Probabilistic fracture mechanics benchmarking study of PASCAL-SP code with xLPR code regarding primary water stress corrosion cracking

Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya

International Journal of Pressure Vessels and Piping, 222, p.105792_1 - 105792_11, 2026/08

https://doi.org/10.1016/j.ijpvp.2026.105792

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.

Development of phenomenological degradation models for Cr-Coated Zr alloy cladding under high-temperature oxidation conditions

Taniguchi, Yoshinori; Luu, V. N.; Tasaki, Yudai; Udagawa, Yutaka; Katsuyama, Jinya

Annals of Nuclear Energy, 231, p.112177_1 - 112177_16, 2026/06

https://doi.org/10.1016/j.anucene.2026.112177

Ultrasonic analysis of columnar crystal propagation echo in acoustically anisotropic weld metals; Targeting intergranular cracking of stainless-steel piping in pressurized water reactor primary system "FY2024 Nuclear Regulation Research Technology Infrastructure Construction Project Subsidy 'The Grant to Establish Technology Bases for Enhancement of Nuclear Regulation Research'"

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.

High-temperature oxidation failure in reactivity-initiated accidents; An Evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments

Luu, V. N.; Taniguchi, Yoshinori; Udagawa, Yutaka; Katsuyama, Jinya

Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10

https://doi.org/10.1016/j.nucengdes.2025.114222

Effect of rod internal gas state on FFRD behavior of high burnup fuel during LOCA conditions

Kakiuchi, Kazuo; Narukawa, Takafumi*; Udagawa, Yutaka; Katsuyama, Jinya; Mihara, Takeshi; Amaya, Masaki

Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1440 - 1449, 2025/10

Applicability of fracture evaluation method based on local approach to an irradiated low-alloy steel

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

https://doi.org/10.1115/PVP2025-151993

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.

Effects of welding and constraint conditions on the welding residual stress and hardness of Type 316 stainless steel pipe

Li, S.; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 8 Pages, 2024/07

https://doi.org/10.1115/PVP2024-122731

Fracture toughness evaluation of weld-HAZ in RPV steel using Mini-C(T) specimens

Ha, Yoosung; Shimodaira, Masaki; Katsuyama, Jinya

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 7 Pages, 2024/03

Heat-affected zone (HAZ) produced by butt-welding in reactor pressure vessel (RPV) steel is one of the representative materials for surveillance program. The fracture toughness values of HAZ may show a large uncertainty due to inhomogeneous metallurgical structures. Also, the inhomogeneous microstructure in HAZ may influence on the degree of uncertainty in the fracture toughness and the sensitivity to irradiation embrittlement. We investigated the fracture toughness in HAZ of unirradiated material with respect to its distance from the fusion line of welds, where the amount of mixed microstructures change due to the thermal history during the welding. Mini-C(T) specimens of HAZ were harvested from the crack position at 0.5 mm, 1 mm and 2 mm from the fusion line of welds. The uncertainty of fracture toughness in HAZ, from the fusion line at 0.5 mm in particular, was larger than those of base metal at a quarter thickness. From the results of fracture toughness evaluation considering the standard deviation, there was the difference of reference temperature, in each position of HAZ. in all positions of HAZ was significantly lower than that of base metal, which means the fracture toughness in HAZ was greater than that of base metal at a quarter thickness.

Development of stress intensity factor solution for surface crack at nozzle corner in reactor pressure vessel

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.

Fracture toughness evaluation of the heat-affected zone under the weld overlay cladding in reactor pressure vessel steel

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya

Journal of Pressure Vessel Technology, 145(2), p.021501_1 - 021501_9, 2023/04

https://doi.org/10.1115/1.4055626