Zhang, T.; Lu, K.; Katsuyama, Jinya; Li, Y.
International Journal of Pressure Vessels and Piping, 189, p.104262_1 - 104262_12, 2021/02
Mano, Akihiro; Katsuyama, Jinya; Miyamoto, Yuhei*; Yamaguchi, Yoshihito; Li, Y.
International Journal of Pressure Vessels and Piping, 179, p.103945_1 - 103945_6, 2020/01
Weld residual stress (WRS) is one of the most important factors in the structural integrity assessment of piping welds, and it is considered a driving force for crack growth. It is characterized by large uncertainty. For more rational assessment, it is important to consider the uncertainty of WRS for evaluating crack growth behavior in probabilistic fracture mechanics (PFM) analysis. In existing PFM analysis codes, WRS uncertainty is set by statistically processing the results of multiple finite element analyses. This process depends on the individual performing PFM analysis, which may lead to uncertainties whose sources would be different from the original WRS. In this study, we developed a new WRS evaluation model based on Fourier transformation, and the model was incorporated into PASCAL-SP, which has been developed by Japan Atomic Energy Agency. Through improvements to the code, WRS uncertainty can be considered automatically and appropriately by inputting multiple WRS analysis results directly as input data for PFM analysis.
Li, Y.; Azuma, Kisaburo*; Hasegawa, Kunio
International Journal of Pressure Vessels and Piping, 171, p.305 - 310, 2019/03
Udagawa, Makoto; Li, Y.; Nishida, Akemi; Nakamura, Izumi*
International Journal of Pressure Vessels and Piping, 167, p.2 - 10, 2018/11
It is important to assure the structural Integrity of piping systems under severe earthquakes because those systems comprise the pressure boundary for coolant with high pressure and temperature. In this study, we examine the seismic safety capacity of piping systems under severe dynamic seismic loading using a series of dynamic-elastic-plastic analyses focusing on dynamic excitation experiments of 3D piping systems which was tested by NIED. Analytical results were consistent with experimental data in terms of natural frequency, natural vibration mode, response accelerations, elbow opening-closing displacements, strain histories, failure position, and low-cycle fatigue failure lives. Based on these results, we concluded that the analytical model used in the study can be applied to failure behavior evaluation for piping systems under severe dynamic seismic loading.
Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio; Li, Y.
International Journal of Pressure Vessels and Piping, 131, p.85 - 95, 2015/07
A number of cracks due to primary water stress corrosion cracking (PWSCC) in pressurized water reactors and Ni-based alloy stress corrosion cracking (NiSCC) in boiling water reactors have been detected around Ni-based alloy welds. The causes of crack initiation and growth due to stress corrosion cracking include weld residual stress, operating stress, the materials, and the environment. We have developed the analysis code PASCAL-NP for calculating the failure probability and assessment of the structural integrity of cracked components on the basis of probabilistic fracture mechanics (PFM) considering PWSCC and NiSCC. This PFM analysis code has functions for calculating the incubation time of PWSCC and NiSCC crack initiation, evaluation of crack growth behavior considering certain crack location and orientation patterns, and evaluation of failure behavior near Ni-based alloy welds due to PWSCC and NiSCC in a probabilistic manner. Herein, actual plants affected by PWSCC have been analyzed using PASCAL-NP. Failure probabilities calculated by PASCAL-NP are in reasonable agreement with the detection data. Furthermore, useful knowledge related to leakage due to PWSCC was obtained through parametric studies using this code.
Katsuyama, Jinya; Ito, Hiroto*; Li, Y.*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 117-118, p.56 - 63, 2014/05
Several probabilistic fracture mechanical (PFM) analysis codes have been improved or developed in Japan, such as PASCAL-SP developed at JAEA, and PRAISE-JNES developed at JNES for structural integrity assessment of aged piping in nuclear power plants. Although they were developed for different purposes, they have similar functions. In this paper, in order to confirm the reliability and applicability of two PFM analysis codes, PASCAL-SP and PRAISE-JNES, benchmark analyses on piping failure probability have been carried out considering typical aging mechanisms, such as fatigue crack growth for piping materials in BWR plants Moreover, a criterion is proposed to judge whether the differences between the analysis results from two codes can be acceptable. Based on the proposed criterion, it is concluded that the analysis results of these two codes are in good agreements.
Li, Y.*; Ito, Hiroto*; Osakabe, Kazuya*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 99-100, p.61 - 68, 2012/11
A benchmark analysis was conducted using two probabilistic fracture mechanics analysis codes for aged piping in nuclear power plants, in order to confirm their reliability and applicability. These analysis codes have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms. In the benchmark analysis, the primary loop recirculation system piping in the boiling water reactor was selected as the typical piping system and stress corrosion cracking and fatigue were taken into account as the typical aging mechanisms. Moreover, a criterion was proposed for judging whether the differences between analysis results from the two codes are acceptable. Based on the benchmark analysis results and numerical investigation, it was concluded that the analysis results of these two codes agree very well.
Kanto, Yasuhiro*; Jhung, M.*; Ting, K.*; He, Y.*; Onizawa, Kunio; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 90-91, p.46 - 55, 2012/02
The International Round Robin (RR) activity was performed by the Probabilistic Fracture Mechanics (PFM) sub-committees of the Atomic Energy Research Committee of the Japan Welding Engineering Society (JWES) in conjunction with Korean, Taiwanese and Chinese research groups. The purposes of this program are to establish reliable procedures for evaluating the fracture probability of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) and to maintain the continuous cooperation among Asian institutes in the probabilistic approach to nuclear safety. This paper describes the outline of the problems and summarizes the results from all participant countries. The RR activity consists of two parts; deterministic analyses on stress and temperature in the reactor pressure vessel wall during PTS, and probabilistic analyses on vessel fracture probability due to PTS transients. The differences caused by the selection of analyzing programs and some input parameters are discussed.
Onizawa, Kunio; Nishikawa, Hiroyuki; Ito, Hiroto
International Journal of Pressure Vessels and Piping, 87(1), p.2 - 10, 2010/01
Probabilistic fracture mechanics (PFM) analysis codes for reactor pressure vessels (RPVs) and piping, called as PASCAL series has been developed at JAEA. The PASCAL2 code can evaluate the conditional probability of flaw initiation and fracture of an RPV under transient conditions such as pressurized thermal shock considering neutron irradiation embrittlement of the materials surrounding the reactor core of the RPV. Recent improvements of PASCAL2 are related to the treatment of weld-overlay cladding. The results of PFM analysis using the improved code have indicated that the residual stress by weld-overlay cladding affects the fracture probability of RPV. The results also indicate that the fracture toughness of cladding should be taken into consideration for the evaluation of realistic through-wall cracking probability. The PASCAL-SP code has been developed recently. The PASCAL-SP evaluates the probabilities of failures such as leakage and break of safety-related piping complying with Japanese regulation and rules. Effects of welding residual stress distribution as well as inspection accuracy are focused in this study. Residual stress distributions, which affect SCC behavior, have been determined by parametric FEM analyses verified through comparisons with welding experiments and incorporated to the code.
Kanto, Yasuhiro*; Onizawa, Kunio; Machida, Hideo*; Isobe, Yoshihiro*; Yoshimura, Shinobu*
International Journal of Pressure Vessels and Piping, 87(1), p.11 - 16, 2010/01
JAEA had sponsored research committees on probabilistic fracture mechanics (PFM) organized by the Japan Welding Engineering Society (JWES) for more than a decade. This work still continues with the same members in JWES. The purpose of the continuous activity is to provide probabilistic approaches in several fields of integrity problems of nuclear power plant. This paper shows some of the newest results of the JWES research committee. First topic is evaluation of the new JSME code case with rules of fitness-for-service from the view of PFM, including reactor pressure vessel subject to pressurized thermal shock loading, piping with a crack of the allowable size and effect of sizing accuracy for piping integrity. Another is development of new PFM techniques including reliability assessment of piping with domestic SCC data and maintenance optimization of LWRs based on risk and economic models. The last topic is the international round robin program just starting from 2008.
Wakai, Takashi; Sukekawa, Masayuki*; Date, Shingo*; Asayama, Tai; Aoto, Kazumi; Kubo, Shigenobu*
International Journal of Pressure Vessels and Piping, 85(6), p.352 - 359, 2008/06
This paper presents the provisional material specification and characteristics of the high chromium (Cr) steel for the sodium-cooled fast reactor (SFR) in Japan and creep-fatigue assessment of the steel. Based on the mechanical test and metallurgical examination results, it is clarified that tungsten (W) should be diminished to achieve better ductility and toughness. Then the provisional specifications of the high Cr steel for SFR components are proposed. Material characteristics, e.g. cyclic stress-strain relationship and creep strain curve, are also presented based on the material test results. Using these characteristics, a creep-fatigue strength assessment of the steel was performed. Conservative predictions were obtained and it was clarified that such conservativeness were resulted from over estimation of creep damage caused by too large initial stress at the beginning of dwell. It can be pointed out there are some rooms for improvement in the assessment procedure.
Shibata, Katsuyuki; Onizawa, Kunio; Li, Y.*; Kato, Daisuke*
International Journal of Pressure Vessels and Piping, 81(9), p.749 - 756, 2004/09
The paper describes the procedure to evaluate the ductile crack extension, where an increase in fracture resistance by a ductile crack extension is considered. Two standard -resistance curves are prepared for applying the elasto-plastic fracture criterion. Case studies concerning the effect of elasto-plastic fracture criterion were carried out using a severe PTS transient. The introduction of the elasto-plastic fracture criterion significantly contributes to remove the over-conservatism in applying the linear elastic fracture criterion. It was also found that the algorithm of the re-evaluation of crack tip characterization also has a significant effect on the failure probability.
Li, Y.*; Kato, Daisuke*; Shibata, Katsuyuki; Onizawa, Kunio
International Journal of Pressure Vessels and Piping, 78(4), p.271 - 282, 2001/04
no abstracts in English