Determination method and device based on earthquake PSA model aseismic design

Abstract

The invention discloses a method and device for determining aseismic design based on an earthquake PSA model, and the method comprises the following steps: firstly, building an earthquake PSA model for analyzing the aseismic design under a non-heap-melting design expansion working condition DEC-A; secondly, performing quantitative analysis on the nuclear power plant reactor core damage frequency caused by an earthquake by using an earthquake PSA model; judging whether the quantitative analysis result is within a preset safety target range or not; if the result of the quantitative analysis is not in the preset safety target range, performing iterative analysis on the aseismic design by using an earthquake PSA model; and finally, according to an iterative analysis result and engineering factors, determining an aseismic design coping with the DEC-A working condition. According to the method and the device, a basis can be provided for determination of aseismic design, and maximization of economic benefits can be realized while the safety of a power plant is not reduced.

Description

Technical Field

[0001] The present invention relates to the technical field of nuclear power seismic design, and in particular to a method and device for determining seismic design based on an earthquake PSA model. Background Art

[0002] An important change in the new nuclear safety requirements is the application of the concept of Design Extension Conditions (DEC). DEC takes into account part of the previous Beyond Design Basis Accident (BDBA) and proposes corresponding design requirements and acceptance criteria for safety measures to deal with DEC. According to the different levels of defense in depth, the design extension conditions are divided into design extension conditions (DEC-A) that do not cause obvious damage to the core, and design extension conditions (DEC-B) with core melting; the corresponding response measures for these conditions are DEC-A measures (severe accident prevention measures) and DEC-B measures (severe accident mitigation measures). The seismic design code for nuclear power plants GB50267-2019 integrates all contents related to the seismic design of nuclear power plants, and coordinates with other seismic codes. It divides seismic resistance into seismic resistance and non-nuclear seismic resistance, and gives definitions and requirements for each seismic resistance type. However, these standards only give principled requirements and cannot be directly applied to the actual work of determining the seismic capacity requirements for DEC condition mitigation systems. As a result, the current seismic design of nuclear power plants' measures to cope with DEC conditions is still mainly based on the analyst's experience and subjective judgment, and lacks the support of consistent implementation steps and analysis methods.

[0003] The existing patent CN115542391A discloses an earthquake PSA vulnerability analysis method and system based on on-site visits to nuclear power plants. The method includes the following steps: obtaining the actual status of structures, systems and equipment and classifying them; assigning different vulnerability reduction factors according to the influence of different types of structures, systems and equipment on seismic performance; reflecting different vulnerability reduction factors in the earthquake PSA vulnerability analysis to obtain the final vulnerability analysis results.

[0004] The existing patent CN115907517A discloses a nuclear power plant earthquake PSA visit system and method based on structured data. The system consists of a mobile terminal and a server terminal. The mobile terminal and the server terminal interact with each other through a data transmission module. The server terminal is responsible for the summary and collation of the visit data, as well as the verification and data storage. The mobile terminal is responsible for data transmission and reading, system login and homepage, and information interaction between users and the server terminal. This solution is used to solve the problem of confusion and error caused by the need to compile a large number of files during the existing earthquake resistance visit.

[0005] In summary, the above two existing patents do not solve the problem that the seismic design of the current nuclear power plant's measures to cope with DEC conditions is mainly based on the analyst's experience and subjective judgment, and lacks a systematic analysis method. Summary of the invention

[0006] Based on the above technical problems, the present invention proposes a method and device for determining seismic design based on the earthquake PSA model, which solves the problem that the seismic design of nuclear power plants' DEC working conditions in the prior art is mainly based on the analyst's experience and subjective judgment, and lacks a systematic analysis method.

[0007] To achieve the above object, the present invention proposes a method for determining seismic design based on an earthquake PSA model.

[0008] A method for determining seismic design based on an earthquake PSA model, comprising:

[0009] Establish a seismic PSA model for analyzing the seismic design under the extended condition DEC-A without stacking melting;

[0010] The earthquake PSA model is used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes;

[0011] Determine whether the results of quantitative analysis are within the preset safety target range;

[0012] If the results of the quantitative analysis are not within the preset safety target range, the earthquake PSA model is used to perform iterative analysis on the seismic design;

[0013] Based on the iterative analysis results and engineering factors, the seismic design for DEC-A condition is determined.

[0014] Furthermore, a seismic PSA model is established to analyze the seismic design under the extended working condition DEC-A without stacking melting, including:

[0015] Identify all possible initiating events caused by earthquakes and obtain accident sequences that may lead to core damage in nuclear power plants based on all initiating events;

[0016] Establish an event tree of the mitigation system corresponding to the accident sequence;

[0017] Based on the earthquake failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model is established to analyze the failure causes of the mitigation system.

[0018] The earthquake PSA model is established based on all initiating events, event trees and fault tree models.

[0019] Furthermore, the accident sequences that may cause core damage in nuclear power plants are obtained based on all initiating events, including:

[0020] Establish an earthquake front tree model based on all initiating events;

[0021] The earthquake front-end tree model and the process analysis of all initiating events are used to determine the accident sequence that may lead to core damage in nuclear power plants.

[0022] Furthermore, the earthquake PSA model is used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes, including:

[0023] Obtain the occurrence frequency of initiating events and the frequency of earthquakes in different earthquake acceleration ranges;

[0024] Obtain the failure probability of mitigation equipment and personnel in the mitigation system;

[0025] Based on the failure probability of mitigation equipment, the failure probability of personnel, the frequency of initiating events, and the frequency of earthquakes in different earthquake acceleration ranges, Formula 1 is used to determine the frequency of superimposed failures of different equipment and human errors after the occurrence of initiating events in different earthquake acceleration ranges. Formula 1: Among them, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n It indicates the probability of failure of mitigation equipment or personnel;

[0026] The frequencies of different equipment failures and personnel failures after the initiating event in different earthquake acceleration ranges are added together to obtain quantitative analysis results.

[0027] Furthermore, the failure probability of the mitigation device in the mitigation system is obtained, including:

[0028] Determine mitigation equipment within the mitigation system based on the fault tree model;

[0029] Analyze the seismic vulnerability of the mitigation equipment and obtain the log-normal model parameters, which include the median seismic resistance, accidental uncertainty, and cognitive uncertainty;

[0030] According to the median seismic resistance capability, accidental uncertainty and cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2. Formula 2:

[0031]

[0032] Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β Ris the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

[0033] Furthermore, the seismic PSA model is used for iterative analysis, including:

[0034] Use the earthquake PSA model to identify weak links in seismic design and provide seismic design targeting the weak links;

[0035] Determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link;

[0036] If no seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link, further determine whether the seismic design is feasible;

[0037] If the seismic design is feasible, the earthquake PSA model is used again for quantitative analysis;

[0038] If earthquake-resistant design is not feasible, the equipment in the countermeasures of the weak links is set to be earthquake-resistant and the earthquake PSA model is updated.

[0039] Furthermore, if the earthquake PSA model is used again for quantitative analysis and the new quantitative analysis results still exceed the preset safety target range, the earthquake PSA model is used to re-identify the weak links in the seismic design, and the equipment in the response measures for the weak links is set to be earthquake-resistant.

[0040] Furthermore, the earthquake PSA model is used to identify weak links in the seismic design and provide seismic design for the weak links, including:

[0041] The earthquake PSA model is used to determine the minimum cut set of the combination of the initiating event caused by the earthquake and the failure of equipment and human error events in each earthquake acceleration range.

[0042] Analyze the minimum cut sets and determine the minimum cut sets in the minimum cut sets that are greater than or equal to the preset risk contribution;

[0043] Analyze the countermeasures of the minimum cut set that is greater than or equal to the preset risk contribution, and propose seismic design for the equipment in the countermeasures.

[0044] Furthermore, the earthquake PSA model is updated, including:

[0045] Update the event tree and fault tree of the mitigation system respectively;

[0046] The earthquake PSA model is updated using the updated event tree and fault tree.

[0047] Furthermore, it also includes:

[0048] Update the failure probability of mitigation devices within the mitigation system;

[0049] The updated failure probability of the mitigation equipment is input into the updated earthquake PSA model for quantitative analysis.

[0050] To achieve the above object, the present invention also proposes a determination device for seismic design based on the earthquake PSA model.

[0051] A determination device for seismic design based on an earthquake PSA model, comprising:

[0052] Establishing a module for establishing a seismic PSA model for analyzing seismic design under the extended condition DEC-A without stacking and melting;

[0053] Quantitative analysis module, used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes using the earthquake PSA model;

[0054] A judgment module is used to judge whether the result of quantitative analysis is within the preset safety target range;

[0055] Iterative analysis module, used to iteratively analyze seismic design using the earthquake PSA model if the result of quantitative analysis is not within the preset safety target range;

[0056] The design module is used to determine the seismic design for the DEC-A condition based on iterative analysis results and engineering factors.

[0057] Furthermore, a module is established for:

[0058] Identify all possible initiating events caused by earthquakes and obtain accident sequences that may lead to core damage in nuclear power plants based on all initiating events;

[0059] Establish an event tree of the mitigation system corresponding to the accident sequence;

[0060] Based on the earthquake failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model is established to analyze the failure causes of the mitigation system.

[0061] The earthquake PSA model is established based on all initiating events, event trees and fault tree models.

[0062] Furthermore, the accident sequences that may cause core damage in nuclear power plants are obtained based on all initiating events, including:

[0063] Establish an earthquake front tree model based on all initiating events;

[0064] The earthquake front-end tree model and the process analysis of all initiating events are used to determine the accident sequence that may lead to core damage in nuclear power plants.

[0065] Furthermore, the quantitative analysis module is used to:

[0066] Obtain the occurrence frequency of initiating events and the frequency of earthquakes in different earthquake acceleration ranges;

[0067] Obtaining the failure probability of mitigation equipment and the failure probability of personnel in the mitigation system;

[0068] Based on the failure probability of the mitigation equipment, the failure probability of the personnel, the occurrence frequency of the initiating event and the earthquake occurrence frequency in different earthquake acceleration intervals, Formula 1 is used to determine the frequency of superimposed failure of different equipment and personnel after the initiating event occurs in different earthquake acceleration intervals. Formula 1: Among them, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n To mitigate the probability of equipment failure or personnel failure;

[0069] The frequencies of different equipment failures and personnel failures after the initiating event in different earthquake acceleration ranges are added together to obtain quantitative analysis results.

[0070] Further, obtaining the failure probability of the mitigation device in the mitigation system includes:

[0071] Determine mitigation equipment within the mitigation system based on the fault tree model;

[0072] Analyze the seismic vulnerability of the mitigation equipment and obtain the log-normal model parameters, which include the median seismic resistance, accidental uncertainty, and cognitive uncertainty;

[0073] According to the median seismic resistance capability, accidental uncertainty and cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2. Formula 2:

[0074]

[0075] Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β R is the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

[0076] Furthermore, the iterative analysis module is used to:

[0077] Use the earthquake PSA model to identify weak links in seismic design and provide seismic design targeting the weak links;

[0078] Determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link;

[0079] If no seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link, further determine whether the seismic design is feasible;

[0080] If the seismic design is feasible, the earthquake PSA model is used again for quantitative analysis;

[0081] If earthquake-resistant design is not feasible, the equipment in the countermeasures of the weak links is set to be earthquake-resistant and the earthquake PSA model is updated.

[0082] Furthermore, if the earthquake PSA model is used again for quantitative analysis and the new quantitative analysis results still exceed the preset safety target range, the earthquake PSA model is used to re-identify the weak links in the seismic design, and the equipment in the response measures for the weak links is set to be earthquake-resistant.

[0083] Furthermore, the earthquake PSA model is used to identify weak links in the seismic design and provide seismic design for the weak links, including:

[0084] The earthquake PSA model is used to determine the minimum cut set of the combination of the initial event caused by the earthquake and the failure of the equipment and human error events under each earthquake acceleration range;

[0085] Analyze the minimum cut sets and determine the minimum cut sets in the minimum cut sets that are greater than or equal to the preset risk contribution;

[0086] Analyze the countermeasures of the minimum cut set that is greater than or equal to the preset risk contribution, and propose seismic design for the equipment in the countermeasures.

[0087] Furthermore, the earthquake PSA model is updated, including:

[0088] Update the event tree and fault tree of the mitigation system respectively;

[0089] The earthquake PSA model is updated using the updated event tree and fault tree.

[0090] Furthermore, it also includes:

[0091] Update the failure probability of mitigation devices within the mitigation system;

[0092] The updated failure probability of the mitigation equipment is input into the updated earthquake PSA model for quantitative analysis.

[0093] Based on the above technical solution, the present invention has at least the following beneficial effects:

[0094] 1. The present invention establishes an earthquake PSA model, utilizes the accident process analysis, earthquake vulnerability analysis and cut set analysis in the earthquake PSA model, and combines the characteristics of the non-melt design extended condition DEC-A to establish a method for determining the seismic design of the DEC-A condition response measures based on the earthquake PSA model. Based on the quantitative analysis results of the earthquake PSA model, the frequency of core damage caused by earthquakes in nuclear power plants is evaluated, providing a basis for determining the seismic design.

[0095] 2. The present invention utilizes the earthquake PSA model to iteratively evaluate the frequency of core damage caused by earthquakes in nuclear power plants, identifies the original earthquake-resistant equipment that needs to be considered in the weak links and the equipment in the mitigation measures for the weak links through cut set analysis, and finally combines the quantitative analysis results and engineering factors to comprehensively determine the earthquake-resistant design of the DEC-A countermeasures. This method and device can ensure that the economic benefits are maximized without reducing the safety of the power plant. BRIEF DESCRIPTION OF THE DRAWINGS

[0096] The accompanying drawings constituting a part of the present invention are used to provide a further understanding of the present invention. The exemplary embodiments of the present invention and their descriptions are used to explain the present invention and do not constitute an improper limitation of the present invention. In the accompanying drawings:

[0097] Figure 1 A flow chart of a method for determining seismic design based on an earthquake PSA model according to an embodiment of the present invention;

[0098] Figure 2 A flowchart of establishing an earthquake PSA model in one embodiment of the present invention;

[0099] Figure 3 A schematic diagram of an earthquake front-end tree established in one embodiment of the present invention;

[0100] Figure 4 A schematic diagram of a pre-tree for a large LOCA caused by an earthquake in a specific embodiment of the present invention;

[0101] Figure 5 This is a schematic diagram of an event tree of a large LOCA caused by an earthquake in a specific embodiment of the present invention;

[0102] Figure 6 A fault tree model corresponding to the isolation failure of a valve in a waste heat removal system in a specific embodiment of the present invention;

[0103] Figure 7 A flowchart for quantitatively analyzing the frequency of nuclear power plant core damage caused by earthquakes using the earthquake PSA model;

[0104] Figure 8 A schematic diagram of an earthquake vulnerability curve in one embodiment of the present invention;

[0105] Fig. 9 It is a schematic diagram of a device for determining seismic design based on an earthquake PSA model according to an embodiment of the present invention. DETAILED DESCRIPTION

[0106] It should be noted that, in the absence of conflict, the embodiments of the present invention and the features in the embodiments can be combined with each other. The present invention will be described in detail below with reference to the accompanying drawings and in combination with the embodiments.

[0107] The present invention is further described in detail below in conjunction with specific embodiments. These embodiments should not be construed as limiting the scope of protection claimed by the present invention.

[0108] Example

[0109] In order to solve the problem that the seismic design of nuclear power plants to cope with DEC conditions is mainly based on the analyst's experience and subjective judgment and lacks a systematic analysis method, the present invention proposes a method and device for determining seismic design based on an earthquake PSA model.

[0110] To achieve the above object, the present invention proposes a method for determining seismic design based on an earthquake PSA model.

[0111] like Figure 1 FIG. 4 is a flow chart of a method for determining seismic design based on an earthquake PSA model according to an embodiment of the present invention, and the flow chart includes the following steps:

[0112] S1. Establish a seismic PSA model for analyzing the seismic design under the extended condition DEC-A without stacking melting.

[0113] Probabilistic Safety Assessment (PSA) is a probabilistic safety analysis of earthquake hazards in nuclear power plants. It evaluates the overall risk of nuclear power plants under earthquakes of different intensities and identifies possible weaknesses in the seismic design of nuclear power plants, thereby providing comprehensive insights into earthquake risks for nuclear power plants.

[0114] The non-melt design extended condition DEC-A is also called the design extended condition without causing obvious core damage. Figure 2 As shown in FIG. 1 , a seismic PSA model for analyzing the seismic design under the design extension condition DEC-A without causing obvious core damage is established, including the following sub-steps:

[0115] S101, determining all possible initiating events caused by an earthquake, and obtaining accident sequences that may cause core damage to a nuclear power plant based on all the initiating events.

[0116] Table 1 Examples of possible initiating events caused by earthquakes

[0117] Serial number Earthquake-induced initiating events 1. Earthquake front tree causes direct core damage 2. Earthquake caused a major circuit rupture accident and a major LOCA 3. Earthquake caused a circuit rupture accident MLOCA 4. Earthquake caused a small circuit rupture accident (SLOCA) 5. Earthquake caused loss of external power supply LOOP accident 6. The earthquake caused the main steam pipe to rupture 7. Earthquake causes water pipe rupture 8. The earthquake caused the loss of all heat sinks 9. The earthquake caused the remaining drainage pipeline to rupture 10. The earthquake caused the loss of 11. The earthquake caused the loss of AC power in Train A 12. The earthquake caused the loss of AC power in train B 13. The earthquake caused the loss of DC power in Train A. 14. Earthquake causes loss of DC power in train B 15. Earthquake causes over-drainage of one circuit

[0118] As shown in Table 1, examples of possible initiating events caused by earthquakes are presented. In the present invention, the possible initiating events caused by earthquakes are considered in the earthquake front-end tree, and earthquake failures that do not cause initiating events are considered in the earthquake fault tree.

[0119] The process of obtaining the accident sequence that may cause core damage of a nuclear power plant according to all initiating events includes: S1011, establishing an earthquake front-end tree model according to all initiating events; S1012, using the earthquake front-end tree model and the process analysis of all initiating events to determine the accident sequence that may cause core damage of a nuclear power plant. Figure 3 , a schematic diagram of an earthquake front-end tree established in an embodiment of the present invention is shown in FIG. Sequences 2 to 21 are examples of establishing an earthquake front-end tree based on possible initiating events of a nuclear power plant caused by an earthquake. Since all SSCs of a nuclear power plant are affected by an earthquake at the same time, multiple initiating event types may occur simultaneously, so the superposition of various initiating events is also considered in the earthquake front-end tree model. For example, the accident sequence 11 corresponds to the simultaneous occurrence of the initiating events of the main steam pipe rupture SLB and the water supply pipe rupture FLB. After the earthquake front-end tree model is established, the earthquake front-end tree model is used to identify the damage state of the nuclear power plant after the earthquake. For the initiating event of catastrophic failure caused by an earthquake, it is conservatively considered that the initiating event directly causes core damage, such as the failure of auxiliary buildings, shielding buildings (auxiliary building roofs), reactor pressure vessels, etc.; for initiating events that do not cause catastrophic failure, their damage states continue to develop according to the response of the nuclear power plant under the earthquake, such as LOCA accidents and other transient accidents.

[0120] S102: Establish an event tree of the mitigation system corresponding to the accident sequence.

[0121] This embodiment determines the event tree of the mitigation system corresponding to the accident sequence by conducting accident process analysis in the early stage. Specifically, except for the sequence in which the earthquake directly causes core damage, the accident process mitigation logic of other initiating events refers to the event process mitigation logic in the internal event level 1 PSA model. Taking a three-loop pressurized water reactor nuclear power plant as an example, after an earthquake causes a large LOCA accident, it is assumed that there is a 1 / 3 probability that the breach will occur in the first-loop pipe section of loop 1, loop 2, and loop 3, such as Figure 4 The earthquake shown leads to a large LOCA pre-tree as shown; combined Figure 5It can be seen from the event tree of large LOCA caused by earthquake that after the breach of each loop, the mitigation process of the large LOCA event in the first-level PSA model of internal events includes: the injection tank provides early water replenishment, the low-pressure injection system injects through the cold section and the cold and hot sections at the same time, provides long-term water replenishment for the primary circuit, and uses the containment spray system (CSP) or the containment heat extraction system (PCS) as the final heat sink to remove the heat in the containment. Based on the above example, the event tree of multiple mitigation systems can be determined by referring to the event process mitigation logic in the first-level PSA model of internal events.

[0122] Different from the traditional earthquake PSA model, the earthquake PSA model in the present invention only considers the earthquake-resistant items determined by the design basis accident (DBA) measures, and the non-seismic SSC is considered to be bound to fail, that is, the functions performed by the non-safety system are set to a state of inevitable failure. Figure 5 The containment heat extraction system (PCS) used to remove heat from the containment is set to a state of inevitable failure, that is, the corresponding failure probability is 1.

[0123] S103, based on earthquake failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model for analyzing the failure causes of the mitigation system is established.

[0124] The earthquake PSA model should consider not only the earthquake failure of SSC, but also the random failure and human error events of SSC. When modeling the earthquake failure of SSC, on the basis of the first-level PSA model of internal events, the basic event of earthquake failure of SSC is connected with its random failure in the form of "OR gate" to reflect the impact of earthquake failure. Figure 6 The fault tree model corresponding to the isolation failure of a valve in the waste heat removal system is shown. Figure 6 The RHR030VP isolation failure is the top event, which will cause the electric valve RHR030VP power failure, electric valve RHR030VP refusal to close, and RHR system safety external isolation valve earthquake failure to be connected through the OR gate, where the electric valve RHR030VP power failure and electric valve RHR030VP refusal to close are random failure events of SSC. In the earthquake PSA model, the top event or related basic event of the non-safe system in the fault tree mitigation system is set to a state of inevitable failure. For example, the fault tree of the chilled water system of the electrical plant used to provide cooling under the loss of heat sink accident is set to a state of inevitable failure.

[0125] S104, establishing an earthquake PSA model based on all initiating events, event trees and fault tree models.

[0126] S2, using the earthquake PSA model to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes.

[0127] The purpose of this step is to complete the risk calculation of core damage caused by earthquakes based on the preliminarily established earthquake PSA model, according to the quantitative requirements of the risk assessment model for DEC-A measures and the setting of the quantitative calculation of the model. Different from the traditional earthquake PSA model, for some sequences that have developed to the DEC-A working condition, the present invention sets its consequence as core damage (CD), which is convenient for the subsequent quantitative calculation of the risk assessment model for DEC-A measures.

[0128] like Figure 7 As shown, the earthquake PSA model is used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes, including four sub-steps from S201 to S204:

[0129] S201, obtaining the occurrence frequency of the initiating event and the earthquake occurrence frequency in different earthquake acceleration ranges.

[0130] S202, obtaining the failure probability of mitigation equipment and the failure probability of personnel in the mitigation system.

[0131] Furthermore, obtaining the failure probability of the mitigation device in the mitigation system includes the following sub-steps:

[0132] S2011, Determine the mitigation equipment within the mitigation system based on the fault tree model.

[0133] like Figure 6 The mitigation equipment in the waste heat removal system includes: electric valve RHR030VP, 380V AC power supply, 110V DC power supply, and 48V DC power supply related equipment.

[0134] S2012, analyze the seismic vulnerability of the mitigation equipment and obtain the log-normal model parameters, which include the median seismic resistance, accidental uncertainty, and epistemic uncertainty.

[0135] The seismic vulnerability is defined as the conditional failure probability of the structures, systems and equipment (SSC) of a nuclear power plant under a given earthquake ground motion intensity. The result of the seismic vulnerability analysis in the seismic PSA is a cluster of seismic vulnerability curves. The log-normal model of vulnerability is given by three parameters: median capacity (A) m ), expressing the best estimate of earthquake resistance; logarithmic standard deviation β U and β R , expressing the degree of dispersion of the median value of seismic resistance, where β R represents randomness, aleatory uncertainty, β R It cannot be reduced, β UIt represents epistemic uncertainty, which can be reduced by deepening understanding and adopting more reasonable models.

[0136] S2013, based on the median seismic resistance capability, accidental uncertainty and cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2.

[0137] In the earthquake PSA model, it is assumed that the ground motion acceleration capacity A of the SSC obeys a double logarithmic normal distribution, which can be expressed as: A = A m e R e U , where A m represents the median value of earthquake resistance, e R and e U They represent the random uncertainty and cognitive uncertainty respectively, and are assumed to obey a log-normal distribution with a median of 1. Their logarithmic standard deviations are β R and β U When the ground motion acceleration a of an earthquake exceeds the ground motion acceleration capacity A of the SSC, the structure or equipment is considered to have failed. Therefore, for a given ground motion acceleration a, the failure probability of the structure or equipment at different confidence levels Q can be expressed as Formula 2:

[0138]

[0139] Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β R is the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

[0140] Based on the above, it can be seen that the seismic vulnerability of SSC is determined by the median value of seismic capacity A m , uncertainty logarithmic standard deviation β U is the logarithmic standard deviation of randomness β R Three parameters are determined. Among them, A m The selected ground motion parameter can be peak ground acceleration (PGA) or spectral acceleration, but it must be consistent with the parameters selected in the seismic hazard analysis. Usually, the peak ground acceleration is selected as the ground motion parameter in the seismic hazard analysis, so the peak ground acceleration is also mainly used as the parameter in the seismic vulnerability analysis. m =0.87g, β U =0.35, β R= 0.25 as an example, earthquake vulnerability curves with different confidence levels can be drawn as follows Figure 8 shown. Figure 8 The seismic vulnerability curves with confidence levels of 5%, 50% and 95% are given in the figure. Point A indicates that when the peak ground acceleration is 0.6g, the conditional failure probability of the mitigation device is 0.79, and the confidence level for this conditional failure probability is 95%, that is, with a confidence level of 95%, it can be considered that the conditional failure probability does not exceed 0.79. Figure 8 Point B in the figure can reflect the median value of earthquake resistance A. m The meaning of is the peak ground acceleration value corresponding to the conditional failure probability of 0.5 on the earthquake vulnerability curve with a confidence level of 50%. It reflects an optimal estimate of the seismic resistance of the equipment. Figure 8 Point C in the middle corresponds to the peak ground acceleration value corresponding to a 5% probability of conditional failure of the mitigation equipment on the earthquake vulnerability curve with a confidence level of 95%, which is called the "High Confidence of Low Probability of Failure" capability value (HCLPF) of the equipment.

[0141] S203, based on the failure probability of mitigation equipment, the failure probability of personnel, the frequency of occurrence of initiating events and the frequency of earthquakes in different earthquake acceleration ranges, the probability of superimposed failure of different equipment and human error after the occurrence of the initiating event in different earthquake acceleration ranges is determined using Formula 1.

[0142] Formula 1: Among them, F is the frequency of superimposed failure of different equipment and human error after the occurrence of the initiating event, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n To reduce the probability of equipment failure and human error.

[0143] S204, adding the frequencies of different equipment failures and personnel failures after the initiating event occurs in different earthquake acceleration ranges to obtain a quantitative analysis result.

[0144] Taking the proposal of some typical DEC-A measures of seismic design for a pressurized water reactor nuclear power plant as an example, firstly, by deleting and setting the modification method of logical state, the measures except the safety system for coping with design basis accidents are eliminated to establish the earthquake PSA model; then, the function headers involving non-safety systems are set to the state of inevitable failure: for example, the emergency boron injection function, the additional diesel engine access function, the secondary side passive residual heat removal system with heat function, the gravity water replenishment function, the water replenishment function through the mobile pump, etc. are set to the state of inevitable failure; the top events or basic events related to the fault tree involving non-safety systems are set to the state of inevitable failure: for example, the fault tree of the diversified drive system and the fault tree of the electrical plant chilled water system are set to the state of inevitable failure, that is, the corresponding failure probability is set to 1.

[0145] By analyzing each event sequence one by one, the consequence of the sequence that has developed to the DEC-A condition (i.e., the sequence of the initiating event superimposed on the failure of the DBA measure) is set to CD: for example, for the sequence that has developed to the loss of all heat sinks or the total power outage (SBO, Station Blackout) condition that may have considered mobile facilities or recovery measures in the PSA model, it is no longer necessary to continue to develop in the DEC-A measure evaluation model, and the corresponding sequence consequence is set to CD.

[0146] Quantitative calculations were performed using the Risk Spectrum software, and the total frequency CDF of the initiating event caused by an earthquake combined with failures of different equipment and human error was found to be 1.44E-06 / reactor-year.

[0147] S3, determining whether the result of the quantitative analysis is within the preset safety target range.

[0148] If the result of the quantitative analysis is within the preset safety target range, it is not necessary to perform iterative analysis, that is, it is not necessary to perform subsequent steps. In the above embodiment, the preset safety indicator is that the core damage frequency is not higher than 1E-06, that is, the result of the quantitative analysis is not within the preset safety target range, and step S4 is continued to perform iterative analysis on the seismic design of the nuclear power plant using the earthquake PSA model, and the PSA model and quantitative analysis are updated.

[0149] S4. If the result of the quantitative analysis is not within the preset safety target range, the earthquake PSA model is used to iteratively analyze the seismic design of the nuclear power plant.

[0150] The purpose of this step is to determine whether to conduct iterative analysis by comparing with the preset safety goals. If the risk results do not meet the safety goals, the seismic measures that should be set are proposed by identifying the weak links so that the nuclear power plant can meet the probabilistic safety goals. Among them, the preset safety goals are based on the design safety goals of specific nuclear power plants, and are comprehensively determined by considering the contributions of internal events, fires, and flooding to risks.

[0151] Furthermore, the seismic PSA model is used for iterative analysis, including five steps from S401 to S405:

[0152] S401, using the earthquake PSA model to identify weak links in seismic design and provide seismic design targeting the weak links.

[0153] Furthermore, the earthquake PSA model is used to identify weak links in the seismic design and provide seismic design for the weak links, including:

[0154] S4011, using the earthquake PSA model to determine the minimum cut set of the combination of failure of the initiating event superimposed on the equipment and human error events caused by the earthquake in each earthquake acceleration range.

[0155] S4012, analyzing the minimum cut sets, and determining the minimum cut sets in the minimum cut sets that are greater than or equal to the preset risk contribution.

[0156] The total cut set of various initiating events caused by earthquakes and equipment failure combinations in each earthquake acceleration range can be expressed as SE i is the ith earthquake acceleration interval, IE j is the jth initiating event, P1, P2, P k To mitigate the probability of equipment or personnel failure.

[0157] In the above embodiment, through the analysis of the minimum cut sets, it is found that the risk contribution of the cut sets in the top 30 reaches 65%, and all of them are core damage caused by the power operation condition SBO (Station Black-Out) accident. The cut sets in the top 30 are taken as the minimum cut sets greater than or equal to the preset risk contribution. Table 2 shows the basic events corresponding to the top five minimum cut sets in risk contribution in this embodiment.

[0158] Table 2 Examples of equipment failure events with top five risk contributions in the minimum cut set

[0159]

[0160]

[0161] S4013: Analyze the response measures of the minimum cut set that is greater than or equal to the preset risk contribution, and propose seismic design for the equipment in the response measures.

[0162] According to the basic events of the minimum cut set in Table 2 above, it is possible to consider improving the seismic resistance level of the power plant by strengthening the seismic resistance of emergency diesel generator-related equipment or other equipment with high risk contribution.

[0163] S402, determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link.

[0164] If a seismic design for improving the seismic resistance of emergency diesel generators or other equipment with high risk contribution has not been provided, execute S403; if a seismic design for improving the seismic resistance of emergency diesel generators or other equipment with high risk contribution has been provided, execute S405, set the equipment in the response measures for weak links to be seismic-resistant, and update the earthquake PSA model.

[0165] S403: If a seismic design to improve the seismic resistance of the original seismic equipment in the weak link has not been provided, further determine whether the seismic design is feasible.

[0166] S404: If the earthquake-resistant design is feasible, the earthquake PSA model is used again for quantitative analysis.

[0167] After the seismic design to improve the seismic resistance of emergency diesel generators or other equipment with high risk contribution, the failure probability of the corresponding equipment will change, and the earthquake PSA model is used again for quantitative analysis, and the corresponding step S3 is performed to determine whether the result of the quantitative analysis is within the preset safety target range, and whether iterative analysis is required according to the judgment result. If the earthquake PSA model is used again for quantitative analysis, and the new quantitative analysis result still exceeds the preset safety target range, the earthquake PSA model is used to re-identify the weak links in the seismic design, and the equipment in the countermeasures for the weak links is set to be earthquake-resistant.

[0168] S405, if the earthquake-resistant design is not feasible, the equipment in the countermeasures of the weak link is set to be earthquake-resistant, and the earthquake PSA model is updated.

[0169] In the above embodiment, if the seismic design to improve the seismic resistance of the original seismic equipment in the weak link is not feasible, the SBO diesel generator related system responding to the SBO accident is set to the seismic class and the earthquake PSA model is updated.

[0170] Specifically, updating the earthquake PSA model includes the following two sub-steps:

[0171] S4051, respectively update the event tree and fault tree of the mitigation system.

[0172] S4052: Update the earthquake PSA model using the updated event tree and fault tree.

[0173] Furthermore, when the earthquake PSA model itself is updated, the input of the earthquake PSA model is further updated, specifically including the following two steps: S4053, updating the failure probability of the mitigation equipment in the mitigation system; S4054, inputting the updated failure probability of the mitigation equipment into the updated earthquake PSA model for quantitative analysis.

[0174] As shown in Table 3, the results of re-quantitative analysis after the earthquake PSA model is updated in the above embodiment. Among them, the basic model refers to the earthquake PSA model initially established, and the corresponding CDF is the core damage frequency obtained by quantitative analysis using the initial earthquake PSA model; Scheme 1 is to execute step S405 to set the SBO diesel generator in the weak link countermeasures to earthquake resistance. After applying Scheme 1 to update the earthquake PSA model, the quantitative analysis result is 2.2E-07 / reactor year, and Scheme 1 has a significant reduction in the CDF result of the initial scheme. According to engineering experience, the CDF caused by internal events and fires is equivalent to the contribution of earthquakes, and the CDF caused by flooding is often one order of magnitude lower than that of earthquakes. Therefore, it is preliminarily judged that the core damage frequency caused by the superposition of internal events and external events can meet the requirements of the preset safety goals.

[0175] Table 3 CDF of each seismic design scheme for the DEC-A system

[0176]

[0177] In view of the fact that the quantitative analysis results after updating the earthquake PSA model by applying Scheme 1 are quite impressive, the iterative analysis process can be terminated at this time, and the impact of other seismic designs on the seismic risk of nuclear power plants can be further studied to provide support for the balance analysis of safety and economy in the design of nuclear power plants. In this embodiment, the necessity of designing other systems as seismic after setting the SBO diesel generator-related systems as seismic is further analyzed. The iterative analysis process based on Scheme 1 is the same as the above steps S401 to S405. After analyzing the minimum cut set, the contribution of the cut sets ranked in the top 30 reaches 13%. As shown in Table 4, it can be seen from the basic events included in the cut set that the main risk contributions come from the accident scenarios of partial rod jamming under power conditions, SBO accidents under one-loop open conditions, and the initiating event superimposed on the unavailability of the auxiliary water supply system. Therefore, it is possible to consider designing the cavity injection and cooling system (CIS) used to mitigate SBO accidents in the first-loop open condition, the emergency boron injection system (REB) to deal with partial rod jams, and the passive residual heat removal system (PRS) on the secondary side as earthquake-resistant types, so as to further improve the safety of the power plant.

[0178] Table 4 Examples of equipment failure events with top five risk contributions in the minimum cut set

[0179]

[0180]

[0181] After iterative analysis, the corresponding seismic design was added on the basis of Scheme 1, and seismic design schemes of Schemes 2, 3, 4, and 5 were formulated. The CDF under each seismic design scheme is shown in Table 2. Based on the above five schemes, it can be seen that Scheme 1 can greatly reduce the CDF and meet the preset safety goals. Schemes 2 to 5, on the basis of Scheme 1, add the way of designing CIS, REB, PRS as seismic type, which reduces the CDF compared with the original basic model, but there is an increase compared with Scheme 1. Therefore, the necessity of designing the relevant equipment of this type of system as seismic type can be further evaluated according to the economic requirements.

[0182] S5. Determine the seismic design for DEC-A condition based on the iterative analysis results and engineering factors.

[0183] For example, if the total construction cost budget of a nuclear power plant is 16,000 yuan / KW, considering that earthquake-resistant plans two, three, and four will cause the cost to exceed the budget, the corresponding earthquake-resistant design plan will not be implemented; in addition, combined with factors such as procurement and equipment post-maintenance, if the earthquake-resistant equipment in plans two, three, and four is difficult to purchase and maintain, the corresponding earthquake-resistant design plan will not be implemented.

[0184] To achieve the same purpose as the above method, the present invention also proposes a determination device for seismic design based on the earthquake PSA model.

[0185] like Fig. 9 9 shows a schematic diagram of a determination device for seismic design based on an earthquake PSA model according to an embodiment of the present invention, the device comprises: an establishment module 901, a quantitative analysis module 902, a judgment module 903, an iterative analysis module 904 and a design module 905. The functions of each module will be introduced below:

[0186] A module 901 is established for establishing a seismic PSA model for analyzing the seismic design under the extended working condition DEC-A without stack melting design.

[0187] Furthermore, a module 901 is established for:

[0188] Determine all possible initiating events caused by earthquakes, and obtain accident sequences that may lead to core damage in nuclear power plants based on all initiating events.

[0189] Furthermore, the accident sequences that may cause core damage in nuclear power plants are obtained based on all initiating events, including:

[0190] Establish an earthquake front tree model based on all initiating events;

[0191] The earthquake front-end tree model and the process analysis of all initiating events are used to determine the accident sequence that may lead to core damage in nuclear power plants.

[0192] Establish an event tree of the mitigation system corresponding to the accident sequence;

[0193] Based on the earthquake failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model is established to analyze the failure causes of the mitigation system.

[0194] The earthquake PSA model is established based on all initiating events, event trees and fault tree models.

[0195] The quantitative analysis module 902 is used to perform quantitative analysis on the frequency of nuclear power plant core damage caused by earthquakes using the earthquake PSA model.

[0196] Furthermore, the quantitative analysis module 902 is used to:

[0197] Obtain the occurrence frequency of initiating events and the frequency of earthquakes in different earthquake acceleration ranges;

[0198] Obtaining the failure probability of mitigation equipment and the failure probability of personnel in the mitigation system;

[0199] Based on the failure probability of the mitigation equipment, the failure probability of the personnel, the frequency of the initiating event and the earthquake occurrence frequency in different earthquake acceleration intervals, Formula 1 is used to determine the frequency of superimposed failure of different equipment and personnel after the initiating event occurs in different earthquake acceleration intervals. Formula 1: Among them, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n To mitigate the probability of equipment failure or personnel failure;

[0200] The frequencies of different equipment failures and personnel failures after the initiating event in different earthquake acceleration ranges are added together to obtain quantitative analysis results.

[0201] Further, obtaining the failure probability of the mitigation device in the mitigation system includes:

[0202] Determine mitigation equipment within the mitigation system based on the fault tree model;

[0203] Analyze the seismic vulnerability of the mitigation equipment and obtain the log-normal model parameters, which include the median seismic resistance, accidental uncertainty, and cognitive uncertainty;

[0204] According to the median seismic resistance capability, accidental uncertainty and cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2. Formula 2:

[0205]

[0206] Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β R is the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

[0207] The judgment module 903 is used to judge whether the result of the quantitative analysis is within the preset safety target range.

[0208] The iterative analysis module 904 is used to iteratively analyze the seismic design of the nuclear power plant using the seismic PSA model if the result of the quantitative analysis is not within the preset safety target range.

[0209] Furthermore, the iterative analysis module 904 is used to:

[0210] Use the earthquake PSA model to identify weak links in seismic design and provide seismic design targeting the weak links;

[0211] Determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link;

[0212] If no seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link, further determine whether the seismic design is feasible;

[0213] If the seismic design is feasible, the earthquake PSA model is used again for quantitative analysis;

[0214] If earthquake-resistant design is not feasible, the equipment in the countermeasures of the weak links is set to be earthquake-resistant and the earthquake PSA model is updated.

[0215] Furthermore, the earthquake PSA model is used to identify weak links in the seismic design and provide seismic design for the weak links, including:

[0216] The earthquake PSA model is used to determine the minimum cut set of the combination of the initial event caused by the earthquake and the failure of the equipment and human error events under each earthquake acceleration range;

[0217] Analyze the minimum cut sets and determine the minimum cut sets in the minimum cut sets that are greater than or equal to the preset risk contribution;

[0218] Analyze the countermeasures of the minimum cut set that is greater than or equal to the preset risk contribution, and propose seismic design for the equipment in the countermeasures.

[0219] Furthermore, the earthquake PSA model is updated, including:

[0220] Update the event tree and fault tree of the mitigation system respectively;

[0221] The earthquake PSA model is updated using the updated event tree and fault tree.

[0222] Furthermore, it also includes:

[0223] Update the failure probability of mitigation devices within the mitigation system.

[0224] The updated failure probability of the mitigation equipment is input into the updated earthquake PSA model for quantitative analysis.

[0225] The design module 905 is used to determine the seismic design for the DEC-A condition based on the iterative analysis results and engineering factors.

[0226] It should be understood that a device for determining seismic design based on an earthquake PSA model is consistent with the description of a corresponding method for determining seismic design based on an earthquake PSA model, so it will not be repeated in this embodiment.

[0227] In summary, it can be seen from the above description that the above embodiments of the present invention achieve the following technical effects:

[0228] 1. The present invention establishes an earthquake PSA model, utilizes the accident process analysis, earthquake vulnerability analysis and cut set analysis in the earthquake PSA model, and combines the characteristics of the non-melt design extended condition DEC-A to establish a method for determining the seismic design of the DEC-A condition response measures based on the earthquake PSA model. Based on the quantitative analysis results of the earthquake PSA model, the frequency of core damage caused by earthquakes in nuclear power plants is evaluated, providing a basis for determining the seismic design.

[0229] 2. The present invention utilizes the earthquake PSA model to iteratively evaluate the frequency of core damage caused by earthquakes in nuclear power plants, identifies the original earthquake-resistant equipment that needs to be considered in the weak links and the equipment in the mitigation measures for the weak links through cut set analysis, and finally combines the quantitative analysis results and engineering factors to comprehensively determine the earthquake-resistant design of the DEC-A countermeasures. This method and device can ensure that the economic benefits are maximized without reducing the safety of the power plant.

[0230] The above description is only a preferred embodiment of the present invention and is not intended to limit the present invention. For those skilled in the art, the present invention may have various modifications and variations. Any modification, equivalent replacement, improvement, etc. made within the spirit and principle of the present invention shall be included in the protection scope of the present invention.

[0231] It should be noted that, in this article, relational terms such as first and second, etc. are only used to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Moreover, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, the elements defined by the sentence "comprise a ..." do not exclude the existence of other identical elements in the process, method, article or device including the elements.

[0232] The logic and / or steps represented in the flowchart or otherwise described herein, for example, can be considered as an ordered list of executable instructions for implementing logical functions, which can be embodied in any computer-readable medium for use by an instruction execution system, apparatus or device (such as a computer-based system, a system including a processor or other system that can fetch instructions from an instruction execution system, apparatus or device and execute instructions), or used in combination with these instruction execution systems, apparatuses or devices.

[0233] It should be understood that the various parts of the present invention can be implemented by hardware, software, firmware or a combination thereof. In the above-mentioned embodiments, a plurality of steps or methods can be implemented by software or firmware stored in a memory and executed by a suitable instruction execution system. For example, if implemented by hardware, as in another embodiment, it can be implemented by any one of the following technologies known in the art or their combination: a discrete logic circuit having a logic gate circuit for implementing a logic function for a data signal, a dedicated integrated circuit having a suitable combination of logic gate circuits, a programmable gate array (PGA), a field programmable gate array (FPGA), etc.

[0234] It should be noted that, in the description of this specification, the description with reference to the terms "one embodiment", "some embodiments", "example", "specific example", or "some examples" etc. means that the specific features, structures, materials or characteristics described in conjunction with the embodiment or example are included in at least one embodiment or example of the present invention. In this specification, the schematic representations of the above terms do not necessarily refer to the same embodiment or example. Moreover, the specific features, structures, materials or characteristics described may be combined in any one or more embodiments or examples in a suitable manner. In addition, those skilled in the art may combine and combine the different embodiments or examples described in this specification and the features of the different embodiments or examples, unless they are contradictory.

Claims

1. A method for determining seismic design based on the earthquake PSA model, characterized in that: include: Establish a seismic PSA model for analyzing the seismic design under the extended condition DEC-A without stacking melting; The earthquake PSA model is used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes; Determine whether the result of the quantitative analysis is within the preset safety target range; If the result of the quantitative analysis is not within the preset safety target range, performing iterative analysis on the seismic design using the earthquake PSA model; Based on the iterative analysis results and engineering factors, the seismic design for the DEC-A condition is determined.

2. The method according to claim 1, characterized in that The seismic PSA model for analyzing the seismic design under the extended condition DEC-A without stacking melting is established, including: Determine all possible initiating events caused by earthquakes, and obtain accident sequences that may cause core damage to nuclear power plants based on all the initiating events; Establishing an event tree of a mitigation system corresponding to the accident sequence; Based on the seismic failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model for analyzing the failure causes of the mitigation system is established; An earthquake PSA model is established based on all the initiating events, the event tree and the fault tree model.

3. The method according to claim 2, characterized in that Obtain the accident sequence that may cause core damage to the nuclear power plant based on all the initiating events described, including: Establishing an earthquake front-end tree model according to all the initiating events; The earthquake front-end tree model and the process analysis of all the initiating events are used to determine the accident sequence that may cause core damage in the nuclear power plant.

4. The method according to claim 2, characterized in that: The earthquake PSA model is used to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes, including: Obtain the frequency of initiating events and the frequency of earthquakes in different earthquake acceleration ranges; Obtaining the failure probability of mitigation equipment and the failure probability of personnel in the mitigation system; Based on the failure probability of the mitigation equipment, the failure probability of the personnel, the frequency of the initiating event and the earthquake occurrence frequency in different earthquake acceleration intervals, Formula 1 is used to determine the frequency of superimposed failure of different equipment and personnel after the initiating event occurs in different earthquake acceleration intervals. Formula 1: Among them, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n It indicates the probability of failure of the mitigation equipment or personnel; The frequencies of different equipment failures and personnel failures after the initiating event in different earthquake acceleration ranges are added together to obtain quantitative analysis results.

5. The method according to claim 4, characterized in that Obtaining the failure probability of the mitigation device in the mitigation system, including: determining mitigation equipment within the mitigation system based on the fault tree model; Analyzing the seismic vulnerability of the mitigation equipment to obtain log-normal model parameters, wherein the log-normal model parameters include median seismic resistance, accidental uncertainty, and cognitive uncertainty; According to the seismic median capability, the accidental uncertainty and the cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2. The formula 2 is: Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β R is the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

6. The method according to claim 2, characterized in that An iterative analysis is performed using the earthquake PSA model, including: Using the earthquake PSA model to identify weak links in the seismic design, and providing seismic design for the weak links; Determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link; If no seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link, further determining whether the seismic design is feasible; If the seismic design is feasible, the earthquake PSA model is used again for quantitative analysis; If the earthquake-resistant design is not feasible, the equipment in the countermeasures for the weak link is set to be earthquake-resistant, and the earthquake PSA model is updated.

7. The method according to claim 6, characterized in that If the earthquake PSA model is used again for quantitative analysis and the new quantitative analysis results still exceed the preset safety target range, the earthquake PSA model is used to re-identify the weak links in the seismic design and set the equipment in the response measures for the weak links to be earthquake-resistant.

8. The method according to claim 6, characterized in that The earthquake PSA model is used to identify weak links in the seismic design, and seismic design for the weak links is provided, including: The earthquake PSA model is used to determine the minimum cut set of the combination of the initiating event caused by the earthquake and the failure of the equipment and the human error event under each earthquake acceleration interval; Analyzing the minimum cut sets, and determining a minimum cut set in the minimum cut sets that is greater than or equal to a preset risk contribution; The countermeasures of the minimum cut set that is greater than or equal to the preset risk contribution are analyzed, and a seismic design is proposed for the equipment in the countermeasures.

9. The method according to claim 6, characterized in that The earthquake PSA model is updated, including: updating the event tree and the fault tree of the mitigation system respectively; The earthquake PSA model is updated using the updated event tree and fault tree.

10. The method according to claim 8, characterized in that Also includes: updating a failure probability of a mitigation device within the mitigation system; The updated failure probability of the mitigation equipment is input into the updated earthquake PSA model for quantitative analysis.

11. A device for determining seismic design based on earthquake PSA model, characterized in that: include: Establishing a module for establishing a seismic PSA model for analyzing seismic design under the extended condition DEC-A without stacking and melting; A quantitative analysis module, used to use the earthquake PSA model to quantitatively analyze the frequency of nuclear power plant core damage caused by earthquakes; A judgment module, used to judge whether the result of the quantitative analysis is within a preset safety target range; An iterative analysis module, configured to iteratively analyze the seismic design using the earthquake PSA model if the result of the quantitative analysis is not within the preset safety target range; The design module is used to determine the seismic design for the DEC-A condition based on the iterative analysis results and engineering factors.

12. The device according to claim 11, characterized in that The establishment module is used to: Determine all possible initiating events caused by earthquakes, and obtain accident sequences that may cause core damage to nuclear power plants based on all the initiating events; Establishing an event tree of a mitigation system corresponding to the accident sequence; Based on the seismic failure events, random failure events and human error events of the SSC in the mitigation system, a fault tree model for analyzing the failure causes of the mitigation system is established; An earthquake PSA model is established based on all the initiating events, the event tree and the fault tree model.

13. The device according to claim 12, characterized in that Obtain the accident sequence that may cause core damage to the nuclear power plant based on all the initiating events described, including: Establishing an earthquake front-end tree model according to all the initiating events; The earthquake front-end tree model and the process analysis of all the initiating events are used to determine the accident sequence that may cause core damage in the nuclear power plant.

14. The device according to claim 12, characterized in that The quantitative analysis module is used for: Obtain the frequency of initiating events and the frequency of earthquakes in different earthquake acceleration ranges; Obtaining the failure probability of mitigation equipment and the failure probability of personnel in the mitigation system; Based on the failure probability of the mitigation equipment, the failure probability of the personnel, the frequency of the initiating event and the earthquake occurrence frequency in different earthquake acceleration intervals, Formula 1 is used to determine the frequency of superimposed failure of different equipment and personnel after the initiating event occurs in different earthquake acceleration intervals. Formula 1: Among them, SE is a certain earthquake acceleration interval, IE is the occurrence frequency of a certain initiating event, P1, P2, P n To mitigate the probability of equipment failure or personnel failure; The frequencies of different equipment failures and personnel failures after the initiating event in different earthquake acceleration ranges are added together to obtain quantitative analysis results.

15. The device according to claim 14, characterized in that Determining the failure probability of the mitigation device in the mitigation system according to the fault tree model includes: determining mitigation equipment within the mitigation system based on the fault tree model; Analyzing the seismic vulnerability of the mitigation equipment to obtain log-normal model parameters, wherein the log-normal model parameters include median seismic resistance, accidental uncertainty, and cognitive uncertainty; According to the seismic median capability, the accidental uncertainty and the cognitive uncertainty, the failure probability of the mitigation equipment at different confidence levels under different ground motion accelerations is determined by Formula 2. The formula 2 is: Where Φ represents the cumulative distribution function of the standard normal distribution, a represents the ground motion acceleration, and A m is the median earthquake resistance, β U is the epistemic uncertainty, β R is the accidental uncertainty, Q represents the confidence level, and f(a) is the failure probability of the mitigation equipment.

16. The device according to claim 12, characterized in that Iterative analysis module for: Using the earthquake PSA model to identify weak links in the seismic design, and providing seismic design for the weak links; Determine whether a seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link; If no seismic design has been provided to improve the seismic resistance of the original seismic equipment in the weak link, further determining whether the seismic design is feasible; If the seismic design is feasible, the earthquake PSA model is used again for quantitative analysis; If the earthquake-resistant design is not feasible, the equipment in the countermeasures for the weak link is set to be earthquake-resistant, and the earthquake PSA model is updated.

17. The device according to claim 16, characterized in that If the earthquake PSA model is used again for quantitative analysis and the new quantitative analysis results still exceed the preset safety target range, the earthquake PSA model is used to re-identify the weak links in the seismic design and set the equipment in the response measures for the weak links to be earthquake-resistant.

18. The device according to claim 16, characterized in that The earthquake PSA model is used to identify weak links in the seismic design, and seismic design for the weak links is provided, including: The earthquake PSA model is used to determine the minimum cut set of the combination of failure of the initiating event superimposed on the equipment and human error event caused by the earthquake in each earthquake acceleration interval; Analyzing the minimum cut sets, and determining a minimum cut set in the minimum cut sets that is greater than or equal to a preset risk contribution; The countermeasures of the minimum cut set that is greater than or equal to the preset risk contribution are analyzed, and a seismic design is proposed for the equipment in the countermeasures.

19. The device according to claim 16, characterized in that The earthquake PSA model is updated, including: updating the event tree and the fault tree of the mitigation system respectively; The earthquake PSA model is updated using the updated event tree and fault tree.

20. The device according to claim 18, characterized in that Also includes: updating a failure probability of a mitigation device within the mitigation system; The updated failure probability of the mitigation equipment is input into the updated earthquake PSA model for quantitative analysis.

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