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Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya
Journal of Pressure Vessel Technology, 145(2), p.021501_1 - 021501_9, 2023/04
Times Cited Count:1 Percentile:59.37(Engineering, Mechanical)Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi
Journal of Pressure Vessel Technology, 144(1), p.011304_1 - 011304_7, 2022/02
Times Cited Count:0 Percentile:0(Engineering, Mechanical)In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (K) should be higher than the stress intensity factor at the crack tip of an under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and K evaluation. In this study, we performed fracture toughness tests and finite element analyses (FEAs) to investigate the effect of cladding on K evaluation. FEA showed that the cladding decreased the plastic constraint in the UCC rather than the surface crack. Moreover, it was also found that the apparent K for the UCC was higher than that for the surface crack from tests and the local approach.
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
Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Shimodaira, Masaki; Tobita, Toru; Nagoshi, Yasuto*; Lu, K.; Katsuyama, Jinya
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 8 Pages, 2021/07
In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (K) should be higher than the stress intensity factor at the crack tip of a semi-elliptical shaped under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and K evaluation. In this study, we performed fracture toughness tests and finite element analyses to investigate the effect of plastic constraint and cladding on the semi-elliptical shaped crack in K evaluation. The apparent K value evaluated at the deepest point of the crack exceeded 5% fracture probability based on the Master Curve method estimated from C(T) specimens, and the conservativeness of the current integrity assessment method was confirmed. Few initiation sites were observed along the tip of semi-elliptical shaped crack other than the deepest point. The plastic constraint state was also analyzed along the crack tip, and it was found that the plastic constraint at the crack tip near the surface was lower than that for the deepest point. Moreover, it was quantitatively showed that the UCC decreased the plastic constraint. The local approach suggested higher K value for the UCC than that for the surface crack, reflecting the low constraint effect for the UCC.
Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.
Ha, Yoosung; Shimodaira, Masaki; Tobita, Toru; Hanawa, Satoshi; Yamasaki, Shota*; Uno, Sadanori*
2018-Nendo Ryoshi Kagaku Gijutsu Kenkyu Kaihatsu Kiko Shisetu Kyoyo Jisshi Hokokusho (Internet), 3 Pages, 2019/09
no abstracts in English
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
no abstracts in English
Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio
JAEA-Data/Code 2018-013, 60 Pages, 2018/11
Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.
Ha, Yoosung; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Nishiyama, Yutaka
Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10
Times Cited Count:6 Percentile:32.89(Engineering, Mechanical)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
Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka
JAEA-Research 2017-018, 122 Pages, 2018/03
Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.
Shamsuzzaman, M.*; Horie, Tatsuro*; Fuke, Fusata*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Tagami, Hirotaka; Suzuki, Toru*; Tobita, Yoshiharu
Annals of Nuclear Energy, 111, p.474 - 486, 2018/01
Times Cited Count:17 Percentile:85.18(Nuclear Science & Technology)Ha, Yoosung; Tobita, Toru; Takamizawa, Hisashi; Nishiyama, Yutaka
Proceedings of 2017 ASME Pressure Vessels and Piping Conference (PVP 2017) (CD-ROM), 5 Pages, 2017/07
The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. value determined from irradiated Mini-C(T) specimens was in good agreement with that determined from the irradiated pre-cracked Charpy-type (PCCv) specimens. Also, the scatter of the 1T-equivalent fracture toughness values obtained from the irradiated Mini-C(T) specimens was not significantly different from that obtained from the irradiated PCCv. values determined from Mini-C(T) specimens agreed very well with the correlation between Charpy 41J transition temperature and of commercially manufactured RPV steels.
Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu; Suzuki, Toru
Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04
The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.
Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04
no abstracts in English
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Toyooka, Junichi; Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Matsuba, Kenichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Toru; Tobita, Yoshiharu
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10
In order to evaluate the distance for fragmentation of molten core material discharged into the lower sodium plenum during core disruptive accidents in sodium-cooled fast reactors, experiments with simulated molten materials and coolants (water, sodium) was carried out, where an empirical correlation of the distance for fragmentation was developed. The empirical correlation developed by this study showed a good agreement with the measurement results obtained by the present experiments. It was found that in order to well-predict the distance for fragmentation in sodium, thermal phenomena, such as sodium boiling and resultant vapor expansion, needed to be considered.