Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Nuclear Engineering and Radiation Science, 5(3), p.031505_1 - 031505_8, 2019/07
Probabilistic fracture mechanics (PFM) analysis is expected as a rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and can evaluate the quantitative value 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 the structural integrity assessment of piping welds in nuclear power plants. In the latest few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in the nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus the structural integrity assessment taking account of PWSCC has become important. In this paper, we improved PASCAL-SP for the assessment considering PWSCC by introducing the several analytical functions such as the evaluation models of crack initiation time, crack growth rate and probability of crack detection. By using improved PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were evaluated as numerical examples. We also evaluated the influence of a leak detection and a non-destructive examination on the failure probabilities. On the basis of the numerical results, we concluded that the improved PASCAL-SP is useful for evaluating the failure probability of pipe taking PWSCC into account.
Chimi, Yasuhiro; Sato, Kenji*; Kasahara, Shigeki; Umehara, Ryuji*; Hanawa, Satoshi
Proceedings of Contribution of Materials Investigations and Operating Experience to Light Water NPPs' Safety, Performance and Reliability (FONTEVRAUD-9) (Internet), 10 Pages, 2018/09
To investigate the influence of Zinc (Zn) injection on primary water stress corrosion cracking (PWSCC) growth behavior, crack growth tests of 10% cold-worked Alloy 600 were performed in simulated primary water environment of pressurized water reactor (PWR) at 320
C with a low-concentration (5-10 ppb) Zn injection under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgH
O. As a result of the crack growth tests, DH-dependence of crack growth rate (CGR) showed a similar tendency to the predicted CGR based on the CGR data without Zn injection, indicating almost no effect of a low-concentration Zn injection on the crack growth behavior. Moreover, the microstructural analyses of oxide films formed inside the crack and on the specimen surface were conducted, and the intake of Zn in the oxides was detected on the specimen surface, but not detected inside the crack. This result was considered to be the cause of no Zn injection effect on the crack growth behavior.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Recently, 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. Structural integrity assessments taking PWSCC into account have become important. Probabilistic fracture mechanics (PFM) is expected as one of rational methods for the assessments because it can account for uncertainty of the influencing factors and evaluate the failure probabilities of components. In JAEA, a PFM analysis code PASCAL-SP was developed to evaluate the failure probability of nuclear pipe. This paper details improvement of the PASCAL-SP to evaluate the failure probability taking PWSCC into account. As numerical examples, the failure probabilities for circumferential and axial cracks due to PWSCC are evaluated. Influence of inspection on failure probabilities are evaluated. As the results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability taking PWSCC into account.
Watanabe, Norio
JAERI-Review 2004-015, 144 Pages, 2004/07
JAERI-Review-2004-015.pdf:25.58MB
In March 6, 2002, significant degradation to the reactor vessel head material adjacent to a control rod drive nozzle was discovered at a U. S. pressurized water reactor (PWR), Davis Besse. Based on the documents issued by the U.S. Nuclear Regulatory Commission (USNRC) and the licensee, this report provides a brief description on circumstances related to discovery of the reactor vessel head degradation, and describes the degraded condition and causes, the USNRC's responses to this event and so on. In attachments, given are the contents of the generic communications (Bulletins, Generic Letters and Information Notices) issued by the USNRC for the events involving boric acid corrosion and degradation of reactor coolant pressure boundary components and the summary of the investigation reports issued by the USNRC's lessons-learned task force.
Chimi, Yasuhiro; Sato, Kenji*; Kasahara, Shigeki; Umehara, Ryuji*; Hata, Kuniki; Hanawa, Satoshi; Nishiyama, Yutaka
no journal, ,
To investigate the influence of Zinc (Zn) injection on primary water stress corrosion cracking (PWSCC) growth behavior, crack growth tests of 10% cold-worked Alloy 600 in simulated primary water environment of pressurized water reactor (PWR) at 320C and 360C, with and without Zn injection, under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgHO. As a result of the crack growth tests at 360C, it is implied that Zn injection suppresses the oxidation inside the crack and the crack growth rate (CGR). From the crack growth tests at 320C with Zn injection, DH-dependence of CGR showed a similar tendency to the predicted CGR based on the CGR data without Zn injection. Obvious influence of Zn injection on DH-dependence of CGR at 320C was not observed in the present study.
