A probabilistic safety evaluation method for dismounting the water room primary side baffle of a nuclear power plant steam generator

Abstract

The application discloses a kind of nuclear power plant dismounting steam generator water chamber primary side stop board during probability safety evaluation method, this method is by defining the boundary condition and duration of MID-LOOP working condition, ensure the accuracy of analysis object;Further through the fusion originating event screening, the intervention time window (success / failure criterion) of safety system is determined by thermal-hydraulic calculation, and personnel reliability analysis, constructs the comprehensive probability safety analysis logic model including safety system failure and human error.Can quantitatively calculate the core damage frequency and uncertainty under this special working condition, so as to fill the gap of refined risk assessment under the condition of specific low water level during refueling overhaul, provide quantitative basis for nuclear power plant optimization overhaul schedule and safety management.

Description

Technical Field

[0001] This invention relates to the field of probabilistic safety assessment of nuclear power plant operations, and more particularly to a probabilistic safety assessment method during the dismantling and installation of a primary side plug plate in a steam generator water chamber of a nuclear power plant. Background Technology

[0002] Currently, during nuclear power unit overhauls, in-service inspection items occupy time during low-low water levels and also block the critical path of the overhaul, hindering the optimization of the overall overhaul schedule. Therefore, the power plant plans to adopt a strategy of removing and installing the steam generator (SG) primary side water chamber plugs while the reactor core is fueled. This will ensure that SG heat transfer tube eddy current inspections can be carried out in parallel while the reactor core is fueled, reducing the time occupied by in-service inspection items during low-low water levels and optimizing the overall overhaul schedule.

[0003] However, for the above operating modes or strategies, the power plant cannot yet assess whether the safety risks introduced during this operation are acceptable, which limits the application of the strategy, affects the optimization of the nuclear power plant's overhaul schedule, and also affects the flexibility of operation, which is not conducive to improving the power plant's capacity factor. Therefore, it is necessary to study and establish a corresponding probabilistic safety evaluation model for nuclear power plants during the primary side plugging of the SG water chamber when the reactor core has fuel, and to quantitatively assess the risk of core damage and the risk of early large-scale release of radioactivity under this special operating condition. Summary of the Invention

[0004] The technical problem to be solved by the present invention is to address at least one deficiency of the related technologies mentioned in the background: power plants are unable to assess whether the safety risks introduced during the operation of removing and installing the steam generator primary side water chamber plugs when there is fuel in the reactor core are acceptable, which limits the application of this strategy and affects the optimization of the overhaul schedule of nuclear power plants.

[0005] The technical solution adopted by this invention to solve its technical problem is: to construct a probabilistic safety assessment method during the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant, including the following steps: S1: When modeling a nuclear power plant, the operating state of the power plant is defined using relatively constant power plant operating parameters. The power plant operating parameters during the disassembly and assembly of the primary side plug plate of the steam generator water chamber are determined as the target operating condition. The target operating condition is the MID-LOOP condition. The starting point and ending point of the safety analysis under the MID-LOOP condition are identified, and the initial conditions and duration of the target operating condition are analyzed and obtained. S2: Using initial conditions and duration as input, the identification method based on safety analysis technology analyzes the initiating events of various disturbances to the stable operation of power plants under the MID-LOOP condition to determine the list of initiating events and the frequency of initiating events; S3: Starting with each initiating event, and combining the corresponding security system response after the initiating event occurs, an event sequence logical model is established through event tree analysis; S4: Determine the intervention time window for each safety system response by defining the core damage criterion and conducting thermal-hydraulic calculations to obtain the success criterion for each safety system response, and determine the failure criterion for each safety system response based on each success criterion; S5: Perform fault tree analysis on the safety system based on the failure criteria of each safety system response to determine the failure probability of the safety system; S6: Classify the personnel error events that occurred before and after the initial event, and select personnel reliability analysis methods to determine the probability of personnel error events according to the classification. S7: Integrate the input analysis software with the frequency of initiating events, the probability of safety system failures, and the probability of human error events to establish a probabilistic safety analysis logic model. Perform quantitative analysis on the probabilistic safety analysis logic model to obtain the core damage frequency of each initiating event and conduct uncertainty analysis on the core damage frequency of each initiating event.

[0006] Optionally, the power plant operating parameters include mandatory core power level, primary loop water level and primary loop temperature, as well as optional primary loop opening status, containment status, decay heat level and decay heat removal mechanism.

[0007] Optionally, the identification methods for security analysis techniques include at least one of the following: conducting system engineering evaluation, logical deduction of correlations, referring to a list of past initiating events, and statistical analysis of historical operational experience; when there are at least two identification methods for security analysis techniques, one of them shall be selected as the primary identification method, and the others shall be selected as auxiliary identification methods.

[0008] Optionally, when the identification methods of security analysis technology include at least two methods, including the identification method of referring to a list of previous originating events, the identification method of referring to a list of previous originating events shall be selected as the primary identification method, and the other identification methods shall be used as auxiliary identification methods.

[0009] Optionally, the identification method based on safety analysis technology analyzes the initiating events that interfere with the stable operation of power plants to determine the list of initiating events and the frequency of initiating events, including: analyzing the initiating events that interfere with the stable operation of power plants to obtain a preliminary list of initiating events, and grouping the initiating events of similar power plants of PSA in previous lists of initiating events to determine the final list of initiating events. The final method for calculating the frequency of initiating events is determined based on whether the data source is sufficient and the preset operating boundary conditions.

[0010] Optionally, the core damage criteria can be defined as follows: the coolant collapse level in the reactor pressure vessel remains below the top of the active fuel assembly for a preset extended time threshold, or the peak temperature of the fuel cladding surface calculated based on a simplified core model is higher than a preset temperature.

[0011] Optionally, the definition of core damage criteria also includes: in the event of a reactive-introduced accident, the reactor does not return to a critical state.

[0012] Optionally, the security system includes: a front-end system for directly mitigating the initiating event and a support system for supporting the front-end system.

[0013] Optionally, the reliability data used for various analyses in steps S1 to S7 comes from a large amount of general data in existing databases and specific data collected from specific objects.

[0014] Optionally, S7 further includes: analyzing and identifying the dominant event sequence leading to core damage from the core damage sequence caused by each initiating event, decomposing the dominant event sequence into multiple minimal cut sets leading to core damage, analyzing and identifying the dominant minimal cut sets leading to core damage from the multiple minimal cut sets leading to core damage, and outputting the equipment importance and important human factors events in the personnel error events in the safety system based on the dominant minimal cut sets.

[0015] By implementing this invention, the following beneficial effects are achieved: This invention ensures the accuracy of the analysis by defining the boundary conditions and duration of the MID-LOOP condition. Furthermore, by integrating initiation event screening, thermo-hydraulic calculations to determine the intervention time window (success / failure criteria) for the safety system, and personnel reliability analysis, it constructs a comprehensive probabilistic safety analysis logical model that includes both safety system failures and human error. This model can quantitatively calculate the core damage frequency and uncertainty under this specific condition, thus filling the gap in refined risk assessment under specific low-water-level conditions during refueling overhauls and providing a quantitative basis for optimizing overhaul schedules and safety management in nuclear power plants. Attached Figure Description

[0016] The present invention will be further described below with reference to the accompanying drawings and embodiments. In the accompanying drawings: Figure 1 This invention illustrates an overall flowchart of an embodiment of a probabilistic safety assessment method for the disassembly and assembly of a primary side plugging plate in a steam generator water chamber of a nuclear power plant. Figure 2 This invention provides a probabilistic safety assessment method for the installation and removal of a primary side plugging plate in a steam generator water chamber at a nuclear power plant. The diagram illustrates the water level flow during the installation and removal of the plugging plate. Figure 3This invention illustrates an example diagram of a probabilistic safety assessment method for the dismantling and installation of a primary side plugging plate in a steam generator water chamber of a nuclear power plant, based on an embodiment of the present invention. (The diagram shows an example of a tree diagram illustrating the loss of column B, a safety-critical power source.) Figure 4 This paper presents an embodiment of the probabilistic safety assessment method for the primary side plugging plate of a steam generator water chamber in a nuclear power plant, based on an example of the uncertainty analysis of the core damage frequency of the MID-LOOP condition initiation event group. Detailed Implementation

[0017] To provide a clearer understanding of the technical features, objectives, and effects of the present invention, specific embodiments of the present invention will now be described in detail with reference to the accompanying drawings.

[0018] It should be noted that the flowcharts shown in the accompanying drawings are merely illustrative and do not necessarily include all content and operations / steps, nor do they necessarily have to be performed in the described order. For example, some operations / steps can be broken down, while others can be combined or partially combined; therefore, the actual execution order may change depending on the specific circumstances.

[0019] The block diagrams shown in the accompanying drawings are merely functional entities and do not necessarily correspond to physically independent entities. That is, these functional entities can be implemented in software, in one or more hardware modules or integrated circuits, or in different network and / or processor devices and / or microcontroller devices.

[0020] It should be noted that "at least two" refers to at least two, which can be two, three, or any number. "At least one" can be one, two, or any number.

[0021] like Figure 1 As shown, some embodiments of the present invention disclose a probabilistic safety assessment method during the dismantling and installation of a primary side plugging plate in a steam generator water chamber at a nuclear power plant, including: S1: When modeling a nuclear power plant, the plant's operating state is defined using relatively constant plant operating parameters (constant during modeling). The plant operating parameters during the removal and installation of the primary side plug plate of the steam generator water chamber are determined as the target operating condition, which is the MID-LOOP condition. The start and end points of the safety analysis under the MID-LOOP condition are identified, and the initial conditions and duration of the target operating condition are analyzed and obtained. The plant operating parameters include mandatory core power level, primary coolant level, and primary coolant temperature, as well as optional primary coolant opening status, containment status, decay heat level, and decay heat removal mechanism, etc., and their impact on risk differs from other configurations.

[0022] It is important to note that the operating status of a power plant can be defined using multiple operating parameters. Using as few parameters as possible, and ensuring they align with the parameters of interest under different operating conditions, facilitates model understanding and configuration control. Conversely, using as many parameters as possible provides a more realistic representation of the power plant's operating status. In practical applications, parameters can be selected based on specific engineering requirements.

[0023] Mid-loop operation (MID-LOOP) is a special operating condition where the unit operates at the MLO level, also known as LOI-RRA level, which is a challenging situation. Several major risks exist: 1) Due to the minimal primary coolant level, operator intervention time is shorter in the event of core decooling and core exposure; 2) The low primary coolant level makes the RRA pump inlet prone to vortex formation and damage; 3) During the removal of the SG chamber baffle, the primary coolant level may rise and enter the SG chamber, potentially causing personnel flooding. Nuclear power plant refueling overhauls mainly fall into two categories: those without a mid-loop operation and those with a mid-loop operation. Based on nuclear power plant overhaul practice, entering the mid-loop operation is only necessary under special circumstances (such as venting due to fuel damage or foreign objects in the steam generator).

[0024] Relevant system technical specifications clearly require that when using a steam generator plug, the pressurizer manhole or pressure vessel must be open, meaning the primary loop must be in a fully open state. At this time, the reactor pool must not be filled. For example... Figure 2 As shown, during the disassembly and assembly of the plug, the water level reaches the LOI-RRA level, which is the MLO level.

[0025] For the special condition of the primary circuit water level being in the MID-LOOP state during the installation and removal of the primary side water chamber blockage plate of the steam generator, it is necessary to identify and clarify the start and end points of the safety analysis, determine the initial conditions of this condition, such as the primary circuit water level, temperature and the current operating status of the RRA pump, and analyze the duration of this condition as input for subsequent initiation event analysis.

[0026] After clarifying the initial operating state, it is necessary to further identify the potential initiating events of the power plant. Initiating events refer to various internal events (occurring within the power plant) or external events (occurring outside the power plant) that disrupt the stable operation of the power plant. Initiating events trigger a series of events that require responses from the power plant's control and safety systems. Failure of these systems may lead to core damage (CD) or the early and large-scale release of radionuclides.

[0027] S2: Using initial conditions and duration as input, an identification method based on safety analysis techniques is used to analyze the initiating events that disrupt the stable operation of the power plant under the MID-LOOP condition to determine the list and frequency of initiating events. The determination of initiating events aims to identify and establish a comprehensive list of initiating events within the analysis scope, ensuring that the model established by the analysis realistically reflects the core damage risks faced by the power plant.

[0028] In some embodiments, the identification methods of safety analysis techniques include at least one of the following: conducting system engineering evaluation, logical deduction of correlations, referring to a list of past initiating events, and statistical analysis of historical operational experience. When at least two identification methods of safety analysis techniques are included, one is selected as the primary identification method, and the others are used as auxiliary identification methods. Each of the above methods has its limitations. Therefore, in order to obtain a set of initiating events as complete as possible, it is best to use multiple methods in combination, with one method as the primary one, when identifying initiating events. It should be noted here that the operational experience here does not refer to the subjective experience and judgment of the staff, but rather to the relevant records of various events and equipment failures in the power plant.

[0029] Furthermore, when the identification methods of safety analysis techniques include at least two methods, including the identification method of referring to a list of previous originating events, the method of referring to a list of previous originating events is selected as the primary identification method, and the other identification methods are used as auxiliary identification methods. For PSA projects that have existing PSAs of similar nuclear power plants as references, the second method (referring to a list of previous originating events) is generally used as the primary method, as it can directly provide a reference basis.

[0030] Preferably, the identification methods of security analysis technology include: conducting system engineering evaluation, logical deduction of correlation, referring to the list of previous initiating events, and statistical analysis of historical operating experience. The list of previous initiating events is selected as the primary determination method, and the other identification methods are used as auxiliary determination methods.

[0031] In this embodiment, specifically, the establishment of the initiation event list referenced initiation events under other operating conditions of the same reactor type, analyzed the initiation event study on the MID-LOOP condition in a previous nuclear engineering report, and also combined the initiation event comparative study with PSA reports from three other units. The analysis of internal initiation events during the primary side plugging plate period of the SG water chamber when the core is fueled was carried out in the following steps: 1) Compare the operating status and initiation events of the three reactor types under other operating conditions to conduct a similarity and difference analysis, and preliminarily compile a list of possible initiation events under the MID-LOOP operating condition; 2) Referring to the initiation event analysis in a previous nuclear engineering report and confirming its conformity with the unit's design characteristics, compile a list of initiation events applicable to the three reactor types; 3) Conduct a comparative study of the initiation events of the three units to analyze whether there are any initiation event types that are applicable to the three reactor types but have been omitted; 4) Conduct engineering analyses, such as Failure Mode and Effects Analysis (FMEA), to systematically evaluate each system to determine whether its failure will trigger an initiating event.

[0032] After the above steps, the final list of initiating events for the MID-LOOP condition is shown in Table 1.

[0033] Table 1 Preliminary List of Originating Events In some embodiments, the identification method based on security analysis technology analyzes various initiating events that interfere with the stable operation of power plants to determine the initiating event list and the initiating event frequency. This includes: analyzing various initiating events that interfere with the stable operation of power plants to obtain a preliminary initiating event list; grouping initiating events of similar power plants under PSA in previous initiating event lists to determine the final initiating event list; and determining the final initiating event frequency calculation method based on whether the data source is sufficient and preset operating boundary conditions to calculate the final initiating event frequency.

[0034] Specifically, the identification method based on safety analysis technology analyzes various initiating events that interfere with the stable operation of the power plant to obtain a preliminary list of initiating events as shown in Table 1 above. The final list of initiating events is determined by mainly referring to the grouping of initiating events in existing PSA-type power plants. The final grouping and description of initiating events are shown in Table 2.

[0035] Table 2 Grouping and Explanation of Originating Events In some embodiments, the grouping method for initiating events also includes: similar incident processes, system responses, personnel operations, event identification, and comparison of operational experience.

[0036] Specifically, a list of initiating events (groups) is formed based on the principle of merging and grouping initiating events, and the frequency of initiating events (groups) must be determined. The final initiating event frequency calculation method is determined based on the sufficiency of data sources and preset operational boundary conditions.

[0037] Furthermore, if the necessary power plant-specific data is sufficient in the data source for the initial event, the power plant-specific data can be used to calculate the corresponding frequency of the initial event; if the necessary power plant-specific data is insufficient in the data source for the initial event, data from similar power plants or general data combined with data processed by Bayes can be used to calculate the corresponding frequency of the initial event; if the preset operating boundary conditions (such as design, site, and operating environment) of the initial event are different from the data source, a special analysis is required to calculate the corresponding frequency of the initial event.

[0038] Specifically, this unit has over two hundred reactor years of operating history and has accumulated some important data related to PSA (Power Supply Availability). Therefore, when conducting initiation event analysis, especially in determining initiation events and their frequencies, the operating history of similar power plants should be fully considered first. For some initiation event groups that rely on the operating experience of multiple plants but lack sufficient data, the French General Data (EPS900) can be used as prior data for Bayesian calculations, such as boron dilution initiation events. For initiation events that have not occurred before, such as small breaks on the RRA (Residual Radiation Protection), EPS900 data is used. For specialized calculations, such as LOCA (Location of Action) caused by maintenance, calculations need to be carried out in conjunction with actual overhaul work, valve quantity, etc.; loss of heat sink, loss of residual heat removal system, loss of safety-critical power supply, and loss of external power are all calculated using the fault tree method to determine the frequency of initiation events. The final list of initiation events (groups) and frequency analysis results are summarized in Table 3.

[0039] Table 3. List and frequency of initiating events (groups) after grouping Once the initiating event is determined, the next step is to determine the plant's response to each initiating event group, identifying the sequence of events that may cause core damage or, in the event of reactor shutdown, ensure residual heat removal and maintain the plant in a safe state. After an initiating event occurs, to mitigate the consequences and bring the reactor to a safe state, systems related to safety functions, i.e., front-end systems, need to respond sequentially according to the event's progression (i.e., safety measures taken after the initial event, such as personnel following accident procedures and activating safety systems as required, or automatic activation of safety systems). Failure of a safety system response may lead to core damage. PSA uses event tree analysis to model and analyze the initiating event and subsequent responses (i.e., the functioning of safety systems), deriving the sequence of events based on the success or failure of the initiating event and subsequent safety functions. Conventionally, a tree-like bifurcated logic diagram, starting from the initiating event and arranged according to the success or failure of each safety function / safety system and personnel intervention, is called an event tree.

[0040] S3: Starting with each initiating event, and combining the corresponding security system response after the initiating event occurs, an event sequence logical model is established through event tree analysis.

[0041] In some embodiments, a functional event tree corresponding to the security system response is established by taking each initiating event as the starting point and combining it with the corresponding security system response after the initiating event occurs. The functional event tree is then converted into a system event tree for the corresponding security system, and an event sequence logic model is established through event tree analysis. Alternatively, a system event tree for the corresponding security system is obtained by taking each initiating event as the starting point and combining it with the corresponding security system response after the initiating event occurs, and the security system response is gradually derived according to the initiation time. The event sequence logic model is then established through event tree analysis.

[0042] Specifically, the latter method directly performs event sequence analysis. It starts from the initial event and deduces the power plant system's response step by step according to the development of events, thereby obtaining the system event tree. This method is generally suitable for personnel familiar with power plants and event tree analysis methods, and the system event tree is generally also directly referred to as an event tree.

[0043] Specifically, the event sequence modeling method employs the so-called small event tree / large fault tree approach. In this method, the correlation between the leading system and the supporting systems does not appear in the event tree; that is, the header of the event tree only includes the commissioning of the leading system and the required operator actions. Any system that provides services to the leading system (such as power systems, cooling water systems, and instrument gas systems) is generally referred to as a supporting system, and it does not appear in the event tree. Supporting systems are considered in the fault tree.

[0044] For each initiating event, a detailed event sequence logical model is established, combining the actions of personnel after the incident and the response of the mitigation system, such as... Figure 3 The image shows an example of an event tree for the loss of safety-critical power in column B. The numbers in the upper left corner of the image represent the starting event and each subsequent processing node for the corresponding example. The right side shows the sequence number, frequency of occurrence, result, and corresponding event sequence number for each event sequence.

[0045] S4: Determine the intervention time window for each safety system response by defining the core damage criterion and conducting thermal-hydraulic calculations to obtain the success criterion for each safety system response, and determine the failure criterion for each safety system response based on each success criterion.

[0046] Specifically, the overall success criterion for PSA aims to define core damage (and early mass release of radionuclides), which means first determining the definition of core damage. Core damage is defined as: the reactor core being exposed and heated to a degree that is expected to cause prolonged cladding oxidation and severe fuel damage, and the affected portion of the core being sufficient to cause a mass release of radionuclides.

[0047] Furthermore, since the shutdown time is already sufficiently long, the core damage criteria are defined as follows: the coolant collapse level in the reactor pressure vessel remains below the top of the active fuel assemblies for a predetermined extended time threshold; or, the peak temperature of the fuel cladding surface calculated based on the simplified core model is higher than a predetermined temperature, which is set to 982°C. For reactive introduction accidents (such as boron dilution accidents), the reactor does not return to criticality.

[0048] In some embodiments, in actual analysis, an indirect core damage criterion may be adopted, in which core damage is considered to have occurred as long as it is unlikely that mitigation measures can be taken to prevent long-term core exposure.

[0049] In addition to the overall success criteria for PSA, the success criterion analysis also identifies and describes success criteria for system, structure, equipment, and personnel actions. These success criteria must be consistent with the characteristics, procedures, and operating principles of the power plant. It can be noted that, due to the long unit shutdown time and very low decay heat, a minimum safety series is sufficient to meet the mitigation functions required to prevent core damage or radioactive material release, such as one RRA for residual heat removal and one RCV pump for primary loop makeup water supply. The results of the thermal-hydraulic analysis also support this point. The time window for operator intervention was determined during the development of this invention through thermal-hydraulic calculations, the results of which are as follows: 1) The success criteria and time window for personnel intervention in a complete loss of RRA cooling accident are: before the first calculation result after the accident (the first calculation result is 6660s), one RCV charging pump is started for reactor makeup water (flow rate 20m³ / h). 2) The success criteria and personnel intervention time window for RRA pipeline rupture accidents are: before the second calculation result after the accident (the second calculation result is 6490s), one RCV charging pump is started for reactor makeup water (flow rate 20m³ / h). 3) The success criteria and personnel intervention time window for an accident caused by RCV leakage failure leading to uncontrolled water level drop are: before the third calculation result after the accident (the third calculation result is 6494s), one RCV charging pump is started for reactor makeup water (flow rate 20m³ / h). 4) The success criteria and personnel intervention time window for a plant-wide power outage accident are: before the fourth calculation result after the accident (the fourth calculation result is 6660s), one RCV charging pump is started for reactor makeup water (flow rate 20m³ / h); conservatively, the effect of gravity makeup water is not considered, and the complete loss of RRA cooling accident can be referred to.

[0050] 5) The success criteria and personnel intervention time window for boron dilution accidents are: the operator must manually isolate the dilution source within the fifth calculation result after the accident (the fifth calculation result is 20 minutes).

[0051] To quantitatively determine the occurrence frequency of incident sequences in an event tree, incident sequence analysis is necessary. The foundation of incident sequence analysis is determining the heads of each event in the event tree and their quantification methods. For mitigation systems, fault tree methods are used to quantify their failure probabilities; for personnel actions, personnel reliability analysis methods are used. After an initiating event occurs, safety systems need to be deployed to mitigate the consequences. However, these systems may be able to fulfill their designed functions, or they may fail and be unable to perform their designed functions. Therefore, system reliability analysis is required. Constructing a system fault tree requires determining the top event and the system success criteria. Success criteria are the opposite of failure criteria (e.g., if a system consists of three columns with 100% operational capability, and the success criterion is one column success, the failure criterion is all three columns failure). Therefore, in system analysis, once the top event and success criteria are determined, the fault tree can be constructed.

[0052] S5: Perform fault tree analysis on the safety system based on the failure criteria of each safety system response to determine the failure probability of the safety system. The safety system includes: a front-end system for directly mitigating the initiating event and a support system for supporting the front-end system.

[0053] Specifically, the purpose of systems analysis is to evaluate the pathways and probabilities of system failures needed to mitigate incidents. The primary method of systems analysis is fault tree analysis. Fault tree analysis is a logical analysis method for a specific undesirable event (such as the system failing to perform its designed functions) to determine the various causes of that undesirable event and their probabilities.

[0054] In probabilistic safety analysis, systems are generally divided into two categories: front-end systems and support systems. Front-end systems are those that directly mitigate accidents after the initial event occurs. Support systems are those that are deployed to support the front-end systems in fulfilling their functions, such as systems providing power or cooling to components within the front-end system. The scope of system analysis includes the analysis of both front-end and support systems. The analysis software used in this invention for safety system analysis is the RS program.

[0055] S6: Classify personnel error events based on their occurrence before and after the initial event, and determine the probability of personnel error events using personnel reliability analysis methods according to the classification. Specifically, this includes classifying personnel error events into pre-accident and post-accident human error events. For pre-accident human error events, the ASEP method is used to determine the probability of personnel error events, and for post-accident human error events, the SPAR-H method is used to determine the probability of personnel error events.

[0056] Specifically, based on the personnel error events occurring before and after the initial event, personnel error events (human error events) are divided into pre-accident human error events and post-accident human error events. Pre-accident human error events refer to situations where, when the unit is under normal operating conditions, errors made by personnel during previous component maintenance or calibration cause these components to be in an abnormal state or position. When the system needs to be put into operation after an accident, these components may prevent the system or a part of it from performing its designed function. For example, a manual valve that should be in the open position may have been left open after maintenance; or a setting error in the dedicated safety facility trigger system may prevent the dedicated safety facility from triggering when it should. Post-accident human error events refer to errors that occur after an accident when the operator diagnoses the accident and performs the corresponding operations according to the accident procedures. Post-accident human error events involve the operator's diagnosis and operation after the accident, which differs in each initial event; therefore, post-accident human error events are mainly considered in the event tree. Human error events considered in the fault tree are mainly pre-accident human error events. However, a very small number of human errors following an accident are due to routine system operations or operations performed in the later stages of an accident. These events are basically the same in each initial event and are also considered in the fault tree.

[0057] Furthermore, during the unit operation phase, the ASEP method is used to systematically evaluate the identified pre-accident human errors. When determining the probability of pre-accident human errors, the ASEP method considers the following factors: 1) Are there any obvious alarms in the main control room? 2) Whether there is a valid test after maintenance or setting; 3) Are there any independent verifications unrelated to the experiment? 4) Are there routine inspections? 5) Regularly test recovery factors.

[0058] Post-initial event human reliability errors refer to errors made by power plant personnel during the response process after an initial event occurs. These types of human error events are represented in both event trees and fault trees.

[0059] Post-accident human error events in the event tree are relatively complex, with a structure basically the same as Class C human error events in the fault tree above: they mainly consist of two parts of personnel response behavior, namely, immediate diagnostic behavior after the accident and intervention operation behavior according to relevant procedures after diagnosis. Therefore, the possible error probability of post-accident human error events also consists of these two parts. For personnel reliability under the MID-LOOP condition, the SPAR-H method is used for evaluation.

[0060] When conducting quantitative analysis of human error events following an accident, the response process is complex and varies depending on the accident. Therefore, the analysis is conducted in the following steps: 1) Event description and time analysis: Based on the results of the event tree analysis, determine the prerequisites and conditions for personnel reliability, the main circumstances of the event, the personnel and equipment involved, and the time window for analyzing the event.

[0061] 2) Quantitative Analysis: Based on feedback from reference operating nuclear power plants, the required time for diagnosis and operation during an accident is determined. The SPAR-H method is used to assess personnel reliability in two parts, and the corresponding Personnel Performance Formation Factor (PSF) is determined according to the specific type of diagnosis and operation. Finally, the probability of personnel reliability is calculated. If necessary, the correlation between personnel reliability profiles should also be evaluated, and an error factor should be calculated.

[0062] Post-accident personnel response in the fault tree is relatively simple, generally involving only the starting and stopping of backup equipment, especially starting equipment when the automatic start signal fails. These are routine system operations, but may not be covered or directly addressed in the accident procedures. Situations where operations are not directly addressed in the procedures mean that the main control room operator is primarily focused on executing the relevant operations in the accident procedures and has not noticed the failure of the normally operating system and automatic actions, thus failing to take timely manual intervention. For operations directly involved in the accident procedures, these generally occur in the later stages of the accident, when the operator is already executing the accident handling procedures; therefore, the possibility of diagnostic errors is small or nonexistent, and only operational errors are generally considered. It is important to note that operations on different system equipment may be related. For example, if an RRI or SEC fails and requires manual activation of the backup column, the backup columns for both RRI and SEC must be activated simultaneously. In this case, the operation of activating the backup columns for both RRI and SEC should be considered as a single event in the fault tree.

[0063] S7: The frequency of initiating events, the probability of safety system failures, and the probability of human error events are input into the analysis software to establish a probabilistic safety analysis logic model. Quantitative analysis is performed on the probabilistic safety analysis logic model to obtain the core damage frequency of each initiating event, and uncertainty analysis is conducted on the core damage frequency of each initiating event. In this invention, the analysis software Risk Spectrum is used to integrate and establish the probabilistic safety analysis logic model. It should be noted that the above analysis software is for illustrative purposes only, and other software can be selected according to actual conditions. The obtained core damage frequencies of initiating events are shown in Table 4, and the uncertainty analysis results of the core damage frequencies of initiating events are shown in Table 5. Figure 4 As shown.

[0064] Table 4. Distribution of estimated core damage frequency points under MID-LOOP condition with the initiating event group Table 5 Uncertainty Analysis of Core Damage Frequency in MID-LOOP Condition Initiation Event Group This embodiment, taking into account the characteristics of the MID-LOOP operating condition, completed the research on the above technical elements, and studied and established a probabilistic safety model development method applicable to this operating condition. It can be used to assess the safety risks during the primary side plugging of the SG water chamber when the reactor core is fueled. The assessment results show that the risk level of this operating condition and the risk increment caused by carrying out this work are within an acceptable range. The results fill the gap in the inability to assess the risk of this operating condition, strongly support the safety risk assessment of the strategy of installing and removing the SG primary side water chamber plugging plate when the reactor core is fueled, and support the optimization of the overhaul period of nuclear power plants.

[0065] In some embodiments, S7 further includes: analyzing and identifying the dominant event sequence leading to core damage from the core damage sequence caused by each initiating event, decomposing the dominant event sequence into multiple minimal cut sets leading to core damage, analyzing and identifying the dominant minimal cut sets leading to core damage from the multiple minimal cut sets leading to core damage, and outputting the equipment importance and important human factors events in the personnel error events in the safety system based on the dominant minimal cut sets.

[0066] Furthermore, the dominant event sequence is the first preset number of core damage sequences (e.g., the first 20, the first 30) arranged in descending order of frequency among the core damage sequences caused by each initiating event. The dominant minimum cut set is the second preset number of minimum cut sets (e.g., the first 20, the first 30) arranged in descending order of frequency among multiple minimum cut sets leading to core damage. It should be noted that the specific values ​​of the first and second preset number of positions listed above are for illustrative purposes, and the specific values ​​can be set to other values ​​according to actual needs.

[0067] For example, the dominant event sequence leading to core damage, the dominant minimal cut set leading to core damage, the equipment importance in the safety system, and the important human factors in personnel error events are shown in Tables 6, 7, 8, and 9, respectively.

[0068] Table 6. Dominant event sequence (first 20) leading to core damage under MID-LOOP conditions. Table 7 Dominant Minimal Cut Sets Leading to Core Damage under MID-LOOP Conditions (Top 30) Table 8 Equipment Importance Table 9 Significant Human Factors Events Reliability data is one of the foundations for conducting PSA (Power Supply Analysis) work in nuclear power plants and serves as the basis for nuclear power plant safety analysis. In Level 1 PSA work, in order to assess the frequency of occurrence of each event sequence, it is necessary to quantify the initiating event, the basic event-type header event, and the bottom event in the system fault tree within the event sequence.

[0069] In some embodiments, the reliability data used for various analyses in steps S1 to S7 comes from a large amount of general data in existing databases and specific data collected from specific objects; the specific data includes actual data of specific equipment collected from power plants.

[0070] Furthermore, using general data as prior data, Bayesian methods are employed to combine general data with specific data to obtain reliability data for analysis. The main components of this reliability data can be categorized as follows: 1) Frequency of initiating events; 2) Equipment startup failure probability γ or operational failure rate λ; 3) Unavailability caused by equipment testing and maintenance; 4) Common cause failure probability; 5) Probability of human error.

[0071] General data refers to statistical data derived from a large amount of operational experience in a database. This type of data typically provides the results of a single event and indicates the boundary conditions and uncertainties of the failed device. In the design phase of a nuclear power plant, when conducting system reliability or probabilistic safety analyses, general data is generally used due to the lack of specific data.

[0072] Specific data refers to data collected from the operational experience of a specific object, applicable to PSA research on that specific object. The application scope of a specific database is limited. After collecting specific data, Bayesian techniques are generally needed to process it and combine it with general data.

[0073] Taking a unit in Nuclear Power Plant A as an example, since the equipment and systems of Units 1 and 2 in Nuclear Power Plant A are quite similar to those in Nuclear Power Plant B, and Nuclear Power Plant B adopts French standard design, the design and manufacturing of most of its equipment are no different from those in France. Therefore, the general reliability data of the equipment used in PSAs 1 and 2 of Nuclear Power Plant A are basically taken from the French EDF manual. The specific reliability data of Nuclear Power Plant A equipment developed are basically based on the statistical data since the operation of Nuclear Power Plant A. For equipment with insufficient failure data, the equipment reliability parameters in the French EDF manual are used as prior data, and Bayesian methods are used to combine them with the data collected by Nuclear Power Plant A. However, corresponding corrections have been made for some design differences in the power plant (such as LLS system generator units and equipment that appeared after the adoption of the digital instrumentation and control system DCS).

[0074] By implementing this invention, the following beneficial effects are achieved: This invention ensures the accuracy of the analysis by defining the boundary conditions and duration of the MID-LOOP condition. Furthermore, by integrating initiation event screening, thermo-hydraulic calculations to determine the intervention time window (success / failure criteria) for the safety system, and personnel reliability analysis, it constructs a comprehensive probabilistic safety analysis logical model that includes both safety system failures and human error. This model can quantitatively calculate the core damage frequency and uncertainty under this specific condition, thus filling the gap in refined risk assessment under specific low-water-level conditions during refueling overhauls and providing a quantitative basis for optimizing overhaul schedules and safety management in nuclear power plants.

[0075] Filling a technological gap: A probabilistic safety model development method applicable to the MID-LOOP operating condition was established, solving the technical problem that nuclear power plants cannot assess the safety risks during the installation and removal of the primary side plug plate of the SG water chamber when there is fuel in the reactor core.

[0076] Precise risk assessment: This model can quantitatively assess the risk level and risk increment under this special operating condition, verify the controllability of the risk, and provide a scientific basis for power plant safety decisions.

[0077] Support for overhaul optimization: The research results directly support the safe implementation of the SG primary side plugging plate installation and removal strategy in nuclear power plants when the reactor core is fueled, significantly saving overhaul time and improving the economic efficiency of nuclear power plants.

[0078] This technology provides an innovative solution for risk assessment under special operating conditions in nuclear power plants, offering both safety and economic benefits.

[0079] It is understood that the above embodiments only illustrate some implementation methods of the present invention, and their descriptions are relatively specific and detailed, but they should not be construed as limiting the scope of the present invention. It should be noted that those skilled in the art can freely combine the above embodiments or technical features without departing from the concept of the present invention, and can also make several modifications and improvements, all of which fall within the protection scope of the present invention. That is, the embodiments described "in some embodiments" can be freely combined with any of the preceding and following embodiments. Therefore, all equivalent transformations and modifications made within the scope of the claims of the present invention should be covered by the claims of the present invention.

Claims

1. A probabilistic safety assessment method for the dismantling and installation of a primary side plug plate in a steam generator water chamber at a nuclear power plant, characterized in that... Includes the following steps: S1: When modeling a nuclear power plant, the operating state of the power plant is defined using relatively constant power plant operating parameters. The power plant operating parameters during the disassembly and assembly of the primary side plug plate of the steam generator water chamber are determined as the target operating condition. The target operating condition is the MID-LOOP condition. The start and end points of the safety analysis under the MID-LOOP condition are identified, and the initial conditions and duration of the target operating condition are analyzed and obtained. S2: Using the initial conditions and the duration as input, the identification method based on safety analysis technology is used to analyze the initiating events of various disturbances to the stable operation of the power plant under the MID-LOOP condition to determine the list of initiating events and the frequency of initiating events; S3: Starting with each of the aforementioned initiating events and combining the corresponding security system responses after the occurrence of the initiating event, an event sequence logical model is established through event tree analysis; S4: Determine the intervention time window for each of the safety system responses by defining the core damage criteria and performing thermo-hydraulic calculations to obtain the success criteria for each of the safety system responses, and determine the failure criteria for each of the safety system responses based on each of the success criteria; S5: Perform fault tree analysis on the safety system according to the failure criteria of each safety system response to determine the failure probability of the safety system; S6: Classify the personnel error events that occurred before and after the initiating event, and determine the probability of personnel error events by selecting personnel reliability analysis methods according to the classification. S7: Integrate the frequency of the initiating event, the probability of the safety system failure, and the probability of the personnel error event into the analysis software to establish a probabilistic safety analysis logic model. Perform quantitative analysis on the probabilistic safety analysis logic model to obtain the core damage frequency of each initiating event and perform uncertainty analysis on the core damage frequency of each initiating event.

2. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, The power plant operating parameters include mandatory core power level, primary loop water level and primary loop temperature, as well as optional primary loop open state, containment state, decay heat level and decay heat removal mechanism.

3. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, The identification method of the security analysis technology includes at least one of the following: conducting system engineering evaluation, logical deduction of correlation, referring to the list of previous initiating events, and statistical analysis of historical operating experience; when the identification method of the security analysis technology includes at least two methods, one of them shall be selected as the primary determination method and the others shall be selected as auxiliary determination methods.

4. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 3, characterized in that, When the identification method of the security analysis technology includes at least two methods, and includes the identification method of referring to a list of previous originating events, the identification method of referring to a list of previous originating events shall be selected as the primary identification method, and the other identification methods shall be used as auxiliary identification methods.

5. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 4, characterized in that, The identification method based on security analysis technology analyzes various initiating events that interfere with the stable operation of power plants to determine the initiating event list and the frequency of the initiating events. This includes: analyzing various initiating events that interfere with the stable operation of power plants to obtain a preliminary initiating event list, and grouping initiating events of similar power plants under PSA in previous initiating event lists to determine the final initiating event list. The final method for calculating the frequency of the initiating event is determined based on whether the data source is sufficient and the preset operating boundary conditions, so as to calculate the final frequency of the initiating event.

6. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, The defined core damage criteria include: the coolant collapse level in the reactor pressure vessel remains below the top of the active fuel assembly for a preset extended time threshold, or the peak temperature of the fuel cladding surface calculated based on the simplified core model is higher than a preset temperature.

7. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 6, characterized in that, The definition of core damage criteria also includes: in the event of a reactive-introduced accident, the reactor does not return to a critical state.

8. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, The security system includes: a front-end system for directly mitigating the initiating event and a support system for supporting the front-end system.

9. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, The reliability data used for various analyses in steps S1 to S7 comes from a large amount of general data in existing databases and specific data collected from specific objects.

10. The probabilistic safety assessment method for the dismantling and installation of the primary side plugging plate of the steam generator water chamber in a nuclear power plant according to claim 1, characterized in that, S7 further includes: analyzing and identifying the dominant event sequence leading to core damage from the core damage sequence caused by each of the initiating events, decomposing the dominant event sequence into multiple minimal cut sets leading to core damage, analyzing and identifying the dominant minimal cut sets leading to core damage from the multiple minimal cut sets leading to core damage, and outputting the equipment importance and important human factor events in the personnel error events in the safety system based on the dominant minimal cut sets.

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