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.

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.

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.

Allowable Flaw Size of Japanese Cast Stainless Steel Pipe Using Probabilistic Fracture Mechanics Method

2017 ◽  
Author(s):  
Shotaro Hayashi ◽  
Mayumi Ochi ◽  
Kiminobu Hojo ◽  
Takahisa Yamane ◽  
Wataru Nishi
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Ageing ◽  
Probabilistic Fracture Mechanics ◽  
Stainless Steel Pipe ◽  
Technical Basis ◽  
Cast Stainless Steel

The cast austenitic stainless steel (CASS) that is used for the primary loop pipes of nuclear power plants is susceptible to thermal ageing during plant operation. The Japanese JSME rules on fitness-for-service (JSME rules on FFS)[1] for nuclear power plants specify the allowable flaw depths. However, some of these allowable flaw sizes are small compared with the smallest flaw sizes, which can be detected by nondestructive testing. ASME Section XI Code Case N-838[2] recently specified the maximum tolerable flaw depths for CASS pipes determined by probabilistic fracture mechanics (PFM). In a similar way, the allowable flaw depths of CASS pipes were calculated by PFM analysis code “PREFACE”[3] which considers uncertainty of the mechanical properties of Japanese PWR CASS materials. In order to confirm the validity of PREFACE, the allowable flaw depths calculated by PREFACE were compared with the maximum tolerable flaw depths in the technical basis of Code Case N-838. As a result, although the J calculation method and the embrittlement prediction model of CASS are different, these were qualitatively consistent. In addition, the sensitivity of ferrite content to the allowable flaw depths was investigated.

Initiation and Growth of Small Fatigue Cracks of Steels Used for Nuclear Power Plants Under Low Cycle Regime

2014 ◽  
Author(s):  
Shota Hasunuma ◽  
Takeshi Ogawa
Keyword(s):  
Fatigue Crack ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Cycle Fatigue ◽  
Fatigue Cracks ◽  
Cycle Fatigue

Low cycle fatigue tests were conducted for carbon steel, STS410, low alloy steel, SFVQ1A, and austenitic stainless steel, SUS316NG, which were used for nuclear power plants, in order to investigate the mechanism of fatigue damage when the plants were subjected to huge seismic loads. In these tests, the surface behavior of fatigue crack initiation and growth was observed in detail using cellulose acetate replicas, while the interior behavior was detected in terms of fracture surface morphology developed by multiple two-step strain amplitude variations with periodical surface removals. Fatigue crack growth rates were evaluated by elasto-plastic fracture mechanics approach. For SFVQ1A and SUS316NG, the fracture mechanics approach is available in order to predict the crack growth life from the metallurgical crack initiation size to the final crack length of the specimens. For STS410, numerous small cracks initiated, grew and coalesced each other on the specimen surface under low cycle fatigue regime.

Study on Probabilistic Fracture Mechanics Analysis Codes for Pipes With Stress Corrosion Cracks

2008 ◽  
Author(s):  
Hideo Machida ◽  
Manabu Arakawa ◽  
Norimichi Yamashita ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Nuclear Power ◽  
Power Plants ◽  
Reliability Assessment ◽  
Multiple Cracks ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Integrity Management ◽  
Technical Basis

Risk-Informed integrity management methodologies have been developed in Japanese nuclear power plants. One of the issues of concern is the reliability assessment of piping with flaws due to stress corrosion cracking (SCC). Therefore, the probabilistic fracture mechanics analysis code have been developed, which can perform the reliability assessment for the austenitic stainless steel piping with flaws due to SCC. This paper describes technical basis of this code. This method is based on Monte-Carlo technique considering many sample cases in a piping section, where the initiation and growth of cracks are calculated and piping failures, including leaks and rapture, are evaluated. A notable feature is that multiple cracks can be treated, consequently, assessment of coalescence of cracks and intricate break evaluation of piping section have been included. Moreover, the in-service inspection (ISI) and integrity evaluation by Fitness-for-Service (FFS) code are integrated into the analysis, and the contribution to failure probability decrease can be assessed. Key parameters are determined on a probability basis with the designated probability type throughout the procedure. Size, location and time of crack initiation, coefficients of crack growth due to SCC and factors for piping failure are included in those parameters. With this method the reliability level of the piping through the operation periods can be estimated and the contribution of various parameters including ISI can be quantitatively evaluated.

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.

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