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

Author(s):  
Shotaro Hayashi ◽  
Mayumi Ochi ◽  
Kiminobu Hojo ◽  
Takahisa Yamane ◽  
Wataru Nishi

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.

Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  

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.


Author(s):  
Seiji Asada ◽  
Masao Itatani ◽  
Naoki Miura ◽  
Hideo Machida

Not only nonmandatory Appendix C, “Evaluation of Flaws in Piping,” in ASME Boiler & Pressure Vessel Code Section XI but also Appendix E-9, “Elastic-Plastic Fracture Mechanics Evaluation,” in the JSME Rules on Fitness-for-Service for Nuclear Power Plants use the load multiplier Z-factor that is applied to elastic-plastic fracture mechanics evaluation for a circumferential flaw of austenitic stainless steel piping and ferritic steel piping. The Z-factor is defined as the ratio of the limit load to the load at fracture load. Basically, the Z-factor equations were conservatively formulated by using the Z-factors for circumferential through-wall flaws. However, the Codes require flaw evaluation for circumferential surface flaws. Accordingly, Z-factors for circumferential surface flaws should be developed to have the consistency. Therefore Z-factor equations of austenitic stainless steel piping and ferritic steel piping have been developed for circumferential surface flaws.


2016 ◽  
Vol 3 (4) ◽  
pp. 16-00083-16-00083 ◽  
Author(s):  
Kiminobu HOJO ◽  
Shotaro HAYASHI ◽  
Wataru NISHI ◽  
Masayuki KAMAYA ◽  
Jinya KATSUYAMA ◽  
...  

Author(s):  
Hideo Machida ◽  
Manabu Arakawa ◽  
Norimichi Yamashita ◽  
Shinobu Yoshimura

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.


2021 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Yinsheng Li

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.


2014 ◽  
Vol 487 ◽  
pp. 54-57 ◽  
Author(s):  
Meng Yu Chai ◽  
Li Chan Li ◽  
Wen Jie Bai ◽  
Quan Duan

304 stainless steel and 316L stainless steel are conventional materials of primary pipeline in nuclear power plants. The present work is to summarize the acoustic emission (AE) characteristics in the process of pitting corrosion of 304 stainless steel, intergranular corrosion of 316L stainless steel and weldments of 316L stainless steel. The work also discussed the current shortcomings and problems of research. At last we proposed the coming possible research topics and directions.


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