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Omasa, Yoshinori*; Takagi, Shigeyuki*; Toshima, Kento*; Yokoyama, Kaito*; Endo, Wataru*; Orimo, Shinichi*; Saito, Hiroyuki*; Yamada, Takeshi*; Kawakita, Yukinobu; Ikeda, Kazutaka*; et al.
Physical Review Research (Internet), 4(3), p.033215_1 - 033215_9, 2022/09
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Nogiwa, Kimihiro; Nishimura, Akihiko; Yokoyama, Atsushi; Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Okubo, Tadakatsu*; Hono, Kazuhiro*
Journal of Nuclear Materials, 417(1-3), p.201 - 204, 2011/10
Times Cited Count：8 Percentile：53.23(Materials Science, Multidisciplinary)Du se 9Cr-ODS (oxide dispersion-strengthened) steel consisting of residual-
ferrite and prime martensite has excellent high-temperature strength. This study describes the microstructure of dual-phase 9Cr-ODS steels characterized by atom-probe tomography in order to compare oxide-particle dispersion states in each phase. This revealed that nano-size oxide particles were of the same chemical composition and that their mean size was about 3 nm in each phase. On the other hand, the number density in the residual- phase was about four times higher than that of the prime phase. These results indicate that the dense distribution of the oxide particles in the residual- phase contribute to the excellent high-temperature strength of 9Cr-ODS steel.
Nakabachi, Kaito*; Yokoyama, Kenichi*; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi
no journal, ,
The effects of thermal aging on hydrogen embrittlement behavior were investigated for Zr/Ta/SUS explosive joints used in spent nuclear fuel reprocessing plants, especially at the Zr/Ta interface, which is considered to be susceptible to hydrogen embrittlement. The results showed that brittle fracture occurred at the Zr/Ta interface when the hydrogen concentration of Ta was over 20 ppm, and the ductility of the specimens with such hydrogen concentration was recovered when the hydrogen concentration of Ta decreased below 10 ppm by thermal aging at 300
C for 1000 hr, while the hydrogen concentration was reduced to The ductility of the Ta/Zr interface was not recovered without significant decrease in hydrogen concentration. These results suggest that the hydrogen concentration in Ta at the Ta/Zr interface affects the hydrogen embrittlement behavior of the explosion joint and that the hydrogen state changes with thermal aging.
Nakabachi, Kaito*; Yokoyama, Kenichi*; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Abe, Hitoshi
no journal, ,
To investigate the effects of strain and thermal aging on hydrogen embrittlement of Ta, tensile tests and internal friction measurements were carried out on cold rolled Ta after hydrogen charging and thermal aging. As a result, the internal friction increased with cold rolling. The internal friction was further increased by hydrogen charging. However, the internal friction of these specimens decreased after thermal aging. These results suggest that aging causes the annihilation or re-arrangement of defects or a change in the hydrogen trapping. In addition, the ductility of the hydrogen-charged and cold-rolled specimens decreased after thermal aging, suggesting that the change in hydrogen and defect state due to thermal aging affects the mechanical properties.
