Failure Probability Analysis of Aged Piping Using Probabilistic Fracture Mechanics Methodology Considering Seismic Loads

2016 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Fragility Curves ◽  
Safety Margin ◽  
Probability Analysis ◽  
Seismic Loads ◽  
Probabilistic Fracture Mechanics ◽  
Age Related ◽  
Piping Systems

Several nuclear power plants in Japan have been operating for more than 30 years and cracks due to age-related degradations have been detected in some piping systems during in-service inspections. Furthermore, several of them have experienced severe earthquakes in recent years. Therefore, failure probability analysis and fragility evaluation for piping systems, taking both age-related degradations and seismic loads into consideration, has become increasingly important for the structural integrity evaluation and the seismic probabilistic risk assessment. Probabilistic fracture mechanics (PFM) is recognized as a rational methodology for failure probability analysis and fragility evaluation of aged piping, because it can take the scatters and uncertainties of influence parameters into account. In our Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed for aged piping considering age-related degradations. In this study, we improved PASCAL-SP for the fragility evaluation taking both age-related degradations and seismic loads into account. The details of the improvement of PASCAL-SP are explained and some example analysis results of failure probabilities, fragility curves and a preliminary investigation on seismic safety margin are presented in this paper.

Development of Guideline on Seismic Fragility Evaluation for Aged Piping

2021 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Yinsheng Li
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Seismic Fragility ◽  
Degradation Mechanisms ◽  
Probabilistic Fracture Mechanics ◽  
Age Related ◽  
Piping Systems

Abstract The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regardto PWSCC

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Crack Detection ◽  
Structural Integrity ◽  
Energy Agency ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel

2021 ◽  
Vol 143 (4) ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Nowadays, it has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PFM analysis of structural components in aging light water reactor (PASCAL) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against nonductile fracture. A series of activities has been performed to verify the applicability of PASCAL. As a part of the verification activities, a working group was established with seven organizations from industry, universities, and institutes voluntarily participating as members. Through one-year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group, including the verification plan, approaches, and results.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

A Probabilistic Evaluation Model for Welding Residual Stress Distribution at Piping Joint in Probabilistic Fracture Mechanics Analysis

2008 ◽  
Author(s):  
Hiroto Itoh ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Residual Stress ◽  
Stress Distribution ◽  
Fracture Mechanics ◽  
Failure Probability ◽  
Evaluation Model ◽  
Residual Stress Distribution ◽  
Welding Residual Stress ◽  
Probabilistic Fracture Mechanics ◽  
Probabilistic Evaluation ◽  
Aging Knowledge

Stress corrosion cracking (SCC) has been observed at some piping joints made by Austenitic stainless steel in BWR plants. In JAEA, we have been developing probabilistic fracture mechanics (PFM) analysis methods for aged piping based on latest aging knowledge and an analytical code, PASCAL-SP. PASCAL-SP evaluates the failure probability of piping at aged welded joints under SCC by a Monte Carlo method. We proposes a simplified probabilistic model which can be applied to the failure probability analysis based on PFM for welded joint of piping considering the uncertainty of welding residual stress. And the probabilistic evaluation model is introduced to PASCAL-SP. A parametric PFM analysis concerning uncertainties of residual stress distribution using PASCAL-SP was performed. The PFM analysis showed that the uncertainties of residual stress distribution largely influenced break probability. The break probability increased with increasing the uncertainties of residual stress.

Accounting for Constraint in Probabilistic Fracture Mechanics Analysis of Welds in Floating Production, Storage and Off-Loading (FPSO) Vessels

2003 ◽  
Author(s):  
Jens P. Tronskar
Keyword(s):  
Residual Stress ◽  
Stress Distribution ◽  
Fracture Mechanics ◽  
Crack Growth ◽  
North Sea ◽  
Failure Probability ◽  
Crack Tip ◽  
Probabilistic Fracture Mechanics ◽  
Crack Tip Constraint

Revision 4 of the British Energy R6 document: “Assessment of the integrity of structures containing defects” provides methods to allow for loss of crack tip constraint for shallow weld flaws. The document also provides methods to estimate upper-bound values of the through thickness residual stress distribution for a range of common weld joint configurations. The present paper presents results of analyses where approaches to modify the R6 Option 1 or 2 failure assessment diagrams (FADs) for loss of crack tip constraint pertaining to primary and non-uniform residual stress have been applied. The modified FAD were formulated for probabilistic fracture mechanics analyses of semi-elliptical surface cracks located at transverse deck welds of Floating Production, Storage and Off-loading (FPSO) vessels designed to operate in the North Sea. The objective was to study the influence on the failure probability of modifying the FAD for constraint and allowing for non-uniform residual stress. Another objective was to study the influence of constraint correction on the combined fatigue and fracture failure probability for the vessels subjected to wave loading. Material and weld tensile properties and fracture toughness distributions for input to the probabilistic fracture mechanics analyses were obtained from testing of welded panels prepared using welding procedures for actual FPSO fabrication. The loading conditions were derived based on North Sea wave data pertaining to the offshore field where the FPSO is operating. The stresses were obtained from global FE analysis and fitted Weibull long-term and extreme value distributions. The results of the analyses demonstrate clearly the importance of correcting for crack tip constraint pertaining to both primary and secondary stress and to allow for non-uniform residual stress for shallow surface flaws of known crack heights. However, in combination with fatigue crack growth the effects become less prominent as the failure probability is governed by the uncertainty in the parameters of the crack growth relationship and the long-term stress distribution.

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.

Sign in / Sign up

Export Citation Format

Share Document