A method for fatigue analysis and damage management of a primary loop pipeline of a nuclear power plant based on digital twinning

Abstract

The present application relates to a kind of nuclear power plant one loop pipe fatigue analysis and damage management method based on digital twinning, comprising the following steps: a) based on the frequency of occurrence, nuclear power plant one loop transient is classified and managed: (a.1) the transient-A class transient of the transient of one loop main equipment occurrence;(a.2) the transient-B class transient of the transient of branch system connected with main equipment and not often used;(a.3) the transient-C class transient of the transient of branch system connected with main equipment and often used;B) fatigue analysis is carried out to each class of transient respectively: (b.1) A class transient is carried out fatigue fast assessment based on design transient statistics;(b.2) B class transient is carried out real-time assessment based on online monitoring of operating parameter;(b.3) C class transient is carried out conservative assessment based on transient cumulative occurrence time or crack propagation;C) nuclear power plant operation control optimization based on digital twinning: based on the real-time monitoring of operating parameter and the classification management of transient, operation control strategy is proposed to reduce the fatigue damage of nuclear power plant one loop pipe.

Description

Technical Field

[0001] This invention belongs to the field of structural integrity assessment technology, specifically relating to a fatigue analysis and damage management method for primary loop pipelines in nuclear power plants based on digital twins, providing accurate technical basis for the safety assessment of key nuclear power equipment. Background Technology

[0002] Feedback from actual operation experience at pressurized water reactor nuclear power plants indicates that the fluctuations in transient parameters of the primary loop piping are often more complex than assumed during the design phase. Operational experience shows that multiple cracking events have occurred in the primary loop piping, most of which were caused by rapid transient fluctuations in the primary loop piping.

[0003] Swedish scholar Studies have shown that thermal shock and thermal oscillation on the inner surface of pipes significantly reduce their fatigue life, and the propagation rate of thermal shock-induced cracks decreases significantly after the depth exceeds 200 μm. German scholar Paffumi E's research indicates that when the crack depth is less than 1 mm, the existing Paris criterion for calculating crack propagation rate is not conservative, suggesting that macroscopic fracture mechanics analysis in engineering should focus on cracks with a depth less than 1 mm (specifically related to load type and material microstructure). Domestically, studies by Zhao Zhide, Wang Donghui, and others have shown that transient thermal stress is mainly concentrated on the inner surface of the pipe and decays rapidly along the pipe wall thickness. Related numerical simulation results indicate that as near-surface thermal fatigue cracks propagate along the wall thickness, the crack stops propagating after reaching a certain size (the fluctuation of fracture parameters is less than the threshold value for crack fatigue propagation), or propagates at a very low rate.

[0004] The evaluation of primary loop piping requires consideration of numerous influencing factors, including geometric information, transient parameters (a large amount of design transient and actual operational transient information), material properties (considering material thermal aging, etc.), and evaluation criteria. Therefore, developing a digital twin-based fatigue analysis and damage management method for nuclear power plant primary loop piping is crucial for improving the accuracy of evaluations and ensuring the safe and economical operation of nuclear power plants. However, existing publicly available literature lacks a suitable method for piping fatigue analysis and damage management using digital twins. Summary of the Invention

[0005] The purpose of this invention is to overcome the shortcomings of the prior art by providing a method for fatigue analysis (obtaining fatigue damage coefficient CUF) and damage management (optimizing operation and reducing CUF caused by transients in subsequent operation) of primary loop piping in nuclear power plants based on digital twins.

[0006] To achieve the above objectives, the present invention adopts the following technical solution:

[0007] A method for fatigue analysis and damage management of primary loop piping in nuclear power plants based on digital twins includes the following steps:

[0008] a) Classify and manage transient events in the primary loop of nuclear power plants:

[0009] (a.1) Transients occurring in primary circuit main equipment are defined as Class A transients;

[0010] (a.2) Transients occurring in branch systems that are connected to the main equipment and are not frequently used are defined as Class B transients;

[0011] (a.3) Transients occurring in branch systems that are frequently connected to the main equipment are defined as Class C transients;

[0012] b) Perform fatigue damage analysis for each type of transient;

[0013] (b.1) The Class A transients described herein employ a rapid fatigue assessment based on design transient statistics;

[0014] (b.2) The Class B transients are assessed in real time based on online monitoring of operating parameters;

[0015] (b.3) The C-type transient is assessed using a conservative evaluation based on the cumulative occurrence time of the transient or crack propagation.

[0016] c) Based on digital twin-based optimization of nuclear power plant operation control, real-time monitoring and transient classification management of operating parameters are implemented, and operation control strategies that are conducive to reducing the fatigue damage coefficient (CFU) of primary loop piping in nuclear power plants are proposed.

[0017] Optimally, in step (a.1), the primary loop main equipment includes the reactor pressure vessel, steam generator, primary loop main piping, and main pump. The primary loop design of pressurized water reactor nuclear power plants requires the operating temperature change rate to be controlled within 28°C / h. Under this condition, the impact of primary loop transients on the fatigue damage of the main equipment is limited.

[0018] Optimally, in step (a.2), the branch systems connected to the main equipment and not frequently used include chemical and volume control systems, safety injection systems, waste heat removal systems, etc. During commissioning, the temperature difference changes greatly and the temperature change rate is fast, but the number of such transients is moderate. In this case, it is necessary to monitor the overall operating parameters of the primary loop during operation, and supplement the monitoring of the temperature at the local nozzle position and count the number of times the branch pipeline is commissioned, so as to supplement the input data for the real-time evaluation based on online monitoring of operating parameters in step b.2.

[0019] Ideally, in step (a.3), the commonly used branch systems connected to the main equipment include voltage regulator fluctuation pipes, voltage regulator spray lines, etc. During commissioning, temperature differences are large and the rate of temperature change is rapid, and such transients occur frequently, usually accompanied by significant thermal oscillations. In this situation, fatigue analysis based on monitoring data is often insufficient to meet engineering assessment requirements. It is necessary to supplement this with enhanced in-service monitoring and assessments of hypothetical crack propagation and the integrity of defective equipment, providing supplementary input data for the conservative assessment in step b.3 based on the cumulative transient occurrence time or crack propagation.

[0020] Optimally, in step (b.1), the Class A transients are assessed using rapid fatigue evaluation based on design transient statistics. This involves real-time design transient fatigue evaluation based on the number of real-time transient occurrences and the original design fatigue evaluation report. During operation, only the overall operating parameters of the primary loop need to be monitored. The number of design transient occurrences is counted, and transient occurrence frequency analysis is performed. Ensuring that the actual number of transient occurrences during operation is less than the design limit satisfies the requirements of fatigue aging management. After the design transients are counted from the online transient detection parameters, the transient pairing table in the fatigue analysis of the design report is used to assemble the operational monitoring transients. For cases where paired transients are missing, a conservative analysis is performed by supplementing the original design pairings, thereby calculating the pipeline fatigue damage coefficient (CUF) based on the design report and the number of monitored design transient occurrences.

[0021] Optimally, in step (b.2), the Type B transient stress test employs real-time assessment based on online monitoring of operating parameters to obtain local thermal-hydraulic characteristics of the pipeline (such as fluid pressure, temperature, and flow velocity information) from nuclear power plant instruments. The stress change process at the analysis site is determined using a transfer function (a common technique for online assessment). Then, based on the load cycle statistics procedure in the French RCC-M standard, the load spectrum of the analysis site is compiled, and the fatigue damage coefficient (CUF) of the pipeline is calculated by comparing it with the fatigue performance curves of the material.

[0022] Optimally, in step (b.2), the Class B transient is evaluated in real time based on online monitoring of operating parameters, which divides the loads on the pipeline of concern in the Class B transient into three categories: one is thermal stress caused by thermal transient, one is stress caused by quasi-static loads such as internal pressure, and the last is stress caused by gravity, which can be assumed to be a constant load.

[0023] For loads related to the load history (thermal stress caused by thermal transients), the stress is related not only to the current stress state of the object, but also to the change history of the preceding time parameters, and time-consuming calculations are required based on monitoring data; for loads unrelated to the load history (stress caused by quasi-static loads), the stress response under unit load or reference load can be calculated, and then the stress data at the time of analysis can be obtained through data interpolation; stress caused by gravity, etc., can be regarded as a constant load.

[0024] Optimally, in step (b.3), the C-type transient is assessed using a conservative evaluation analysis based on the cumulative occurrence time of the transient or crack propagation, divided into four levels (corresponding to different evaluation accuracies and workloads, with the evaluator selecting the analysis level based on the evaluation accuracies):

[0025] (i) Screening of potential thermal fatigue risk locations: Based on fatigue damage analysis and calculation of typical pipelines, the minimum thermal fluid temperature difference (the temperature difference between the main loop fluid and the branch pipe thermal fluid) that can lead to fatigue damage at the T-joint is obtained, and this is used as a general "qualification threshold". For pipelines whose theoretically maximum operating thermal fluid temperature difference is less than the "qualification threshold", further assessment of thermal fatigue damage is not required. The "qualification threshold" value for thermal fluid temperature difference of austenitic steel pipelines in the primary loop of nuclear power plants is set at 80℃, and the "qualification threshold" value for thermal fluid temperature difference of carbon steel pipelines is set at 50℃.

[0026] (ii) Engineering assessment based on conservative assumptions: It is generally assumed that the transient thermal fluid temperature changes in a sinusoidal manner. The fatigue damage engineering assessment of the structural response characteristics is carried out with reference to the French RCC-M standard. The analysis obtains the thermal shock frequency under the condition of maximum structural fatigue damage (the thermal shock frequency does not necessarily mean that the higher the frequency, the greater the damage; excessively high frequencies may not cause fatigue damage). In this step, only the influence of thermal shock load is considered, and the influence of structural factors can be ignored.

[0027] (iii) Based on the analysis of the monitored transient parameters, temperature measuring points are added at local locations in the pipeline, and detailed structural stress response finite element numerical simulation analysis is performed based on the monitored transient parameters.

[0028] (iv) Evaluation based on crack propagation characteristics: Since the stress caused by thermal shock is mainly concentrated on the pipe surface, the pipe stress will be significantly reduced after the crack propagates 1-2 mm away from the wall surface, and the thermal fatigue crack will stop propagating after it has reached a certain extent.

[0029] Ideally, crack propagation analysis should consider the effects of alternating loads (fatigue crack propagation) and the primary circuit water environment (stress corrosion crack propagation) in accordance with the French RCC-M standard.

[0030] Optimally, in step (b.3), the conservative assessment analysis of the C-type transient based on the cumulative occurrence time of the transient or crack propagation requires the addition of local temperature measurement points in the pipeline to assist in determining the timing and frequency of thermal shock transients. When a thermal shock is identified, it is necessary to assume that a transient process with the maximum thermal shock temperature difference has occurred.

[0031] Optimally, in step (b.3), the C-type transient adopts a conservative assessment analysis based on the cumulative occurrence time of the transient or crack propagation, and the crack propagation characteristics are calculated using the Paris formula (corresponding to the assessment based on crack propagation characteristics in (iv) above). The assessment based on crack propagation characteristics can significantly extend the service life of the component and reduce the cycle requirements of in-service inspection.

[0032] In step (b.3), the conservative assessment analysis of the C-type transient based on the cumulative occurrence time of the transient or crack propagation needs to refer to industry experience feedback, especially the experience feedback of replacement of components caused by thermal fatigue cracking in similar power plants. For replacement events caused by thermal fatigue cracking, a detailed assessment based on crack propagation characteristics is required for the corresponding components.

[0033] In an optimized manner, step (c) divides the software processing aspect of the nuclear power plant primary loop piping in the digital twin into four platforms: a data acquisition platform (used to collect power plant monitoring data in steps a and b), a unified management platform (to realize the identification and processing of bad points in the data acquisition platform), a digital twin platform (to reproduce the analysis focus in steps a and b), and a comprehensive business platform (to realize CUF calculation, analysis, and management in steps a and b).

[0034] Ideally, in step (c), the data acquisition platform includes defect identification and processing; the unified management platform includes optimization processing of data requiring digital twin submission; the digital twin platform includes real-time display of fatigue damage; and the integrated business platform includes suggestions for optimization strategies for transient operation.

[0035] In step (c), the data transfer between four platforms in the software processing of the primary loop piping of the nuclear power plant in the digital twin is optimized by accumulating and comparing the characteristics of different operations during each operation of the nuclear power plant, comparative experience learning is carried out, and good operating practices with small impact on the pipeline fatigue damage coefficient CUF are selected as the recommended operating methods for the subsequent operation, so as to reduce the fatigue damage of the primary loop piping of the nuclear power plant during long-term operation.

[0036] Due to the application of the above technical solutions, the present invention has the following advantages compared with the prior art: The present invention provides a fatigue analysis and damage management method for primary loop pipelines in nuclear power plants based on digital twins, which overcomes the technical difficulties of unified management and quantitative evaluation of fatigue damage in different transient states of nuclear power plants. It proposes a unified management method for fatigue damage based on online monitoring of transient parameters, and provides an operation control strategy to reduce fatigue damage in primary loop pipelines of nuclear power plants, thus providing technical support for the long-term safe operation of primary loop pipelines in nuclear power plants. Attached Figure Description

[0037] To more clearly illustrate the technical solutions in the embodiments of the present invention, the accompanying drawings used in the description of the embodiments will be briefly introduced below. Obviously, the accompanying drawings described below are only some embodiments of the present invention. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0038] Figure 1 This is a flowchart of the analysis method of the present invention;

[0039] Figure 2 This invention relates to a transient classification and management strategy for the primary loop of nuclear power plants.

[0040] Figure 3 This is a flowchart of the real-time evaluation process based on online monitoring of operating parameters according to the present invention;

[0041] Figure 4 This invention relates to a local temperature monitoring instrument for the primary loop piping of a nuclear power plant;

[0042] Figure 5 This is a schematic diagram illustrating the online monitoring of operating parameters of the reactor primary loop piping according to the present invention;

[0043] Figure 6 This is a schematic diagram of the stress results calculated by online monitoring of the operating parameters of the reactor primary loop piping according to the present invention;

[0044] Figure 7 This is a schematic diagram of the hot and cold water mixing area of ​​the pipe joint of the present invention. Detailed Implementation

[0045] To enable those skilled in the art to better understand the technical solutions of the present invention, the technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort should fall within the scope of protection of the present invention.

[0046] This invention proposes a fatigue analysis and damage management method for primary loop pipelines in nuclear power plants based on digital twins. It overcomes the technical difficulties of unified management and quantitative evaluation of fatigue damage in different transient states of nuclear power plants. It also proposes a unified management method for fatigue damage based on online monitoring of transient parameters and provides an operation control strategy to reduce fatigue damage in primary loop pipelines of nuclear power plants, thus providing technical support for the long-term safe operation of primary loop pipelines in nuclear power plants.

[0047] like Figure 1 As shown in this embodiment, the fatigue analysis and damage management method for primary loop piping in a nuclear power plant based on digital twins includes the following steps:

[0048] 1) such as Figure 1 , 2 As shown, transients in the primary loop of a nuclear power plant are classified and managed accordingly:

[0049] (1.1) Define Class A transient

[0050] Type A transient refers to the requirement in the design of a pressurized water reactor nuclear power plant that the rate of temperature change during operation be controlled within 28°C / h for primary loop main equipment (such as reactor pressure vessel, steam generator, primary loop main piping, etc.).

[0051] In this case, the impact of transients in the primary circuit on the fatigue damage of the main equipment is limited. During operation, it is only necessary to focus on the overall operating parameters of the primary circuit, count the number of designed transients, and conduct transient occurrence analysis to ensure that the actual number of transients during operation is less than the design limit, which can meet the requirements of fatigue aging management.

[0052] (1.2) Define Class B transient

[0053] Type B transients refer to those infrequently connected branch systems (such as chemical and volume control systems, safety injection systems, and waste heat removal systems) that experience large temperature differences and rapid temperature changes during commissioning, but occur in a moderate number of times.

[0054] In this case, it is necessary to monitor the overall operating parameters of the primary circuit during operation, and supplement the monitoring of the temperature at the local nozzle location and count the number of times the branch pipeline is put into operation.

[0055] (1.3) Define Class C transient

[0056] Type C transients refer to common branch systems connected to the main equipment (such as voltage regulator oscillation tubes, voltage regulator spray lines, etc.): during commissioning, the temperature difference changes greatly, the rate of temperature change is fast, and such transients occur frequently, usually accompanied by obvious thermal oscillation processes.

[0057] In this context, fatigue analysis based on monitoring data is often insufficient to meet engineering assessment requirements, necessitating enhanced in-service monitoring and assessments of hypothetical crack propagation and the integrity of defective equipment. Crack propagation analysis must consider the effects of alternating loads (fatigue crack propagation) and the primary circuit water environment (stress corrosion crack propagation).

[0058] 2) such as Figure 2 As shown, fatigue analysis and damage management are performed for each type of transient.

[0059] (2.1) For Class A transients, fatigue rapid assessment based on design transient statistics is adopted.

[0060] Type A transients employ a rapid fatigue assessment based on design transient statistics, which evaluates the design transient fatigue damage coefficient (CUF) in real time based on the number of real-time transient occurrences and the original design fatigue evaluation report.

[0061] After the design transients are statistically determined from the transient online detection parameters, the transient pairing table of fatigue analysis in the design report is used to match the operational monitoring transients. For cases where there are missing paired transients, a conservative analysis is performed by supplementing the original design pairings, thereby calculating the pipeline fatigue damage coefficient CUF based on the design report and the number of monitoring design transients.

[0062] (2.2) Type B transients are evaluated in real time based on online monitoring of operating parameters.

[0063] like Figure 3 As shown, the Class B transient analysis uses real-time evaluation based on online monitoring of operating parameters to obtain local thermal-hydraulic characteristics of the pipeline (such as fluid pressure, temperature, flow velocity, etc.) from nuclear power plant instruments, and determines the stress change process of the analysis part through transfer function.

[0064] Based on the load cycle statistics procedure in the French RCC-M standard, the load spectrum of the analysis part is sorted out, and then the fatigue damage coefficient CUF of the pipeline is calculated by comparing it with the fatigue performance curve of the material.

[0065] like Figure 4 As shown, in Type B transient analysis, it is necessary to add temperature measuring points at local locations on the pipeline. The transient information monitored for a certain pipeline is as follows: Figure 5 As shown, the calculated stress results are as follows: Figure 6 As shown.

[0066] In the real-time assessment based on online monitoring of operating parameters, Class B transient loads are divided into three categories: thermal stress caused by thermal transients, stress caused by quasi-static loads such as internal pressure, and stress caused by gravity, which can be assumed to be constant loads.

[0067] For loads related to the load history (thermal stress caused by thermal transients), the stress is related not only to the current stress state of the object but also to the change history of the preceding time parameters, requiring time-consuming calculations based on monitoring data; for loads unrelated to the load history (stress caused by quasi-static loads such as internal pressure), the stress response under a unit load or reference load can be calculated, and then the stress data at the analysis time can be obtained through data interpolation; stress caused by gravity, etc., can be regarded as a constant load.

[0068] (2.3) Class C transients are assessed using a conservative evaluation based on the cumulative occurrence time of the transient or crack propagation.

[0069] Type C transient analysis employs a conservative assessment based on the cumulative occurrence time of the transient or crack propagation, divided into four levels (corresponding to different assessment accuracies and workloads, with the assessor selecting the analysis method according to assessment requirements):

[0070] (i) Screening of potential thermal fatigue risk locations: Based on fatigue damage analysis and calculation of typical pipelines, the maximum hot fluid temperature difference (the temperature difference between the main loop fluid and the branch pipe hot fluid) that can cause fatigue damage at the T-joint is obtained, and this is used as a general "qualification threshold". For pipelines whose theoretically maximum operating hot fluid temperature difference is less than the "qualification threshold", no further assessment of thermal fatigue damage is required. In this application, the "qualification threshold" value for the hot fluid temperature difference of the primary loop austenitic steel pipeline in a nuclear power plant is set at 80°C, and the "qualification threshold" value for the hot fluid temperature difference of the carbon steel pipeline is set at 50°C.

[0071] (ii) Engineering assessment based on conservative assumptions: This typically assumes that the transient thermal fluid temperature changes sinusoidally. A fatigue damage engineering assessment of the structural response characteristics is then performed, analyzing the thermal shock frequency that maximizes structural fatigue damage (higher thermal shock frequencies do not necessarily lead to greater damage; excessively high frequencies may not cause fatigue damage). This step only considers the effect of thermal shock loads and ignores the influence of structural factors.

[0072] (iii) Based on the analysis of the monitored transient parameters, temperature measuring points are added at local locations in the pipeline to conduct a detailed structural stress response analysis based on the monitored transient parameters.

[0073] (iv) Evaluation based on crack propagation characteristics: Since the stress caused by thermal shock is mainly concentrated on the pipe surface, the pipe stress will be significantly reduced after the crack propagates 1-2 mm away from the wall surface, and the thermal fatigue crack will stop propagating after it has reached a certain extent.

[0074] like Figure 7 As shown, there is a certain degree of uncertainty in the hot and cold water mixing region. Therefore, in the conservative assessment analysis of Class C transients based on the cumulative occurrence time of transients or crack propagation, the Paris formula is used to calculate the crack propagation characteristics. The assessment based on crack propagation characteristics can significantly extend the service life of components and reduce the cycle requirements for in-service inspections.

[0075] In the conservative assessment analysis of Class C transients based on the cumulative occurrence time of the transient or crack propagation, it is also necessary to refer to industry experience feedback, especially the experience feedback of replacement of components that have experienced thermal fatigue cracking in similar power plants.

[0076] 3) Optimization of nuclear power plant operation control based on digital twins: Based on real-time monitoring and transient classification management of operating parameters, an operation control strategy is proposed that is conducive to reducing fatigue damage of primary loop pipelines in nuclear power plants.

[0077] By accumulating experience from the operation of nuclear power plants and comparing the characteristics of different operations during each operation, comparative learning is conducted to select good operating practices that have a small impact on the pipeline fatigue damage coefficient (CUF) as recommended operating methods for the future, thereby reducing fatigue damage to the primary loop pipelines of nuclear power plants during long-term operation.

[0078] As shown in Table 1, the software processing of the primary loop piping in a nuclear power plant using digital twins is divided into four platforms: a data acquisition platform, a unified management platform, a digital twin platform, and a comprehensive business platform. The data acquisition platform includes defect identification and processing; the unified management platform includes optimization processing of data requiring digital twin submission; the digital twin platform includes real-time display of fatigue damage; and the comprehensive business platform includes suggestions for optimization strategies during transient operation.

[0079] Table 1. Digital Twin Software Processing Platform for Nuclear Power Plant Primary Circuit Piping

[0080]

[0081]

[0082] In the software processing of the primary circuit piping of the nuclear power plant in digital twins, there is mutual data transfer between four platforms. By accumulating and comparing the characteristics of different operations during each operation of the nuclear power plant, comparative experience learning is carried out to select the best operating practices that minimize pipeline fatigue damage as the recommended operating methods for subsequent operation, so as to reduce fatigue damage to the primary circuit piping of the nuclear power plant during long-term operation.

[0083] Example 1

[0084] The primary loop piping of a nuclear power plant is made of austenitic stainless steel, material grade Z2CND18-12N2. The pipe's inner diameter R... i =142.1mm, outer diameter R3 =177.8mm. The specific implementation process of this invention's method for fatigue analysis and damage management of primary loop piping in nuclear power plants based on digital twins is as follows:

[0085] 1) such as Figure 2 As shown, transients in the primary loop of a nuclear power plant are classified and managed accordingly:

[0086] (1.1) Low frequency transients, Class A transients include: nuclear power plant primary loop start-up, shutdown, cold hydrostatic test, hot hydrostatic test, etc.

[0087] (1.2) Transients with moderate frequency of occurrence. Type B transients include: cold injection transients, hot injection transients, charging transients, draining transients, and residual heat discharge transients.

[0088] (1.3) High frequency transients, Class C transients include transients such as thermal stratification of horizontal pipe sections of voltage regulator fluctuation pipe, spraying of spray pipeline, and thermal oscillation of valve leakage.

[0089] 2) Perform fatigue analysis and damage management for each type of transient event;

[0090] (2.1) For Class A transients, fatigue rapid assessment based on design transient statistics is adopted.

[0091] In this case study, the fatigue evaluation results from the pipeline design phase are shown in Table 2. The fatigue damage factor (CUF) of the pipeline in the design is 0.2197. After service operation, the actual transient types and quantities identified are shown in Table 3. In the rapid fatigue assessment based on design transient statistics, real-time design transient fatigue evaluation can be performed based on the real-time transient occurrence quantity and the original design fatigue evaluation report. As shown in Table 4, the actual transients are matched according to the pairing table from the design phase. For cases where paired transients are missing, a conservative analysis is performed by supplementing the original design pairings. In this case, the calculated CUF caused by pipeline design transients is 0.001108.

[0092] Table 2 Load spectrum statistics during the design phase

[0093]

[0094] Table 3 shows the transients identified by the TAC module in the case studies.

[0095]

[0096] Table 4. Calculation results of the rapid fatigue assessment case based on design transient statistics.

[0097]

[0098] (2.2) Class B transients are assessed in real-time based on online monitoring of operating parameters. Transient information monitored for a certain pipeline is as follows: Figure 5 As shown. The stress change process of type B transient is calculated in real time based on online monitoring of operating parameters, as follows: Figure 6 As shown, the CUF caused by the transient is less than 0.00001.

[0099] (2.3) Class C transient tests employ a conservative assessment based on the cumulative occurrence time of the transient or crack propagation. In the Class C transient test analysis using a conservative assessment based on the cumulative occurrence time of the transient or crack propagation, the Paris formula is used to calculate crack propagation characteristics. Assessment based on crack propagation characteristics can significantly extend the service life of components and reduce the frequency of in-service inspections. For example, for a nuclear power plant's safety injection pipeline, the crack propagation amount calculated based on the Paris formula can reach approximately 3 mm after every 500 hours of operation. Therefore, it is recommended to add an in-service defect size inspection every 500 hours of operation for this pipeline to ensure the safety performance of the pipeline in service.

[0100] 3) Optimization of nuclear power plant operation control based on digital twins.

[0101] If the CUF formed by the transient F1, F2 and F3 of the safety injection pipeline is minimized after calculation, the relevant operations of the safety injection transient F1 can be optimized, which can effectively reduce the fatigue damage of the primary loop pipeline in the nuclear power plant.

[0102] The above embodiments are only for illustrating the technical concept and features of the present invention, and are intended to enable those skilled in the art to understand the content of the present invention and implement it accordingly. They should not be construed as limiting the scope of protection of the present invention. All equivalent changes or modifications made in accordance with the spirit and essence of the present invention should be covered within the scope of protection of the present invention.

Claims

1. A method for fatigue analysis and damage management of primary loop piping in nuclear power plants based on digital twins, characterized in that, Includes the following steps: a) Classify and manage transient events in the primary loop of nuclear power plants: (a.1) Transients occurring in primary circuit main equipment are defined as Class A transients; (a.2) Transients occurring in branch systems that are connected to the main equipment and are not frequently used are defined as Class B transients; (a.3) Transients occurring in branch systems that are frequently connected to the main equipment are defined as Class C transients; b) Perform fatigue analysis for each type of transient; (b.1) The Class A transients described herein employ a rapid fatigue assessment based on design transient statistics; (b.2) The Class B transients are assessed in real time based on online monitoring of operating parameters; (b.3) The C-type transient is assessed using a conservative evaluation based on the cumulative occurrence time of the transient or crack propagation. The conservative assessment analysis based on transient cumulative occurrence time or crack propagation described in step (b.3) is divided into four levels: (i) Screening of potential thermal fatigue risk locations: Based on fatigue damage analysis and calculation of typical pipelines, the minimum thermal fluid temperature difference that can form fatigue damage at T-joints is obtained and used as the identification threshold; for pipelines whose theoretically maximum operating thermal fluid temperature difference is less than the identification threshold, no further evaluation of the thermal fatigue damage coefficient (CUF) is required. (ii) Engineering assessment based on conservative assumptions: Assuming that the transient thermal fluid temperature changes in a sinusoidal manner, the fatigue damage engineering assessment of the structural response characteristics is carried out with reference to the French RCC-M standard, and the thermal shock frequency under the condition of maximum structural fatigue damage can be obtained by analysis. (iii) Based on the analysis of monitored transient parameters, temperature measuring points are added at local locations in the pipeline, and finite element numerical simulation structural stress response analysis is performed based on the monitored transient parameters. (iv) Evaluation based on crack propagation characteristics: The stress caused by thermal shock is concentrated on the pipe surface. After the crack propagates away from the wall surface by 1-2 mm, the thermal stress of the pipe will decrease, and the thermal fatigue crack will stop propagating after it has reached a certain extent. c) Real-time monitoring of nuclear power plant operating parameters and classification management of transient damage based on digital twins.

2. The fatigue analysis and damage management method according to claim 1, characterized in that: The primary loop main equipment includes the reactor pressure vessel, steam generator, primary loop main piping, and main pump.

3. The fatigue analysis and damage management method according to claim 1, characterized in that: The branch systems that are connected to the main equipment but are not frequently used include the chemical and volume control system, the safety injection system, and the waste heat removal system.

4. The fatigue analysis and damage management method according to claim 1, characterized in that: The commonly used branch system connected to the main equipment includes voltage regulator fluctuation pipes and voltage regulator spray lines.

5. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.1), the rapid fatigue assessment based on design transient statistics is to perform real-time design transient fatigue assessment based on the number of real-time transient occurrences and the original design fatigue evaluation report: after the design transients that occur are statistically identified by the online transient detection parameters, the transient pairing table of the fatigue analysis in the original design fatigue evaluation report is used to match the operational monitoring transients. For cases where there are missing paired transients, a conservative analysis is performed by supplementing the original design pairings, thereby calculating the pipeline fatigue damage coefficient CUF based on the design report and the number of monitoring design transient occurrences.

6. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.2), the real-time assessment based on online monitoring of operating parameters involves obtaining the local thermal-hydraulic characteristics of the pipeline from the nuclear power plant instruments and determining the stress change process of the analysis location through the transfer function. Based on the statistical procedure for load cycles, the load spectrum of the analysis area is sorted out, and then the fatigue damage coefficient (CUF) of the pipeline is calculated by comparing it with the fatigue performance curve of the material.

7. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.2), the loads borne by the pipes of concern in the transient B category are divided into thermal stress caused by thermal transients, stress caused by quasi-static loads, and stress that can be assumed to be constant loads.

8. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.2), the real-time evaluation based on online monitoring of operating parameters needs to include local temperature measurement points in the pipeline to accurately analyze fatigue damage in the B-type transient state.

9. The fatigue analysis and damage management method according to claim 1, characterized in that: In the conservative assessment analysis based on the cumulative occurrence time of transients or crack propagation described in step (b.3), it is necessary to add local temperature measurement points in the pipeline to help determine the time and number of thermal shock transients.

10. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.3), the conservative assessment analysis based on transient cumulative occurrence time or crack propagation is adopted. The crack propagation characteristics are calculated using the Paris formula. The assessment based on crack propagation characteristics can significantly extend the service life of the component and reduce the cycle requirements of in-service inspection.

11. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (b.3), the crack propagation analysis needs to consider fatigue crack propagation and stress corrosion crack propagation.

12. The fatigue analysis and damage management method according to claim 1, characterized in that: In step (c), the software processing aspect of the nuclear power plant primary loop pipeline in the digital twin is divided into four platforms: a data acquisition platform for collecting power plant monitoring data in steps a and b; a unified management platform for identifying and processing bad points in the data acquisition platform; a digital twin platform for reproducing the analysis focus objects in steps a and b; and a comprehensive business platform for realizing CUF calculation, analysis and management in steps a and b.

13. The fatigue analysis and damage management method according to claim 12, characterized in that: In step (c), based on digital twins, the characteristics of different operations during each operation are accumulated and compared through the operation of nuclear power plants. Comparative experience learning is carried out to select good operating practices that minimize pipeline fatigue damage as recommended operating methods for the future, thereby achieving the goal of reducing the fatigue damage coefficient (CUF) of the primary loop pipelines in nuclear power plants during long-term operation.

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