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Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya
International Journal of Pressure Vessels and Piping, 222, p.105792_1 - 105792_11, 2026/08
Times Cited Count：0 Percentile：0.00(Engineering, Multidisciplinary)A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants while considering age-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the confidence in the results of PASCAL-SP, a benchmarking study was performed with the PFM analysis code, xLPR, which was developed by the U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are performed using common analysis conditions. This paper presents the details of these conditions and comparisons of the results between the two aforementioned codes. Both codes were found to provide nearly the same results in both deterministic and probabilistic analyses for a dissimilar metal weld subjected to primary water stress corrosion cracking.
Matsui, Tetsuya; Shimodaira, Masaki; Yamaguchi, Yoshihito; Toyama, Takeshi; Katsuyama, Jinya
JAEA-Research 2025-017, 41 Pages, 2026/03
JAEA-Research-2025-017.pdf:4.52MB
The JAEA Safety Research Center has been conducting fundamental research on advanced inspection and structural integrity assessment technologies since FY2024, including the development of a machine-learning-based ultrasonic flaw detection method using an ultrasonic simulator. To assess the simulator's applicability, phased array ultrasonic testing (PAUT) results produced by the simulator were compared with actual measurement data. Due to limited publicly available datasets, an intergranular crack in the pressurizer spray line piping of Kansai Electric Power Co. Inc.'s Ohi Nuclear Power Station Unit 3 was selected as the reference case. PAUT linear scanning analysis at a 45
incident angle detected the crack's corner and edge echoes. Strong columnar-crystal propagation echoes were also observed within the weld metal, with their intensity showing dependence on the symmetric axis angle. Analysis at a 31 incident angle similarly identified strong columnar-crystal propagation echoes, which connected to the crack's corner echoes and propagated into the weld region. These results align with actual measurements, indicating that the observed weld-metal echoes are likely attributable to columnar-crystal propagation.
Yamaguchi, Yoshihito; Li, S.; Katsuyama, Jinya
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07
Yamaguchi, Yoshihito; Hasegawa, Kunio; Negyesi, M.*
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07
Shimodaira, Masaki; Ha, Yoosung; Yamaguchi, Yoshihito; Hata, Kuniki; Katsuyama, Jinya
Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07
In the structural integrity assessment of the reactor pressure vessels (RPVs), the stress intensity factor acting on the tip of a postulated crack compares with fracture toughness evaluated by the fracture toughness test. The plastic constraint of the postulated crack is lower than the fracture toughness specimen due to the shallow crack depth. Assessing the structural integrity of the RPV with a low-constraint specimen may give an overly conservative result. Recently, a rational fracture assessment method has been developed for RPVs based on the local approach (LA). The LA can estimate the fracture toughness distribution using the Weibull stress, an index of the fracture independent of the plastic constraint. To apply the LA for RPV, it must be confirmed to accurately estimate the Weibull stress and the fracture toughness distribution. In this study, we focused on Weibull parameters, such as shape parameter m and the scaling factor 
, which are used to calculate the Weibull stress and the fracture probability, respectively. The effect of neutron irradiation on these parameters was investigated by conducting fracture toughness tests and finite element analyses. As a result, the m value corresponding to the uncertainty of the Weibull stress was not affected by irradiation. In contrast, it was found that the value should be optimized to accurately estimate the fracture toughness distribution for irradiated steel, considering the change in the tensile property.
Hasegawa, Kunio; Yamaguchi, Yoshihito; Udyawar, A.*
Journal of Pressure Vessel Technology, 147(3), p.034501_1 - 034501_7, 2025/06
Times Cited Count：0 Percentile：0.00(Engineering, Mechanical)
for negative stress ratios R based on trend in experimental data for fatigue crack growth rates of austenitic stainless steels for ASME code Section XINegyesi, M.*; Yamaguchi, Yoshihito; Hasegawa, Kunio; Lacroix, V.*; Morley, A.*
Journal of Pressure Vessel Technology, 147(2), p.021201_1 - 021201_7, 2025/04
Times Cited Count：0 Percentile：0.00(Engineering, Mechanical)Shimodaira, Masaki; Yamaguchi, Yoshihito; Iwata, Keiko; Katsuyama, Jinya; Chimi, Yasuhiro
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 10 Pages, 2024/07
no abstracts in English
Li, S.; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07
Negyesi, M.*; Yamaguchi, Yoshihito; Hasegawa, Kunio; Lacroix, V.*; Morley, A.*
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07
Ha, Yoosung; Yamaguchi, Yoshihito; Hasegawa, Kunio; Negyesi, M.*
Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 6 Pages, 2024/07
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
The steam generator (SG) is an important component of a pressurized water reactor. In addition, local wall-thinning has been reported in SG tubes. The burst differential pressure, considering both the internal and external pressures from the primary and secondary coolant systems, should be predicted for the failure probability evaluation or structural integrity assessment of SG tubes. In this study, based on the results of burst tests performed in Japan and the United States, we improved the existing burst pressure estimation method for SG tubes with wall-thinning. In addition, as an example of the utilization of the improved burst pressure estimation method, the conditional failure probabilities for SG tubes with local wall-thinning, which is necessary for probabilistic risk assessment and risk-informed decision making, are calculated considering the dimensions of the wall-thinning.
Yamaguchi, Yoshihito; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.
Proceedings of the ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 9 Pages, 2023/07
The stress intensity factor (SIF) for crack at nozzle corner is a key parameter in structural integrity assessment of nozzle in reactor pressure vessel (RPV). Although various SIF solutions for surface cracks at nozzle corners have been proposed, most of them are only focusing on the deepest point of the crack, and the information about geometric dimension of the nozzle corner is not clear. According to the previous fatigue test results regarding the surface crack at the nozzle corner, the amounts of crack growth at the surface points were larger than that at the deepest point of the crack. Such results imply that SIFs at the surface points may be higher than that at the deepest point. To increase the reliability of the structural integrity assessment, it is necessary to provide SIF solutions for both surface and deepest points. In this study, SIF solutions for two surface points and the deepest point of surface crack at nozzle corners are developed through finite element analyses and the solutions are provided corresponding to the geometric dimensions of nozzle corner and crack size.
Li, S.; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.; Deng, D.*
Proceedings of the ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 7 Pages, 2023/07
Yamaguchi, Yoshihito; Nishida, Akemi; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
The wall-thinning is one of the most important age-related degradation phenomena in nuclear piping systems. Furthermore, in recent years, several nuclear power plants in Japan have experienced severe earthquakes. Therefore, failure probability analysis and fragility evaluation of piping systems, taking both wall-thinning and seismic response stresses into consideration, have become increasingly important in seismic probabilistic risk assessment. In Japan Atomic Energy Agency, in order to evaluate the failure probability of aged piping system with wall-thinning, a probabilistic analysis code PASCAL-EC was developed. In this study, to evaluate the seismic fragility of a wall-thinned pipe, a model of seismic response stress considering the wall-thinning effect, a failure evaluation method for wall-thinned pipes, and functions related to uncertainties treatment for important influence parameters have been introduced to PASCAL-EC. In this paper, the improved PASCAL-EC is outlined and preliminary results of the seismic fragility evaluation performed using this code are provided.
Yamaguchi, Yoshihito; Mano, Akihiro; Li, Y.
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 10 Pages, 2022/07
The steam generator (SG) tube is one of the important components in pressurized water reactors. Flaws such as wall-thinning or stress corrosion cracking have been reported in SG tubes. The burst pressure where both the internal and external pressures from the primary and secondary coolant systems are considered must be predicted to assess the structural integrity of SG tubes. Burst tests were performed by various organizations. On the basis of the test results, failure estimation methods were proposed. In this study, previous burst test data and existing failure estimation methods for SG tubes with wall-thinning or crack were investigated. As a result, the coefficient of the existing estimation method for SG tube with uniform wall-thinning was updated. In addition, failure estimation methods that are suitable for SG tubes with crack or local wall-thinning were proposed by considering the effects of the flaw shape and size on the burst pressure. The applicability of the failure estimation methods was confirmed by comparing the predicted results with the burst test data in actual SG tubes.
Lacroix, V.*; Hasegawa, Kunio; Li, Y.; Yamaguchi, Yoshihito
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 7 Pages, 2022/07
Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Furuta, Takuya; Kaji, Yoshiyuki
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07
Yamaguchi, Yoshihito; Hasegawa, Kunio; Li, Y.; Lacroix, V.*
Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 4 Pages, 2022/07
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 144(1), p.011506_1 - 011506_9, 2022/02
Times Cited Count：2 Percentile：11.57(Engineering, Mechanical)In the past few decades, the cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints welded using nickel based alloy in piping system of boiling water reactors. Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a probabilistic structural integrity assessment of weld joint in pipe in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint welded using nickel based alloy in boiling water reactor susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.
