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.

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.

Application of Probabilistic Fracture Mechanics Methodology for Japanese Reactor Pressure Vessels Using PASCAL4

2019 ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) methodology, which represents the influence parameters in their inherent probabilistic distributions, is deemed to be promising in the structural integrity assessment of pressure-boundary components in nuclear power plants. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs, Version 4) which can be used to evaluate the failure frequency of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analyses are performed for a Japanese model RPV using PASCAL4, and the effects of non-destructive examination and neutron fluence mitigation on failure frequency of RPV are quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for the structural integrity assessment of RPVs and can enhance the applicability of PFM methodology.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

2020 ◽  
Vol 143 (2) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Thermal Transients ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

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 Through Benchmark Analyses

2018 ◽  
Author(s):  
Yinsheng Li ◽  
Shumpei Uno ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Terry Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Pressurized Thermal Shock ◽  
Great Confidence ◽  
Reactor Pressure

A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency based on Japanese methods and data to evaluate failure probabilities and failure frequencies of Japanese reactor pressure vessels (RPVs) considering pressurized thermal shock (PTS) events and neutron irradiation embrittlement. To verify PASCAL, we have been performing benchmark analyses by using a PFM code FAVOR which has been developed in the United States and utilized in nuclear regulation. Based on two-year activities, the applicability of PASCAL in failure probability and failure frequency evaluation of Japanese RPVs was confirmed with great confidence. The analysis conditions, approaches and results are given in this paper.

Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Considering Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants

2014 ◽  
Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Multiple Cracks ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

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.

Development of Crack Evaluation Models for Probabilistic Fracture Mechanics Analyses of Japanese Reactor Pressure Vessels

2018 ◽  
Author(s):  
Kai Lu ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Evaluation Models ◽  
Reactor Pressure

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency for structural integrity assessment of reactor pressure vessels (RPVs). The most recent release is PASCAL Version 4 (hereafter, PSACAL4) which can be used to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and pressurized thermal shock events. For the integrity assessment of RPVs, development of crack evaluation models is important. In this study, finite element analyses are performed firstly to verify the stress intensity factor calculations of cracks in PASCAL4. In addition, the applicability of the crack evaluation models in PASCAL4 such as the location of embedded cracks, crack shape and depth of surface cracks, and the increment of crack propagation is investigated. Based on sensitivity analyses of crack evaluation models for Japanese RPVs using PASCAL4, the effects of these evaluation models on failure frequency are clarified. From the analysis results, crack evaluation models recommended to the failure frequency evaluation for a Japanese model RPV are discussed.

Uncertainties in Probabilistic Fracture Mechanics Calculations

2010 ◽  
Author(s):  
F. A. Simonen
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Flaw Detection ◽  
Pressure Vessels ◽  
Operating Conditions ◽  
Statistical Distributions ◽  
Probabilistic Fracture Mechanics ◽  
Crack Growth Rates ◽  
Failure Probabilities

This paper addresses uncertainties in probabilistic fracture mechanics (PFM) calculations for pressure boundary components at commercial nuclear power plants. Such calculations can predict the probability that a component will have failed after a specified period of operation, but with large uncertainties that are difficult to quantify. PFM models only approximate details of as-built components as well as actual operating conditions over the lifetime of the component. Statistical distributions used as inputs to the calculations are subject to uncertainties, which also results in large uncertainties in calculated failure probabilities. This paper describes from the author’s perspective various uncertainties that are associated with PFM calculations. Efforts to quantify PFM uncertainties are described along with their impacts on calculated failure probabilities. Many uncertainties are explicitly addressed by statistical distributions for input parameters to the PFM models (e.g. crack growth rates, material strengths, probabilities of flaw detection, etc.). Other calculations have gone further by estimating uncertainties in the parameters of these statistical distributions along with uncertainties in parameters treated as deterministic inputs to the PFM models. Examples from the author’s experience with uncertainty analyses for pressure vessels and piping components are described.

Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels

2017 ◽  
Author(s):  
Jinya Katsuyama ◽  
Kazuya Osakabe ◽  
Shumpei Uno ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Neutron Irradiation ◽  
Structural Integrity ◽  
The United States ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

In Japan, to prevent nil-ductile fracture of reactor pressure vessels (RPVs) due to neutron irradiation embrittlement, deterministic fracture mechanics evaluation in accordance with the standards developed by the Japan Electric Association is performed for assessing the structural integrity of RPVs under pressurized thermal shock (PTS) events considering neutron irradiation embrittlement. In recent years, a structural integrity assessment methodology based on probabilistic fracture mechanics (PFM) has been introduced into the regulations in the United States and a few European countries. PFM is a rational methodology for evaluating the failure frequency of important pressure boundary components by considering the statistical distributions of various influence factors related to ageing due to the long-term operation. At Japan Atomic Energy Agency (JAEA), a PFM analysis code called PASCAL has been developed to evaluate the failure frequency of RPVs considering neutron irradiation embrittlement and PTS events. In addition, JAEA has developed a guideline for the PFM based structural integrity assessment of RPVs to promote the applicability of PFM in Japan and achieve the objective that an engineer/analyst who familiar with the fracture mechanics to perform PFM analyses and evaluate through-wall cracking frequency (TWCF) of RPVs easily. The guideline consists of a main body (general requirements), explanation (guidance), and several supplements. The technical basis for PFM analysis is also provided, and the new information and better fracture mechanics models are included in the guideline. In this paper, an overview of the guideline and some typical analysis results obtained based on the guideline and the Japanese database related to PTS evaluation are presented.

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