Yamaguchi, Yoshihito; Katsuyama, Jinya; Masaki, Koichi*; Li, Y.
JAEA-Research 2020-017, 80 Pages, 2021/02
The seismic probabilistic risk assessment (seismic PRA) is an important methodology to evaluate the seismic safety of nuclear power plants. Regarding seismic fragility evaluations performed in the seismic PRA, the Probabilistic Fracture Mechanics (PFM) can be applied as a useful evaluation technique for aged piping with crack or wall thinning due to the age-related degradation. Here, to advance seismic PRA methodology for the long-term operated nuclear power plants, a guideline for the fragility evaluation on the typical aged piping of nuclear power plants has been developed taking the aged-related degradation into account.
Yamaguchi, Yoshihito; Mano, Akihiro; Katsuyama, Jinya; Masaki, Koichi*; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2020-021, 176 Pages, 2021/02
In Japan Atomic Energy Agency, as a part of researches on the structural integrity assessment and seismic safety assessment of aged components in nuclear power plants, a probabilistic fracture mechanics (PFM) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed to evaluate failure probability of piping. The initial version was released in 2010, and after that, the evaluation targets have been expanded and analysis functions have been improved based on the state-of-the art technology. Now, it is released as Ver. 2.0. In the latest version, primary water stress corrosion cracking in the environment of Pressurized Water Reactor, nickel based alloy stress corrosion cracking in the environment of Boiling Water Reactor, and thermal embrittlement can be taken into account as target age-related degradation. Also, many analysis functions have been improved such as incorporations of the latest stress intensity factor solutions and uncertainty evaluation model of weld residual stress. Moreover, seismic fragility evaluation function has been developed by introducing evaluation methods including crack growth analysis model considering excessive cyclic loading due to large earthquake. Furthermore, confidence level evaluation function has been incorporated by considering the epistemic and aleatory uncertainties related to influence parameters in the probabilistic evaluation. This report provides the user's manual and analysis methodology of PASCAL-SP Ver. 2.0.
Sun, X. H.*; Wang, H.*; Otsu, Hideaki*; Sakurai, Hiroyoshi*; Ahn, D. S.*; Aikawa, Masayuki*; Fukuda, Naoki*; Isobe, Tadaaki*; Kawakami, Shunsuke*; Koyama, Shumpei*; et al.
Physical Review C, 101(6), p.064623_1 - 064623_12, 2020/06
The spallation and fragmentation reactions of Xe induced by proton, deuteron and carbon at 168 MeV/nucleon were studied at RIKEN Radioactive Isotope Beam Factory via the inverse kinematics technique. The cross sections of the lighter products are larger in the carbon-induced reactions due to the higher total kinetic energy of carbon. The energy dependence was investigated by comparing the newly obtained data with previous results obtained at higher reaction energies. The experimental data were compared with the results of SPACS, EPAX, PHITS and DEURACS calculations. These data serve as benchmarks for the model calculations.
Katsuyama, Jinya; Masaki, Koichi; Lu, K.; Watanabe, Tadashi*; Li, Y.
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07
For reactor pressure vessel (RPV) of pressurized water reactor, temperature of coolant water in emergency core cooling system (ECCS) may have influence on the structural integrity of RPV during pressurized thermal shock (PTS) events. Focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs in order to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. From the analysis results, it was shown that the failure probability of RPV was dramatically reduced when the coolant temperature in accumulator as well as high and low pressure injection systems (HPI/LPI) was raised, although raising the coolant temperature of HPI/LPI only did not cause reduction in the failure probability.
Aratani, Hidekazu*; Nakatani, Yasuhiro*; Fujiwara, Hidenori*; Kawada, Moeki*; Kanai, Yuina*; Yamagami, Kohei*; Fujioka, Shuhei*; Hamamoto, Satoru*; Kuga, Kentaro*; Kiss, Takayuki*; et al.
Physical Review B, 98(12), p.121113_1 - 121113_6, 2018/09
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
Li, Y.; Uno, Shumpei*; Masaki, Koichi; Katsuyama, Jinya; Dickson, T.*; Kirk, M.*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 11 Pages, 2018/07
Katsuyama, Jinya; Masaki, Koichi; Miyamoto, Yuhei*; Li, Y.
JAEA-Data/Code 2017-015, 229 Pages, 2018/03
As a part of the structural integrity research for aging light water reactor components, a probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed in JAEA. The PASCAL code can evaluate the conditional failure probabilities and failure frequencies for core region in reactor pressure vessels under the pressurized thermal shock events. In this study, we improved many functions such as the stress intensity factor solutions, the fracture toughness models, or confidence level evaluation function by considering epistemic and aleatory uncertainties related to influence parameters in the structural integrity assessment. We also developed the analysis module PASCAL-Manager which calculates the failure frequency for the entire core region taking into consideration the failure probabilities obtained from PACAL-RV. Based on these improvements, the new analysis code is upgraded to PASCAL Ver.4. This report provides the user's manual and theoretical background of PASCAL Ver.4.
Masaki, Koichi; Miyamoto, Yuhei*; Osakabe, Kazuya*; Uno, Shumpei*; Katsuyama, Jinya; Li, Y.
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 7 Pages, 2017/07
A probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed by Japan Atomic Energy Agency (JAEA). PASCAL can evaluate failure frequencies of Japanese reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events based on domestic structural integrity assessment models and data of influence factors. In order to improve the engineering applicability of PFM to Japanese RPVs, we have performed verification of the PASCAL. In general, PFM code consists of many functions such as fracture mechanics evaluation functions, probabilistic evaluation functions including random variables sampling modules and probabilistic evaluation models, and so on. The verification of PFM code is basically difficult because it is impossible to confirm such functions through the comparison with experiments. When a PFM code is applied for evaluating failure frequencies of RPVs, verification methodology of the code should be clarified and it is important that verification results including the region and process of the verification of the code are indicated. In this paper, our activities of verification for PASCAL are presented. We firstly represent the overview and methodology of verification of PFM code, and then, some verification examples are provided. Through the verification activities, the applicability of PASCAL in structural integrity assessments for Japanese RPVs was confirmed with great confidence.
Li, Y.; Katsumata, Genshichiro*; Masaki, Koichi*; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. 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. The source program of PASCAL was released to the members of the working group. 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.
Li, Y.; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Kanto, Yasuhiro*; Suzuki, Masahide*; Masaki, Koichi*
JAEA-Review 2017-005, 80 Pages, 2017/03
For the improvement of the structural integrity assessment methodology on reactor pressure vessels (RPVs), the probabilistic fracture mechanics (PFM) analysis code PASCAL has been developed and improved in JAEA based on latest knowledge. The PASCAL code evaluates the failure probabilities and frequencies of Japanese RPVs under transient events such as pressurized thermal shock considering neutron irradiation embrittlement. In order to confirm the reliability of the PASCAL as a domestic standard code and to promote the application of PFM on the domestic structural integrity assessments of RPVs, it is important to verify the probabilistic variables, functions and models incorporated in the PASCAL and summarize the verification processes and results as a document. On the basis of these backgrounds, we established a working group, composed of experts on this field besides the developers, on the verification of the PASCAL3 which is a PFM analysis module of PASCAL, and the source program of PASCAL3 was released to the members of working group. Through one year activities, the applicability of PASCAL in structural integrity assessments of domestic RPVs was confirmed with great confidence. This report summarizes the activities of the working group on the verification of PASCAL in FY2015.
Yamasaki, Atsushi*; Fujiwara, Hidenori*; Tachibana, Shoichi*; Iwasaki, Daisuke*; Higashino, Yuji*; Yoshimi, Chiaki*; Nakagawa, Koya*; Nakatani, Yasuhiro*; Yamagami, Kohei*; Aratani, Hidekazu*; et al.
Physical Review B, 94(11), p.115103_1 - 115103_10, 2016/11
In this study, we systematically investigate three-dimensional(3D) momentum-resolved electronic structures of Ruddlesden-Popper-type iridium oxides SrIrO using soft-X-ray angle-resolved photoemission spectroscopy (SX-ARPES). Our results provide direct evidence of an insulator-to-metal transition that occurs upon increasing the dimensionality of the IrO-plane structure. This transition occurs when the spin-orbit-coupled = 1/2 band changes its behavior in the dispersion relation and moves across the Fermi energy. By scanning the photon energy over 350 eV, we reveal the 3D Fermi surface in SrIrO and -dependent oscillations of photoelectron intensity in SrIrO. To corroborate the physics deduced using low-energy ARPES studies, we propose to utilize SX-ARPES as a powerful complementary technique, as this method surveys more than one whole Brillouin zone and provides a panoramic view of electronic structures.
Hojo, Kiminobu*; Hayashi, Shotaro*; Nishi, Wataru*; Kamaya, Masayuki*; Katsuyama, Jinya; Masaki, Koichi*; Nagai, Masaki*; Okamoto, Toshiki*; Takada, Yasukazu*; Yoshimura, Shinobu*
Mechanical Engineering Journal (Internet), 3(4), p.16-00083_1 - 16-00083_16, 2016/08
Performance demonstration certification of non-destructive inspection for cast stainless steel (CASS) has been planned but the target flaw depth to be detected has not been determined yet in Japan. The target flaw size is closely connected to the allowable flaw size which is determined by flaw evaluation of the rules on fitness-for-service. For rational mitigation of the acceptable flaw size, application of probabilistic fracture mechanics (PFM) is one of the useful countermeasures compared with deterministic approach. In this paper, benchmark problems for a CASS pipe were proposed with intention applying and verifying PFM codes. As the fracture modes, fatigue crack extension, plastic collapse and ductile crack initiation were assumed. Six organizations participated in the benchmark analysis and failure probabilities from them were compared. As a result the failure probability of each problem showed good agreement and the code for application of CASS issue has been verified.
Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Yoshimura, Shinobu*
Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 8 Pages, 2015/07
A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.
Osakabe, Kazuya*; Masaki, Koichi*; Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 7 Pages, 2014/07
The structural integrity of reactor pressure vessel (RPV) during pressurized thermal shock events is judged to be maintained unless the stress intensity factors at the crack tip is smaller than fracture toughness based on deterministic approach in the current Japanese code. Application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RPVs has become attractive recently, because uncertainties of several parameters can be incorporated rationally. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated. In this study, in order to identify the conservatism in the current code, PFM analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007 is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.
Masaki, Koichi; Katsuyama, Jinya; Onizawa, Kunio
Journal of Pressure Vessel Technology, 136(1), p.011208_1 - 011208_7, 2014/02
In order to apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), PFM analysis code has been developed at JAEA. Using the PFM analysis code, PASCAL version 3, the conditional probabilities of crack initiation and fracture for an RPV during pressurized thermal shock events have been analyzed. Sensitivity analyses on some input parameters were performed to clarify the effect on the conditional fracture probability. Comparison between the conditional probabilities and temperature margin from current deterministic analysis method were made for some model plant conditions of domestic typical old-type RPVs. From the analyses, a good correlation between temperature margin and the conditional probability of crack initiation was obtained.
Suzuki, Sadaaki; Yagyu, Junichi; Masaki, Kei; Nishiyama, Tomokazu; Nakamura, Shigetoshi; Saeki, Hisashi; Hoshi, Ryo; Sawai, Hiroaki; Hasegawa, Koichi; Arai, Takashi; et al.
NIFS-MEMO-67, p.266 - 271, 2014/02
no abstracts in English
Katsuyama, Jinya; Masaki, Koichi; Onizawa, Kunio
Journal of Pressure Vessel Technology, 135(6), p.061403_1 - 061403_8, 2013/12
When weld defects are observed during an inspection after welding, repair welding is performed after removing the defects. However, partial repair welding can potentially complicate the weld residual stress distribution. In this study, we performed thermal-elastic-plastic analyses to evaluate the weld residual stress produced by repair welding after pipe butt-welding. In addition, based on the analysis results for varying repair-welding conditions, we also performed structural integrity assessments related to stress corrosion cracking using the probabilistic fracture mechanics analysis code PASCAL-SP. It was clearly observed that the tensile stress in the repair-welded region increased and that compressive stresses occurred outside the repair-welded region. A deeper mechanical cutting depth caused larger increases in the tensile residual stress of the repair-welded region. It was also clarified that partial repair welding may favorably affect the break probability of piping welds.
Shibanuma, Kiyoshi; Arai, Takashi; Hasegawa, Koichi; Hoshi, Ryo; Kamiya, Koji; Kawashima, Hisato; Kubo, Hirotaka; Masaki, Kei; Saeki, Hisashi; Sakurai, Shinji; et al.
Fusion Engineering and Design, 88(6-8), p.705 - 710, 2013/10