a targeted rupture initiating event class may be the average of the rupture probability values ​​corresponding to the relevant pipelines, or the maximum value among the rupture probability values ​​corresponding to the relevant pipelines.

[0099] In one embodiment, the probabilistic safety analysis model can be configured with frequency data bits for each initiating event class in the set of initiating event classes. When the breach initiating event class and the target initiating event class do not overlap, the computer device inputs the occurrence frequency of the breach initiating event class and the occurrence frequency of the target initiating event class into the corresponding frequency data bits in the probabilistic safety analysis model to obtain the index values ​​of the core damage indicator and the radioactive release indicator output by the probabilistic safety analysis model. The frequency data bits are used to input the occurrence frequency.

[0100] In one embodiment, when there is an overlapping initiating event class between the breach initiating event class and the target initiating event class, the computer device can determine the first occurrence frequency when the overlapping initiating event class is used as the initiating event class in the breach initiating event class, and determine the second occurrence frequency when the overlapping initiating event class is used as the initiating event class in the target initiating event class, and determine the target occurrence frequency of the overlapping initiating event class based on the first occurrence frequency and the second occurrence frequency; the target occurrence frequency of the overlapping initiating event class, the occurrence frequency of the initiating event class other than the overlapping initiating event class in the breach initiating event class, and the occurrence frequency of the initiating event class other than the overlapping initiating event class in the target initiating event class are input into the corresponding frequency data bits in the probabilistic safety analysis model to obtain the index value of the core damage index and the index value of the radioactive release index output by the probabilistic safety analysis model.

[0101] In one embodiment, the step of determining the evaluation result according to the index value of the core damage index and the index value of the radioactive release index includes: determining a first risk level under the core damage index dimension according to the index value of the core damage index; determining a second risk level under the radioactive release index dimension according to the index value of the radioactive release index; when the first risk level is a medium risk level or a low risk level, and the second risk level is a medium risk level or a low risk level, obtaining an evaluation result characterizing that a test result of a sealing test can be performed on the primary circuit system in each refueling cycle as an alternative test result of the primary circuit water pressure test.

[0102] The first risk level is the risk level under the core damage indicator dimension. It is determined based on the core damage frequency and the core damage frequency increment. The second risk level is the risk level under the radioactive release indicator dimension. It is determined based on the early large-scale radioactive release frequency and the early large-scale radioactive release frequency increment. The first or second risk level can represent a high, medium, or low risk level.

[0103] The core damage frequency increment indicator is the difference between the core damage frequency calculated based on the probabilistic safety analysis model and the pre-recorded core damage frequency value for the primary hydrostatic test. The early large radioactive release frequency increment indicator is the difference between the early large radioactive release frequency calculated based on the probabilistic safety analysis model and the pre-recorded early large radioactive release frequency value for the primary hydrostatic test.

[0104] In this embodiment, the risk levels in two dimensions are determined based on the quantitative index values ​​of the core damage index and the radioactive release index. The risks of the test results of the sealing test on the primary circuit system in each refueling cycle as an alternative test result to the primary circuit water pressure test can be evaluated in multiple dimensions, and the evaluation results can be determined more accurately.

[0105] In one embodiment, based on the risk-guided decision-making method for applying for a change in the Licensing Basis (LB) for a specific nuclear power plant in the NNSA-014 standard, a first risk level and a second risk level may be determined. Specifically, Figure 2 As shown in the risk assessment guideline diagram under the core damage indicator dimension, when the core damage frequency is less than 10 -3 , and the core damage frequency increment is less than 10 -6 , the computer equipment can obtain the first risk level (region III) representing the low risk level under the core damage index dimension; when the core damage frequency is less than 10 -4 , and the core damage frequency increment is less than 10 -5 and not less than 10 -6 , the computer device can obtain the first risk level (region II) representing the medium risk level under the dimension representing the core damage index.

[0106] In one embodiment, Figure 3 As shown in the risk assessment guideline diagram under the radioactive release indicator dimension, when the frequency of early large-scale radioactive release is less than 10 -4 , and the frequency increment of early large-scale radioactive release is less than 10 -7 , the computer equipment can obtain the second risk level (region III) representing the low risk level under the dimension of radioactive release index; when the frequency of early large-scale radioactive release is less than 10 -5 , and the frequency increment of early large-scale radioactive release is less than 10 -6 and not less than 10 -7 , the computer equipment can obtain the second risk level (region II) representing the medium risk level under the radioactive release index dimension.

[0107] In one embodiment, when the first risk level or the second risk level represents a high risk level, an evaluation result is obtained indicating that the test result of the sealing test performed on the primary circuit system during each refueling cycle cannot be used as a substitute test result for the primary circuit water pressure test. -5 , the first risk level represents a high risk level; when the frequency increment of early large-scale radioactive release is not less than 10 -6 , then the second risk level represents a high risk level.

[0108] In a specific embodiment, Figure 4 The flowchart of the change evaluation steps for the primary circuit water pressure test is shown. Before the change strategy of using the experimental results of other tests as the substitute test results of the primary circuit water pressure test is expected, the risk of the change strategy needs to be evaluated. Specifically, the different parts of the objects tested by the other tests and the primary circuit water pressure test can be compared. When the objects tested by each test are the same, it is determined that the change strategy does not affect the safety of the nuclear power unit, and the experimental results of other tests can be used as substitute test results for the primary circuit water pressure test. When there are different parts of the objects tested by each test, it is determined that the change strategy may affect the safety of the nuclear power unit. It is necessary to obtain an evaluation result through the above-mentioned data processing method for the primary circuit water pressure test of the nuclear power plant to determine whether the experimental results of other tests can be used as substitute test results for the primary circuit water pressure test. For example, other tests can be the sealing test of the primary circuit system in each refueling cycle. Compared with the primary circuit water pressure test, the sealing test conducted in each refueling cycle cannot test the sealing condition when the pressure of the primary circuit system exceeds 1 times the preset design pressure, the connecting pipes at the connection between the primary circuit system and the support system, and the valves connecting the connecting pipes and the support system. Based on Figure 4 The above-mentioned data processing method for the primary circuit water pressure test of a nuclear power plant specifically includes the following steps.

[0109] The computer equipment can obtain a final safety analysis report pre-generated for the nuclear power plant; determine, from the final safety analysis report, a set of accidents that cause a change in the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant; determine, from the accident set, a target accident that causes the pressure of the primary circuit system to exceed a preset design pressure; and determine, from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model, a target initiating event class that matches the target accident.

[0110] The computer device can determine multiple sample nuclear power groups recorded in the preset nuclear power group historical event information that matches the type of the nuclear power group; for each sample nuclear power group, determine from the preset nuclear power group historical event information the target historical events that occurred in multiple preset historical statistical time periods for the sample nuclear power group and matched the target initiating event class; determine the reference occurrence frequency of the target initiating event class corresponding to the sample nuclear power group based on the multiple preset historical statistical time periods and the number of events of the target historical events that occurred in the multiple preset historical statistical time periods for the sample nuclear power group; determine the occurrence frequency of the target initiating event class corresponding to the nuclear power group based on the respective reference occurrence frequencies of the multiple sample nuclear power groups.

[0111] The computer device can determine whether there is a determination criterion that is met for the pipeline in the primary loop system and the support system connected to the primary loop system according to each determination criterion in the preset pipeline degradation mechanism determination criterion table; when it is determined that there is a determination criterion that is met for the pipeline, the degradation mechanism type corresponding to the existing determination criterion is determined as the corresponding degradation mechanism determination result for the pipeline; when it is determined that there is no determination criterion that is met for the pipeline, the degradation mechanism determination result corresponding to the pipeline that indicates the absence of the degradation mechanism is determined; and according to the preset pipeline rupture possibility level table and the degradation mechanism results corresponding to the pipeline, the rupture possibility level corresponding to the pipeline in the primary loop system and the support system is determined.

[0112] The computer equipment can determine the frequency of occurrence of the rupture initiating event class in the set of initiating event classes based on the rupture possibility level corresponding to the pipeline; determine the index value of the core damage index and the index value of the radioactive release index according to the frequency of occurrence of the rupture initiating event class, the frequency of occurrence of the target initiating event class and the probabilistic safety analysis model; determine the first risk level under the core damage index dimension according to the index value of the core damage index; determine the second risk level under the radioactive release index dimension according to the index value of the radioactive release index; when the first risk level represents a medium risk level or a low risk level, and the second risk level represents a medium risk level or a low risk level, determine that the risk of the change strategy is acceptable, and obtain an evaluation result that represents the test result of the sealing test that can be performed on the primary circuit system in each refueling cycle as an alternative test result of the primary circuit water pressure test.

[0113] It should be understood that, although the various steps in the flowcharts involved in the various embodiments described above are displayed in sequence according to the instructions of the arrows, these steps are not necessarily executed in sequence in the order indicated by the arrows. Unless otherwise specified herein, there is no strict order restriction on the execution of these steps, and these steps can be executed in other orders. Moreover, at least a portion of the steps in the flowcharts involved in the various embodiments described above can include multiple steps or multiple stages, and these steps or stages are not necessarily executed and completed at the same time, but can be executed at different times, and the execution order of these steps or stages is not necessarily to be carried out in sequence, but can be executed in turn or alternately with other steps or at least a portion of steps or stages in other steps.

[0114] Based on the same inventive concept, embodiments of the present application also provide a data processing device for a nuclear power plant primary circuit hydraulic test, which is used to implement the aforementioned data processing method for a nuclear power plant primary circuit hydraulic test. The solution provided by this device is similar to the solution described in the aforementioned method. Therefore, the specific limitations of one or more embodiments of the data processing device for a nuclear power plant primary circuit hydraulic test provided below can be found in the aforementioned limitations of the data processing method for a nuclear power plant primary circuit hydraulic test, and will not be further elaborated here.

[0115] In one embodiment, Figure 5 As shown, a data processing device 500 for a primary circuit hydrostatic test of a nuclear power plant is provided, comprising: an originating event management module 510, a pipeline information management module 520, and an evaluation module 530, wherein:

[0116] The initiating event class management module 510 is used to determine the target accident that causes the pressure of the primary circuit system of the nuclear power unit in the nuclear power plant to exceed the preset design pressure based on the final safety analysis report pre-generated for the nuclear power plant; determine the target initiating event class that matches the target accident from the initiating event class set configured for the pre-generated probabilistic safety analysis model; and determine the occurrence frequency of the target initiating event class corresponding to the nuclear power unit based on the preset nuclear power unit historical event information that matches the type of the nuclear power unit.

[0117] The pipeline information management module 520 is used to determine the corresponding rupture possibility level of the pipelines in the primary loop system and the support system connected to the primary loop system according to the preset pipeline degradation mechanism judgment criterion table and the preset pipeline rupture possibility level table.

[0118] Evaluation module 530 is used to determine the evaluation result based on the corresponding rupture probability level of the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the evaluation result indicates whether the test results of the sealing test performed on the primary circuit system during each refueling cycle can be used as an alternative test result for the primary circuit water pressure test.

[0119] In one embodiment, the initiating event class management module 500 is also used to obtain a final safety analysis report pre-generated for the nuclear power plant; from the final safety analysis report, determine a set of accidents that cause changes in the pressure of the primary loop system of the nuclear power unit in the nuclear power plant; and from the accident set, determine a target accident that causes the pressure of the primary loop system to exceed the preset design pressure.

[0120] In one embodiment, the initiating event class management module 500 is also used to determine multiple sample nuclear power groups recorded in the preset nuclear power group historical event information that matches the type of the nuclear power group; for each sample nuclear power group, determine the target historical events that occurred in multiple preset historical statistical time periods for the sample nuclear power group and match the target initiating event class from the preset nuclear power group historical event information; determine the reference occurrence frequency of the target initiating event class corresponding to the sample nuclear power group based on the multiple preset historical statistical time periods and the number of events of the target historical events that occurred in the multiple preset historical statistical time periods for the sample nuclear power group; determine the occurrence frequency of the target initiating event class corresponding to the nuclear power group based on the reference occurrence frequencies of each of the multiple sample nuclear power groups.

[0121] In one embodiment, the preset pipeline degradation mechanism determination criteria table includes multiple degradation mechanism types and corresponding determination criteria for each of the multiple degradation mechanism types. The pipeline information management module 520 is further configured to determine, for the pipelines in the primary loop system and the support system connected to the primary loop system, whether a determination criterion exists for the pipeline in question based on each determination criterion in the preset pipeline degradation mechanism determination criteria table; if it is determined that a determination criterion exists for the pipeline in question, determine the degradation mechanism type corresponding to the existing determination criterion as the corresponding degradation mechanism determination result for the pipeline in question; if it is determined that no determination criterion exists for the pipeline in question, determine a degradation mechanism determination result corresponding to the pipeline in question that indicates the absence of a degradation mechanism; and determine the rupture probability level for the pipeline in the primary loop system and the support system based on the preset pipeline rupture probability level table and the corresponding degradation mechanism results for the pipeline.

[0122] In one embodiment, the evaluation module 530 is also used to determine the occurrence frequency of the rupture initiating event class in the initiating event class set based on the rupture possibility level corresponding to the pipeline; determine the index value of the core damage index and the index value of the radioactive release index according to the occurrence frequency of the rupture initiating event class, the occurrence frequency of the target initiating event class and the probabilistic safety analysis model; and determine the evaluation result according to the index value of the core damage index and the index value of the radioactive release index.

[0123] In one embodiment, the evaluation module 530 is also used to determine a first risk level under the core damage index dimension based on the index value of the core damage index; determine a second risk level under the radioactive release index dimension based on the index value of the radioactive release index; when the first risk level represents a medium risk level or a low risk level, and the second risk level represents a medium risk level or a low risk level, obtain an evaluation result that represents the test result of the sealing test that can be performed on the primary circuit system in each refueling cycle as an alternative test result of the primary circuit water pressure test.

[0124] Each module in the aforementioned data processing device for nuclear power plant primary circuit hydrostatic testing may be implemented in whole or in part through software, hardware, or a combination thereof. Each module may be embedded in or independent of a processor within a computer device in hardware form, or may be stored in a computer device's memory in software form, allowing the processor to call and execute the corresponding operations of each module.

[0125] In one embodiment, a computer device is provided. The computer device may be a terminal, and its internal structure diagram may be as follows: Figure 6 As shown. The computer device includes a processor, a memory, an input / output interface, a communication interface, a display unit, and an input device. The processor, memory, and input / output interface are connected via a system bus, and the communication interface, display unit, and input device are connected to the system bus via the input / output interface. The processor of the computer device is used to provide computing and control capabilities. The memory of the computer device includes a non-volatile storage medium and an internal memory. The non-volatile storage medium stores an operating system and a computer program. The internal memory provides an environment for the operation of the operating system and computer program in the non-volatile storage medium. The input / output interface of the computer device is used to exchange information between the processor and an external device. The communication interface of the computer device is used to communicate with an external terminal in a wired or wireless manner, and the wireless manner can be achieved through WIFI, a mobile cellular network, NFC (near field communication) or other technologies. When the computer program is executed by the processor, a data processing method for a primary circuit water pressure test of a nuclear power plant is implemented. The display unit of the computer device is used to form a visually visible image, and can be a display screen, a projection device, or a virtual reality imaging device. The display screen can be a liquid crystal display screen or an electronic ink display screen, and the input device of the computer device can be a touch layer covering the display screen, or a button, trackball or touchpad set on the computer device casing, or an external keyboard, touchpad or mouse.

[0126] Those skilled in the art will understand that Figure 6 The structure shown in the figure is only a block diagram of a part of the structure related to the solution of the present application, and does not constitute a limitation on the computer device to which the solution of the present application is applied. The specific computer device may include more or fewer components than shown in the figure, or combine certain components, or have a different component arrangement.

[0127] In one embodiment, a computer device is further provided, including a memory and a processor. The memory stores a computer program, and the processor implements the steps in the above method embodiments when executing the computer program.

[0128] In one embodiment, a computer-readable storage medium is provided, on which a computer program is stored. When the computer program is executed by a processor, the steps in the above-mentioned method embodiments are implemented.

[0129] In one embodiment, a computer program product is provided, including a computer program, which implements the steps in the above method embodiments when executed by a processor.

[0130] It should be noted that the user information (including but not limited to user device information, user personal information, etc.) and data (including but not limited to data used for analysis, stored data, displayed data, etc.) involved in this application are all information and data authorized by the user or fully authorized by all parties, and the collection, use and processing of relevant data must comply with relevant regulations.

[0131] Those skilled in the art will appreciate that all or part of the processes in the above-mentioned embodiment methods can be implemented by instructing the relevant hardware through a computer program, and the computer program can be stored in a non-volatile computer-readable storage medium. When the computer program is executed, it can include the processes of the embodiments of the above-mentioned methods. Among them, any reference to memory, database or other media used in the embodiments provided in this application may include at least one of non-volatile and volatile memory. Non-volatile memory may include read-only memory (ROM), magnetic tape, floppy disk, flash memory, optical memory, high-density embedded non-volatile memory, resistive random access memory (ReRAM), magnetic random access memory (MRAM), ferroelectric random access memory (FRAM), phase change memory (PCM), graphene memory, etc. Volatile memory may include random access memory (RAM) or external cache memory, etc. By way of illustration and not limitation, RAM can be in various forms, such as static random access memory (SRAM) or dynamic random access memory (DRAM). The database involved in the various embodiments provided herein may include at least one of a relational database and a non-relational database. Non-relational databases may include, but are not limited to, distributed databases based on blockchains. The processor involved in the various embodiments provided herein may be, but are not limited to, a general-purpose processor, a central processing unit, a graphics processing unit, a digital signal processor, a programmable logic unit, a data processing logic unit based on quantum computing, and the like.

[0132] The technical features of the above embodiments can be combined arbitrarily. To make the description concise, not all possible combinations of the technical features in the above embodiments are described. However, as long as there is no contradiction in the combination of these technical features, they should be considered to be within the scope of this specification.

[0133] The above-described embodiments merely represent several implementation methods of the present application. While the descriptions are relatively specific and detailed, they should not be construed as limiting the scope of the present application. It should be noted that a person of ordinary skill in the art may make various modifications and improvements without departing from the spirit of the present application, and these modifications and improvements fall within the scope of protection of the present application. Therefore, the scope of protection of the present application shall be determined by the appended claims.

Claims

1. A data processing method for a primary circuit hydraulic test of a nuclear power plant, characterized in that: The method comprises: Determining, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure; Determining a target initiating event class that matches the target accident from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model; Determining the occurrence frequency of the target initiating event class corresponding to the nuclear power unit according to preset nuclear power unit historical event information that matches the type of the nuclear power unit; Determining the rupture possibility levels corresponding to the pipelines in the primary loop system and the support system connected to the primary loop system according to a preset pipeline degradation mechanism judgment criterion table and a preset pipeline rupture possibility level table; An assessment result is determined based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the assessment result indicates whether the test result of the sealing test performed on the primary circuit system during each material change cycle can be used as an alternative test result for the primary circuit water pressure test.

2. The method according to claim 1, characterized in that The target accident that causes the pressure of the primary circuit system of the nuclear power unit in the nuclear power plant to exceed the preset design pressure is determined based on the final safety analysis report pre-generated for the nuclear power plant, including: Obtain pre-generated final safety analysis reports for nuclear power plants; determining, from the final safety analysis report, a set of accidents that cause a change in the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant; A target accident causing the pressure of the primary circuit system to exceed a preset design pressure is determined from the accident set.

3. The method according to claim 1, characterized in that The determining, based on preset nuclear power unit historical event information that matches the type of the nuclear power unit, the occurrence frequency of the target initiating event class corresponding to the nuclear power unit includes: Determining a plurality of sample nuclear power units recorded in preset nuclear power unit historical event information that match the type of the nuclear power unit; For each sample nuclear power unit, determining, from preset nuclear power unit historical event information, target historical events that occurred within multiple preset historical statistical time periods for the sample nuclear power unit and matched the target initiating event class; Determining a reference occurrence frequency of the target initiating event class corresponding to the sample nuclear power unit according to the multiple preset historical statistical time periods and the number of events of the target historical events occurring in the sample nuclear power unit within the multiple preset historical statistical time periods; The occurrence frequency of the target initiating event class corresponding to the nuclear power group is determined based on the reference occurrence frequencies of each of the multiple sample nuclear power groups.

4. The method according to claim 1, wherein The preset pipeline degradation mechanism judgment criterion table includes a plurality of degradation mechanism types and judgment criteria corresponding to the plurality of degradation mechanism types; determining the corresponding rupture possibility levels of pipelines in the primary loop system and the support system connected to the primary loop system according to the preset pipeline degradation mechanism judgment criterion table and the preset pipeline rupture possibility level table includes: For the pipelines in the primary loop system and the support system connected to the primary loop system, determine whether the pipeline meets the determination criteria according to each determination criteria in the preset pipeline degradation mechanism determination criteria table; When it is determined that the targeted pipeline meets the determination criteria, the degradation mechanism type corresponding to the existing determination criteria is determined as the corresponding degradation mechanism determination result of the targeted pipeline; When it is determined that the targeted pipeline does not meet the determination criteria, a degradation mechanism determination result corresponding to the targeted pipeline indicating that no degradation mechanism exists is determined; The rupture possibility levels of the pipelines in the primary system and the support system are determined according to a preset pipeline rupture possibility level table and the degradation mechanism results corresponding to the pipelines.

5. The method according to claim 1, wherein Determining an assessment result based on the rupture possibility level corresponding to the pipeline, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model includes: Determining the occurrence frequency of a rupture initiating event class in the set of initiating event classes based on the rupture possibility level corresponding to the pipeline; Determining an index value of a core damage index and an index value of a radioactive release index based on the occurrence frequency of a breach initiating event class, the occurrence frequency of the target initiating event class, and the probabilistic safety analysis model; An assessment result is determined according to the index value of the core damage index and the index value of the radioactive release index.

6. The method according to claim 5, characterized in that Determining the evaluation result according to the index value of the core damage index and the index value of the radioactive release index includes: determining a first risk level under the core damage indicator dimension according to an indicator value of the core damage indicator; determining a second risk level under the radioactive release indicator dimension according to the indicator value of the radioactive release indicator; When the first risk level represents a medium risk level or a low risk level, and the second risk level represents a medium risk level or a low risk level, an evaluation result is obtained that represents the test result of the sealing test that can be performed on the primary circuit system in each material change cycle as an alternative test result of the primary circuit water pressure test.

7. A data processing device for a primary circuit water pressure test of a nuclear power plant, characterized in that: The device comprises: An initiating event class management module is configured to determine, based on a final safety analysis report pre-generated for a nuclear power plant, a target accident that causes the pressure of a primary circuit system of a nuclear power unit in the nuclear power plant to exceed a preset design pressure; determine, from a set of initiating event classes configured for a pre-generated probabilistic safety analysis model, a target initiating event class that matches the target accident; and determine, based on preset nuclear power unit historical event information that matches the type of the nuclear power unit, an occurrence frequency of the target initiating event class corresponding to the nuclear power unit; a pipeline information management module, configured to determine, based on a preset pipeline degradation mechanism determination criterion table and a preset pipeline rupture probability level table, corresponding rupture probability levels of pipelines in the primary loop system and in a support system connected to the primary loop system; An evaluation module is configured to determine an evaluation result based on the rupture probability level corresponding to the pipeline, the frequency of occurrence of the target initiating event class, and the probabilistic safety analysis model; the evaluation result indicates whether the test result of the sealing test performed on the primary circuit system during each refueling cycle can be used as an alternative test result for the primary circuit water pressure test.

8. A computer device comprising a memory and a processor, wherein the memory stores a computer program, wherein: When the processor executes the computer program, the steps of the method according to any one of claims 1 to 6 are 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 steps of the method according to any one of claims 1 to 6 are implemented.

10. A computer program product comprising a computer program, characterized in that When the computer program is executed by a processor, the steps of the method according to any one of claims 1 to 6 are implemented.

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