Analysis of the effect of crack curvature in Mini-C(T) specimen on fracture toughness evaluation

Shimodaira, Masaki; Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Onizawa, Kunio

Journal of Pressure Vessel Technology, 148(2), p.021504_1 - 021504_10, 2026/04

https://doi.org/10.1115/1.4070669

In the current structural integrity assessment of the reactor pressure vessel, the accurate reference temperature (T) based on the Master Curve method is necessary. The T can be estimated by using the Mini-C(T) fracture toughness specimen in accordance with ASTM E1921 and JEAC4216, which prescribe the crack straightness criteria. A requirement in ASTM E1921 has been revised in a decade to increase the accuracy and reasonability, and the applicable crack curvature has been varied by applied codes. The crack curvature of the Mini-C(T) specimen might have an impact on the T because of the variation of the plastic constraint. In this work, the effect of the crack curvature on the fracture toughness (K) evaluation using the Mini-C(T) specimen was quantitatively evaluated by using the finite element analysis (FEA) including the Weibull stress analysis, to discuss the difference in a requirement of the crack straightness in ASTM E1921 and JEAC4216. FEAs showed a possibility that the upper limit curvature would decrease the plastic constraint, and consequently obtain higher K in the Mini-C(T) specimen. Furthermore, if the upper limit curvature according to the ASTM E1921-21 was allowed, the T would be estimated as nonconservative based on the Weibull stress analysis. In contrast, the difference in (T) between the crack with upper limit curvature according to JEAC4216 and the ideal straight crack was not significant.

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.

Assessment and improvement of flaw-to-surface proximity rules accounting for flaw aspect ratio and component thickness

Lacroix, V.*; Dulieu, P.*; Hasegawa, Kunio; Ha, Yoosung; Negyesi, M.*

Proceedings of the ASME 2025 Pressure Vessels & Piping Conference (PVP2025) (Internet), 8 Pages, 2025/07

https://doi.org/10.1115/PVP2025-154613

Allowable circumferential flaw sizes based on code given and actual measured flow stresses for high toughness ductile pipes subjected to bending and tensile loads

Negyesi, M.*; Ha, Yoosung; Hasegawa, Kunio; Lacroix, V.*

Proceedings of the ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 6 Pages, 2024/07

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.

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

Fracture toughness in postulated crack area of PTS evaluation in highly-neutron irradiated RPV steel

Ha, Yoosung; Shimodaira, Masaki; Takamizawa, Hisashi; Tobita, Toru; Katsuyama, Jinya; Nishiyama, Yutaka

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 6 Pages, 2021/07

Ion-induced irradiation hardening of the weld heat-affected zone in low alloy steel

Ha, Yoosung; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi; Nishiyama, Yutaka

Nuclear Instruments and Methods in Physics Research B, 461, p.276 - 282, 2019/12

https://doi.org/10.1016/j.nimb.2019.10.018

Susceptibility to neutron irradiation embrittlement of heat-affected zone of reactor pressure vessel steels

Takamizawa, Hisashi; Katsuyama, Jinya; Ha, Yoosung; Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 8 Pages, 2019/07

https://doi.org/10.1115/PVP2019-94011

no abstracts in English

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

Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Hanawa, Satoshi; Nishiyama, Yutaka

Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 6 Pages, 2018/07