A method for evaluating the structural integrity of a primary loop pipe of a nuclear power plant based on online monitoring of transient parameters

Abstract

The application relates to a kind of online monitoring based on transient parameter evaluation method of nuclear power plant primary loop main pipeline structural integrity, comprising the following steps: a) initial data processing, (a.1) acquisition online monitoring data, and screening is obtained common data, and common data is uniformly processed;(a.2) after the data of uniform processing, optimization is carried out using parallel computing;B) crack initiation evaluation, (b.1) based on the data after parallel computing, damage assessment is carried out using thermal oscillation load in the form of sinusoidal loading, to obtain maximum fatigue damage;(b.2) based on the influence of nuclear power plant primary loop water environment, maximum fatigue damage is corrected;C) crack failure evaluation, if the maximum fatigue damage after correction is greater than 1, or the case where accurate fatigue evaluation cannot be carried out, a fracture-plastic instability double failure mode is used to evaluate the crack integrity, to determine whether the nuclear power plant primary loop main pipeline structure is complete.

Description

Technical Field

[0001] This invention belongs to the field of structural integrity assessment technology, specifically relating to a method for assessing the structural integrity of the primary loop main pipeline of a nuclear power plant based on online monitoring of transient parameters, providing accurate technical basis for the safety assessment of key nuclear power equipment. Background Technology

[0002] Fatigue aging management of the primary loop pressure-bearing piping in pressurized water reactor nuclear power plants is a crucial aspect of ensuring the integrity of the pressure-bearing boundary. In recent years, multiple fatigue cracks have been discovered at the mixing primary loop tees in nuclear power plants. In 2022, during routine safety inspections at nuclear power plants such as Chivo in France, several cracks were found near the welds of reactor injection pipelines. Fatigue cracks typically cause small cracks or leaks and generally do not lead to direct pipe fracture. However, from the perspective of long-term economic operation, it is necessary to fully manage the fatigue issues of metal components. The mixing of fluids at different temperatures causes high-frequency temperature fluctuations (thermal oscillations) in the pipe wall at the tees, which easily leads to the initiation of fatigue cracks in these areas. The degree of fatigue damage at the tees caused by mixed flow depends not only on the temperature difference of the mixed fluids but also on the geometry of the pipe and the flow conditions.

[0003] For low-frequency thermal oscillations, the temperature of the outer wall of the pipe at the mixing tee can be used as input for fatigue damage assessment. For high-frequency thermal oscillations, the unit distribution characteristics of temperature fluctuations need to be obtained through numerical simulation or multi-point temperature measurement. Since the spatial distribution of pipe wall temperature fluctuations at the mixing tee is not uniform due to the influence of pipe and flow conditions, it is necessary to identify the location with the largest temperature difference for fatigue damage assessment.

[0004] Currently, research institutions worldwide have developed various fatigue analysis systems based on online monitoring of transient parameters, such as the FatiguePro fatigue analysis system developed by the Nuclear Power Research Institute in the United States, the FAMOS fatigue analysis system developed by Areva in Germany, and the WESTEMSTM fatigue analysis system developed by Westinghouse. However, these publicly available analysis systems currently lack the capability to perform high-cycle fatigue analysis based on thermal oscillations and structural integrity analysis based on fracture mechanics. Summary of the Invention

[0005] The purpose of this invention is to provide a method for assessing the structural integrity of the primary loop main pipeline of a nuclear power plant based on online monitoring of transient parameters, in order to overcome the shortcomings of the prior art.

[0006] To achieve the above objectives, the technical solution adopted by the present invention is as follows:

[0007] A method for assessing the structural integrity of the primary loop main piping in a nuclear power plant based on online monitoring of transient parameters includes the following steps:

[0008] a) Big data processing technologies in the initial data processing layer:

[0009] (a.1) Collect online monitoring data, filter out common data, and process the common data in a unified manner;

[0010] (a.2) Optimize the data after unified processing using parallel computing;

[0011] b) Evaluation of crack initiation

[0012] (b.1) Based on the data after parallel computing, damage assessment is performed using a sinusoidal thermal oscillation load to obtain the maximum fatigue damage;

[0013] (b.2) Correction of maximum fatigue damage based on the influence of the primary loop water environment of nuclear power plants;

[0014] c) Crack failure evaluation

[0015] If the maximum fatigue damage after correction is greater than 1, or if an accurate fatigue evaluation cannot be performed, the fracture-plastic instability dual failure mode is used to assess the crack integrity and determine whether the primary loop main pipeline structure of the nuclear power plant is intact.

[0016] Optimally, in step (a.1), the unified processing refers to processing the calculation process parameters based on online monitoring data according to the calculation path i(P) i,path ), filter and extract the common data (D) required for the cumulative integration algorithm along the time axis. i,share The system performs unified time step size (Δt) calculations and influence function normalization on the shared data. A series of standardized data preprocessing steps are performed, and the preprocessed data is then added to a standard data bus for subsequent processing based on monitoring parameter D. i,monior Calculate the target parameter G i,path Formula (1) provides the application method.

[0017]

[0018] Ideally, in step (a.1), the influence function in the unified processing flow of shared data is normalized. It refers to the response process of a primary loop pipeline under a unit load, which reflects the comprehensive characteristics of structure, load and material properties.

[0019] Ideally, in step (a.1), for primary loop pipelines with localized thermal shock and thermal oscillation transient loads, it is necessary to add local temperature monitoring instruments on the outer surface of the pipeline to obtain more online local monitoring data.

[0020] Optimally, in step (a.2), the parallel computation of the data is as follows: the existing algorithm for accumulating integrals along the time axis is optimized to perform the computation within the specified decay time (t) of the influence function. d At any point j within the range, the target time series range (t) can be obtained by accumulating integrals simultaneously. j,start , t j,end Algorithms with arbitrary parameters within the target parameter G. i,path The parallel computing method is shown in the following equation:

[0021]

[0022] Optimally, in step (a.2), the data parallel computing enables the online monitoring data based on transient parameters to be divided into several subsets, thereby achieving on-demand data allocation, segmented submission, and time-sharing parallel computing. The subsets are processed on different algorithm computing nodes, improving the computational efficiency of the algorithm. A distributed computing system composed of multiple computing nodes can fully utilize computing resources to achieve efficient processing of big data.

[0023] Optimally, in step (b.1), to supplement step (a) where it is impossible to monitor parameters (D) based on the outer surface of the pipe. i,monior The accurately predicted thermal oscillation load (i.e., thermal shock load) refers to the rapid damage assessment of the conservative loading in the form of sinusoidal thermal oscillation, assuming that the pipeline is subjected to the maximum alternating thermal shock load in the analysis (the temperature fluctuation of the fluid inside the pipeline is taken as the maximum temperature difference between the hot and cold three-dimensional structures). The definition method is shown in the following formula.

[0024]

[0025] In the formula, T fluid-ave ΔT represents the average fluid temperature (°C) in the thermal shock region. fluid f represents the amplitude (°C) of fluid temperature change in the thermal shock region. i Let t be the assumed frequency of thermal shock change, and t be time (s). The directional angle is rad.

[0026] Optimally, in step (b.1), during damage assessment using sinusoidal thermal oscillation loads, sensitivity analysis is used to obtain different thermal shock frequencies f. i The structural stress response is analyzed, and based on this, structural fatigue damage analysis is performed, thereby identifying the impact frequency f that has the greatest impact on structural fatigue damage. max The corresponding maximum fatigue damage is UF max .

[0027] Optimally, in step (b.2), to supplement the consideration of the primary circuit water environment of the nuclear power plant on UFmax The promoting effect, the correction of the maximum fatigue damage based on the influence of the primary loop water environment of the nuclear power plant is as follows: For extreme point i, record its previous extreme point as i-1 and its next extreme point as i+1, and calculate the environmental impact factor F during the load history period (i-1, i) based on the following formula. en(i-1,i) Based on formula (4), the environmental impact factor F during the load history period (i, i+1) can also be calculated. en(i,i+1) .

[0028]

[0029] In the formula, k is the number of time points during the load history period (i-1, i) where transient operating parameters are recorded, j is the j-th recorded load fluctuation, and Δε j This refers to the strain change during a single fluctuation process.

[0030] Optimally, in step (b.2), the correction of the maximum fatigue damage based on the influence of the primary loop water environment of the nuclear power plant is as follows: for the extreme point i, it is recorded in the load spectrum record database as (i, F en(i-1,i) F en(i,i+1) The specific time history of the load spectrum is no longer recorded, which significantly reduces the amount of data stored in the database.

[0031] Optimally, in step (b.2), the correction of the maximum fatigue damage based on the influence of the primary loop water environment of the nuclear power plant is as follows: In the revision of fatigue damage, for the re-interference with the limit point (i, F) en(i-1,i) F en(i,i+1) Any extreme point (q, F) paired en(q-1,q) F en(q,q+1) The environmental impact factor F of the primary loop of the nuclear power plant, which is re-paired (i, q). en(i,q) The calculation method is as follows.

[0032] F en(i,q) =max(F en(i-1,i) ,F en(i,i+1) ,F en(q-1,q) ,F en(q,q+1) (5)

[0033] Ideally, in step (b.2), the impact of the primary circuit water environment of the nuclear power plant on UF should be supplemented. max After the promoting effect, the final fatigue damage of the modified structure [UF] max ] end The calculation method is shown in equation (6). Standard requirements [UF] max ] end If the value is less than 1, then it is necessary to assume that the pipe has developed a crack.

[0034] [UF max ]end =UF max ·F en(i,q) (6)

[0035] Optimally, in step (c), for [UF max ] end In cases where the value exceeds the specification limit by 1, or where accurate fatigue evaluation cannot be performed, the crack integrity assessment using the fracture-plastic instability dual failure mode is as follows: based on the safety assessment on the failure assessment diagram (both fracture and plastic instability are considered structural failures), the failure assessment diagram of this invention ignores the influence of pressure loads, which are not the main load factors causing the failure of the primary loop main pipeline, and the technical development and application research mainly focus on the content of bending moment loads.

[0036] In an optimized manner, in step (c), the assessment of crack integrity using the fracture-plastic instability dual failure mode is as follows: the fracture assessment based on the failure assessment diagram is based on the J integral and is a conservative assessment curve obtained through a large number of finite element numerical simulations. The assessment curve assessment equation is shown below.

[0037]

[0038] In the formula, L r K is the x-axis of the evaluation curve. r To determine the ordinate of the curve.

[0039] Optimally, in step (d), the maximum value of the abscissa of the failure assessment diagram in the crack integrity assessment using the fracture-plastic instability dual failure mode is... Let it be 1.3.

[0040] Ideally, in step (d), when evaluating crack integrity using the fracture-plastic instability dual failure mode, the evaluation point falls on the evaluation curve, the coordinate axis, and the maximum value of the abscissa. If the internal structure meets the specifications, the structure of the primary loop main pipeline of a nuclear power plant is complete and safe; otherwise, the structure is incomplete and unsafe.

[0041] Due to the application of the above technical solutions, this invention has the following advantages compared with existing technologies: This invention proposes a method for assessing the structural integrity of the primary loop main pipeline of a nuclear power plant based on online monitoring of transient parameters. This overcomes the technical difficulty of rapid and accurate operation of big data in online monitoring and analysis of transient parameters in nuclear power plants. It proposes a rapid damage assessment technique for primary loop main pipeline structures subject to thermal oscillation loads and provides a convenient algorithm considering the influence of the primary loop water environment of the nuclear power plant. It also enables convenient storage of data in the form of load spectra under different refueling cycles. Furthermore, this invention provides a crack integrity assessment technique for the unified assessment of the fracture-plastic instability dual failure modes of the primary loop main pipeline of a nuclear power plant, providing in-depth technical support for the long-term safe operation of the primary loop pipeline of a nuclear power plant. Attached Figure Description

[0042] 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.

[0043] Figure 1 This is a flowchart illustrating the method for assessing the structural integrity of the primary loop main pipeline in a nuclear power plant based on online monitoring of transient parameters, as per the present invention.

[0044] Figure 2 This is a schematic diagram of the real-time evaluation process based on online monitoring of operating parameters according to the present invention;

[0045] Figure 3 This is a schematic diagram of the online monitoring instrument for the operating parameters of the primary loop pipeline in a nuclear power plant in this invention;

[0046] Figure 4 This is a schematic diagram of the big data cluster computing based on online monitoring of transient parameters in this invention;

[0047] Figure 5 This is a schematic diagram illustrating the evaluation of the dual failure criteria for the primary loop main pipeline of a nuclear power plant in this invention.

[0048] Figure 6 This is an example of the transient operating parameters obtained from online monitoring of the primary loop piping operating parameters in a nuclear power plant in this invention;

[0049] Figure 7 This is an example of the calculation results of the stress in the online monitoring of transient parameters of the primary loop piping in a nuclear power plant in this invention;

[0050] Figure 8 This is a schematic diagram of the circumferential crack on the inner surface of the primary loop main pipe in a nuclear power plant according to the present invention.

[0051] Figure 9 This is a schematic diagram illustrating the application of the dual failure criteria assessment for the primary loop main pipeline of a nuclear power plant in this invention. Detailed Implementation

[0052] 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.

[0053] like Figure 1 As shown in this embodiment, the method for assessing the structural integrity of the primary loop main piping of a nuclear power plant based on online monitoring of transient parameters includes the following steps:

[0054] 1) The structural integrity assessment process for the primary loop main piping of a nuclear power plant based on online monitoring of transient parameters is as follows: Figure 2 As shown ( Figure 1 This is the key technical point of the present invention. Figure 2 As a key node in the online assessment system process, big data processing technologies are used in the initial data processing layer.

[0055] (1.1) Collect online monitoring data, filter out common data, and process the common data in a unified manner.

[0056] Specifically, the calculation process parameters based on online monitoring data are calculated according to the calculation path i(P) i,path ), filter and extract the common data (D) required for the cumulative integration algorithm along the time axis. i,share Furthermore, the time step size (Δt) and influence function normalization of shared data were standardized. After undergoing a series of standardization processes, it is added to the standard data bus for subsequent use based on monitoring parameter D. i,monior Calculate the target parameter G i,path Formula (1) provides the application method.

[0057]

[0058] like Figure 3 As shown, for pipelines where localized thermal shock occurs, it is necessary to add local monitoring instruments for the temperature of the pipeline's outer surface.

[0059] Normalization of influence functions in a unified data processing workflow It refers to the response process of a primary loop pipeline under a unit load, which reflects the comprehensive characteristics of structure, load and material properties.

[0060] (1.2) Big Data Cluster Computing Technology: Optimizing the data after unified processing using parallel computing.

[0061] The computing cluster code is shown below. Figure 4 As shown, data parallel computing technology refers to optimizing the existing algorithm for accumulating integrals along the time axis to achieve the desired result within a specified decay time (t) of the influence function. d At any point j within the range, the target time series range (t) can be obtained by accumulating integrals simultaneously. j,start , t j,end Algorithms with arbitrary parameters within the target parameter G. i,path The parallel computing method is shown in the following equation:

[0062]

[0063] Data-parallel computing technology enables the partitioning of online monitoring data based on transient parameters into several subsets, thereby achieving on-demand data allocation, segmented submission, and time-sharing parallel computing. Each subset is processed on a different algorithm computing node, improving the algorithm's computational efficiency. A distributed computing system composed of multiple computing nodes can fully utilize computing resources to achieve efficient processing of large datasets.

[0064] 2) A precise crack initiation evaluation method developed in the crack initiation evaluation hierarchy.

[0065] (2.1) Based on the data after parallel computing, damage assessment is performed using thermal oscillation load damage in the form of sinusoidal loading to obtain the maximum fatigue damage.

[0066] The rapid damage assessment technique for thermal oscillation load damage using sinusoidal loading refers to the conservative assumption that the pipeline is subjected to the maximum alternating thermal shock load (taking the maximum temperature difference between the hot and cold three-dimensional structures), and the definition method is shown in the following formula.

[0067]

[0068] In the formula, T fluid-ave ΔT represents the average fluid temperature (°C) in the thermal shock region. fluid f represents the amplitude (°C) of fluid temperature change in the thermal shock region. i Let t be the assumed frequency of thermal shock change, and t be time (s). The directional angle is rad.

[0069] In the assessment of rapid damage using sinusoidal thermal oscillation loads, sensitivity analysis was conducted to obtain different thermal shock frequencies f. i The structural stress response is analyzed, and based on this, structural fatigue damage analysis is performed, thereby identifying the impact frequency f that has the greatest impact on structural fatigue damage. maxThe corresponding maximum fatigue damage is UF max .

[0070] (2.2) Correction of maximum fatigue damage based on the influence of the primary loop water environment of nuclear power plants

[0071] For an extreme point i, record its preceding extreme point as i-1 and its following extreme point as i+1, and calculate the environmental impact factor F during the load history period (i-1, i) based on the following formula. en(i-1,i) Based on formula (4), the environmental impact factor F during the load history period (i, i+1) can also be calculated. en(i,i+1) .

[0072]

[0073] In the formula, k is the number of time points during the load history period (i-1, i) where transient operating parameters are recorded, j is the j-th recorded load fluctuation, and Δε j This refers to the strain change during a single fluctuation process.

[0074] For the extreme point i, it is recorded in the database of load spectrum records as (i, F en(i-1,i) F en(i,i+1) The specific time history of the load spectrum is no longer recorded, which significantly reduces the amount of data stored in the database.

[0075] In the comprehensive assessment of fatigue damage, the re-examination and the limit point (i, F) are considered. en(i-1,i) F en(i,i+1) Any extreme point (q, F) paired en(q-1,q) F en(q,q+1) The environmental impact factor F of the primary loop of the nuclear power plant, which is re-paired (i, q). en(i,q) The calculation method is as follows.

[0076] F en(i,q) =max(F en(i-1,i) ,F en(i,i+1) ,F en(q-1,q) ,F en(q,q+1) (5)

[0077] Additional consideration should be given to the impact of the primary circuit water environment of nuclear power plants on UF. max After the promoting effect, the final fatigue damage of the modified structure [UF] max ] end The calculation method is shown in equation (6). Standard requirements [UF] max ] end If the value is less than 1, then it is necessary to assume that the pipe has developed a crack.

[0078] [UF max ] end =UF max·F en(i,q) (6)

[0079] 3) In the crack failure evaluation hierarchy, the fracture-plastic instability dual failure mode is used to assess crack integrity and determine whether the primary loop main piping structure of the nuclear power plant is intact.

[0080] For [UF] max ] end In cases where the failure exceeds the specification limit by 1, or where accurate fatigue evaluation cannot be performed, the fracture-plastic instability dual failure mode is used to assess crack integrity. The assessment is based on the fracture assessment on the failure assessment diagram. The failure assessment diagram ignores the influence of pressure load, which is not the main load factor causing the failure of the primary loop main pipeline. The technical development and application research mainly focuses on the content of bending moment load.

[0081] like Figure 5 As shown, the crack integrity assessment using the fracture-plastic instability dual failure mode is as follows: based on the fracture assessment on the failure assessment diagram, the failure assessment diagram is obtained by finite element analysis calculation based directly on the J integral, and its assessment equation is as follows.

[0082]

[0083] In the formula, L r K is the x-axis of the evaluation curve. r To determine the ordinate of the curve.

[0084] In assessing crack integrity using a fracture-plastic instability dual failure mode, the maximum value of the horizontal axis of the failure assessment diagram is... The value is taken as 1.3. An example of transient operating parameters obtained from online monitoring of primary loop piping operating parameters in a nuclear power plant is shown below. Figure 6 As shown, the calculated stress of the pipe structure is as follows: Figure 7 As shown (the results of the influence function algorithm used in this invention to calculate the thermal stress of the pipeline can be consistent with the detailed finite element calculation results, with a maximum deviation of less than 3% in this case), a schematic diagram of the circumferential crack on the inner surface of the primary loop main pipeline in a nuclear power plant is shown below. Figure 8 As shown. An example of assessment based on failure assessment diagrams is shown below. Figure 9 As shown. Figure 9 The assessment point is obtained based on online monitoring and calculation of operating parameters. If the assessment point falls inside the failure assessment diagram, it means that the structure is safe. If the assessment point falls outside the failure assessment diagram, the structure faces the risk of fracture failure, that is, the structure of the primary loop pipeline of the nuclear power plant is incomplete.

[0085] Example 1

[0086] 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.

[0087] like Figure 1 As shown in the figure, the specific implementation process of the method for assessing the structural integrity of the primary loop main pipeline of a nuclear power plant based on online monitoring of transient parameters in this embodiment is as follows:

[0088] 1) The structural integrity assessment process for the primary loop main piping of a nuclear power plant based on online monitoring of transient parameters is as follows: Figure 2 As shown, the big data processing technologies in the initial data processing layer are:

[0089] (1.1) Collect online monitoring data, filter out common data, and process the common data in a unified manner.

[0090] Specifically, the computational data is processed according to the computational path i(P) i,path ), filter and extract the common data (D) required for the cumulative integration algorithm along the time axis. i,share And unify the time step size (Δt) and normalize the influence function. After undergoing a series of standardization processes, it is added to the standard data bus for subsequent use based on monitoring parameter D. i,monior Calculate the target parameter G i,path Formula (1) provides the application method.

[0091]

[0092] like Figure 3 As shown, for pipelines where localized thermal shock occurs, it is necessary to add local monitoring instruments for the temperature of the pipeline's outer surface.

[0093] Normalization of influence functions in a unified data processing workflow It refers to the response history of the pipeline structure under unit load in the reference transient state.

[0094] (1.2) Big Data Cluster Computing Technology: Optimizing the data after unified processing using parallel computing.

[0095] The computing cluster code is shown below. Figure 4 As shown, data parallel computing technology refers to optimizing the existing algorithm for accumulating integrals along the time axis to achieve the desired result within a specified decay time (t) of the influence function. d At any point j within the range, the target time series range (t) can be obtained by accumulating integrals simultaneously. j,start , t j,end Algorithms with arbitrary parameters within the target parameter G. i,path The parallel computing method is shown in the following equation:

[0096]

[0097] 2) A precise crack initiation evaluation method developed in the crack initiation evaluation hierarchy.

[0098] (2.1) Based on the data after parallel computing, damage assessment is performed using thermal oscillation load damage in the form of sinusoidal loading to obtain the maximum fatigue damage.

[0099] The conservative assumption is that the pipe interior is subjected to the maximum alternating thermal shock load, which is defined as shown in the following formula.

[0100]

[0101] In the formula, T fluid-ave ΔT represents the average fluid temperature (°C) in the thermal shock region. fluid f represents the amplitude (°C) of fluid temperature change in the thermal shock region. i Let t be the assumed frequency of thermal shock change, and t be time (s). The directional angle is rad.

[0102] use Figure 6 Actual monitoring data, ΔT fluid The temperature is taken as 108℃. When calculating, the material properties are selected based on data from 200℃. This temperature is T in formula (3). fluid-ave In the numerical analysis, the basic physical property parameters of the material were selected with reference to the French RCC-M standard.

[0103] Sensitivity analysis was used to identify the impact frequency f that has the greatest impact on structural fatigue damage. max The impact frequency f was obtained through analysis. max The fatigue damage to the structure is greatest in the range of 0.5–1.5 Hz. Subsequent fatigue damage assessments were conducted based on this frequency range, with the maximum fatigue damage being UF. max It is 0.00017.

[0104] (2.2) Correction of maximum fatigue damage based on the influence of the primary loop water environment of nuclear power plants

[0105] For an extreme point i, record its preceding extreme point as i-1 and its following extreme point as i+1, and calculate the environmental impact factor F during the load history period (i-1, i) based on the following formula. en(i-1,i) Based on formula (4), the environmental impact factor F during the load history period (i, i+1) can also be calculated. en(i,i+1) .

[0106]

[0107] In the formula, k is the number of time points during the load history period (i-1, i) where transient operating parameters are recorded, j is the j-th recorded load fluctuation, and Δε jThis refers to the strain change during a single fluctuation process.

[0108] For the extreme point i, it is recorded in the database of load spectrum records as (i, F en(i-1,i) F en(i,i+1) The specific time history of the load spectrum is no longer recorded, which significantly reduces the amount of data stored in the database.

[0109] In the comprehensive fatigue damage assessment and correction, for the re-examination of the limit point (i, F)... en(i-1,i) F en(i,i+1) Any extreme point (q, F) paired en(q-1,q) F en(q,q+1) The environmental impact factor F of the primary loop of the nuclear power plant, which is re-paired (i, q). en(i,q) The calculation method is as follows.

[0110] F en(i,q) =max(F en(i-1,i) ,F en(i,i+1) ,F en(q-1,q) ,F en(q,q+1) (5)

[0111] Calculate F en(i,q) If the value is approximately 3, then [UF max ] end It is 0.00051. [UF max ] end The value is less than 1, therefore, the primary loop pipe will not develop an initial crack under this condition.

[0112] 3) Crack integrity assessment technology for unified assessment of fracture-plastic instability dual failure modes

[0113] In this example, we further assume that an initial crack has developed in the primary loop piping of the nuclear power plant and conduct a conservative assessment.

[0114] Examples of transient operating parameters obtained from online monitoring of primary loop piping in nuclear power plants are as follows: Figure 6 As shown, the calculated stress of the pipe structure is as follows: Figure 7 As shown in Table 1, an example of calculating the reference defect stress intensity factor (SIF) under transient monitoring is presented.

[0115] Table 1. Example of calculating the reference defect stress intensity factor (SIF) under transient monitoring.

[0116]

[0117] An example of an assessment based on a failure assessment diagram is as follows: Figure 9 As shown, the assessment point falls within the failure assessment diagram, indicating that under the operating conditions of this example monitoring data, the primary loop piping structure of the nuclear power plant is intact and safe.

[0118] This invention presents a method for assessing the structural integrity of primary loop main piping in nuclear power plants based on online monitoring of transient parameters. It overcomes the technical challenge of rapid and accurate operation of big data in online monitoring and analysis of transient parameters in nuclear power plants. The method proposes a rapid damage assessment technique for primary loop main piping structures subjected to thermal oscillation loads and provides a convenient algorithm that considers the influence of the primary loop water environment. It also enables convenient storage of data in the form of load spectra across different refueling cycles. Furthermore, for cases where fatigue damage exceeds the specification limits (or where accurate fatigue evaluation is not possible), this invention provides a crack integrity assessment technique for the unified assessment of fracture-plastic instability dual failure modes in the primary loop main piping of nuclear power plants, providing in-depth technical support for the long-term safe operation of primary loop piping in nuclear power plants.

[0119] 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 assessing the structural integrity of the primary loop main piping of a nuclear power plant based on online monitoring of transient parameters, characterized in that, Includes the following steps: a) Initial Data Processing (a.1) Collect online monitoring data, filter out common data, and process the common data in a unified manner; (a.2) Optimize the data after unified processing using parallel computing; b) Evaluation of crack initiation (b.1) Based on the data after parallel computing, damage assessment is performed using a sinusoidal thermal oscillation load to obtain the maximum fatigue damage; (b.2) Correction of maximum fatigue damage based on the influence of the primary loop water environment of nuclear power plants; c) Crack failure evaluation If the maximum fatigue damage after correction is greater than 1, or if an accurate fatigue evaluation cannot be performed, the fracture-plastic instability dual failure mode is used to assess the crack integrity and determine whether the primary loop main pipeline structure of the nuclear power plant is intact. In step (b.1), the damage assessment using sinusoidal thermal oscillation load assumes that the pipe interior is subjected to the maximum alternating thermal shock load, and a rapid damage assessment using conservative loading in the form of sinusoidal thermal oscillation is performed. The definition method for the rapid damage assessment using conservative loading in the form of sinusoidal thermal oscillation is shown in the following formula: （3）； In the formula, T fluid-ave ΔT represents the average fluid temperature in the thermal shock region, expressed in °C. fluid f represents the amplitude of fluid temperature change in the thermal shock region, in °C. i Here is the assumed frequency of thermal shock change, where t is time, in seconds. The angle is expressed in rad. In step (b.2), the maximum fatigue damage is corrected based on the influence of the primary loop water environment of the nuclear power plant as follows: For extreme point i, the previous extreme point is recorded as i-1 and the next extreme point as i+1. The environmental impact factor F during the load history period (i-1, i) is calculated based on the following formula. en(i-1,i) Environmental impact factor F during the load history period (i, i+1) en(i,i+1) : （4）； In the formula, k is the number of time points during the load history period (i-1, i) where transient operating parameters are recorded, j is the j-th recorded load fluctuation, and Δε j This refers to the strain change during a single fluctuation process.

2. The evaluation method according to claim 1, characterized in that: In step (a.1), the unified processing is as follows: the calculation process parameters based on online monitoring data are processed according to the calculation path P. i,path Filter and extract the common data D required for the cumulative integration algorithm along the time axis. i,share And perform unified time step size Δt calculation and influence function normalization. The standardized data preprocessing process involves adding the preprocessed data to a standard data bus for subsequent processing based on monitoring parameter D. i,monior Calculate the target parameter G i,path use.

3. The evaluation method according to claim 2, characterized in that: The information provided for subsequent monitoring based on parameter D i,monior Calculate the target parameter G i,path The following formula is used: （1）。 4. The evaluation method according to claim 2, characterized in that: In step (a.1), the influence function is normalized. It is the reaction process of a primary loop pipeline under unit load.

5. The evaluation method according to claim 1, characterized in that: In step (a.1), for primary loop pipelines with localized thermal shock and thermal oscillation transient loads, it is necessary to add local temperature monitoring instruments on the outer surface of the pipeline to obtain more online local monitoring data.

6. The evaluation method according to claim 1, characterized in that: In step (a.2), the data parallel computation refers to optimizing the existing algorithm for accumulating integrals along the time axis to achieve the desired result within the decay time t of the specified influence function. d The target time series range (t) can be obtained by accumulating integrals at any time point j within the range. j,start , t j,end An algorithm with arbitrary parameters within ) 7. The evaluation method according to claim 6, characterized in that: Target parameter G i,path The data parallel computing method is shown in the following formula: （2）。 8. The evaluation method according to claim 1, characterized in that: In step (b.1), during the damage assessment using sinusoidal thermal oscillation loads, sensitivity analysis is used to obtain different thermal shock frequencies f. i The structural stress response is analyzed, and based on this, structural fatigue damage analysis is performed, thereby identifying the impact frequency f that has the greatest impact on structural fatigue damage. max The corresponding maximum fatigue damage is UF max .

9. The evaluation method according to claim 1, characterized in that: In step (b.2), for the extreme point i, it is recorded in the database of load spectrum records as (i, F en(i-1,i) F en(i,i+1) ).

10. The evaluation method according to claim 1, characterized in that: In the correction of step (b.2), for the re-interchange with the limit point (i, F) en(i-1,i) F en(i,i+1) Any extreme point (q, F) paired en(q-1,q) F en(q,q+1) The environmental impact factor F of the primary loop of the nuclear power plant, which is re-paired (i, q). en(i,q) The calculation method is as follows: （5）。 11. The evaluation method according to claim 1, characterized in that: In step (b.2), the corrected final fatigue damage [UF] max ] end The calculation method is shown in the following formula: （6）； If [UF max ] end If the value is greater than 1, it is assumed that the pipe has developed a crack.

12. The evaluation method according to claim 1, characterized in that: In step (c), the assessment of crack integrity using the fracture-plastic instability dual failure mode is as follows: a safety assessment is conducted based on the failure assessment diagram, and the structure is considered to have failed if either fracture or plastic instability occurs.

13. The evaluation method according to claim 12, characterized in that: In step (c), the fracture assessment based on the failure assessment diagram is established on the basis of J integral and a conservative assessment curve is obtained through a large number of finite element numerical simulations. The assessment curve assessment equation is shown in the following equation (7): （7）； In the formula, L r K is the x-axis of the evaluation curve. r To determine the ordinate of the curve.

14. The evaluation method according to claim 12, characterized in that: The maximum value of the abscissa of the failure assessment diagram in step (d) of crack integrity assessment. Take 1.

3.

15. The evaluation method according to claim 12, characterized in that: The crack integrity assessment in step (d) involves the assessment point falling on the assessment curve, coordinate axis, and the maximum value of the abscissa. The internal structure must meet the specifications; otherwise, the structure is incomplete.

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