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Nakamichi, Masaru; Kim, Jae-Hwan; Munakata, Kenzo*; Shibayama, Tamaki*; Miyamoto, Mitsutaka*
Journal of Nuclear Materials, 442(1-3), p.S465 - S471, 2013/11
Times Cited Count：11 Percentile：64.2(Materials Science, Multidisciplinary)Ito, Tatsuya*; Yamauchi, Yuji*; Hino, Tomoaki*; Shibayama, Tamaki*; Nobuta, Yuji*; Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato
Journal of Nuclear Materials, 417(1-3), p.1147 - 1149, 2011/10
Times Cited Count：13 Percentile：69.64(Materials Science, Multidisciplinary)Shibayama, Tamaki*; Nakamichi, Masaru; Miyamoto, Mitsutaka*; Kuga, Noriyoshi*; Dorn, C. K.*; Knudson, T.*
Purazuma, Kaku Yugo Gakkai-Shi, 87(4), p.259 - 267, 2011/04
no abstracts in English
Yano, Yasuhide; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Shibayama, Tamaki*; Watanabe, Seiichi*; Takahashi, Heishichiro
Journal of Nuclear Materials, 398(1-3), p.59 - 63, 2010/03
Times Cited Count：10 Percentile：56.32(Materials Science, Multidisciplinary)The in-reactor creep rupture tests of 11Cr-0.5Mo-2W, V, Nb F/M steel were carried out in the temperature range from 823 to 943 K using Materials Open Test Assembly in the Fast Flux Test Facility and tensile and temperature-transient-to-burst specimens were irradiated in the experimental fast reactor JOYO at temperatures between 693 to 1013 K to fast neutron doses ranging from 3.5 to 102 dpa. The results of post irradiation mechanical tests showed that there was no significant degradation in tensile and transient burst strengths even after neutron irradiation below 873 K, but that there was significant degradation in both strengths at neutron irradiation above 903 K. On the other hand, the in-reactor creep rupture times were equal or greater than those of out-reactor creep even after neutron irradiation at all temperatures. This creep rupture behavior was different from that of tensile and transient burst specimens.
Nakamichi, Masaru; Shibayama, Tamaki*; Tatenuma, Katsuyoshi*; Yonehara, Kazuo
Proceedings of 9th IEA International Workshop on Beryllium Technology (BeWS-9), p.40 - 43, 2009/09
Shibayama, Tamaki*; Kishimoto, Hirotatsu*; Koyama, Akira*; Yano, Yasuhide
Materia, 47(12), P. 628, 2008/12
no abstracts in English
Mishima, Yoshinao*; Yoshida, Naoaki*; Kawamura, Hiroshi; Ishida, Kiyohito*; Hatano, Yuji*; Shibayama, Tamaki*; Munakata, Kenzo*; Sato, Yoshiyuki*; Uchida, Munenori*; Tsuchiya, Kunihiko; et al.
Journal of Nuclear Materials, 367-370(2), p.1382 - 1386, 2007/08
Times Cited Count：26 Percentile：84.1(Materials Science, Multidisciplinary)no abstracts in English
Tsuchiya, Kunihiko; Kawamura, Hiroshi; Mishima, Yoshinao*; Yoshida, Naoaki*; Tanaka, Satoru*; Uchida, Munenori*; Ishida, Kiyohito*; Shibayama, Tamaki*; Munakata, Kenzo*; Sato, Yoshiyuki*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 83(3), p.207 - 214, 2007/03
no abstracts in English
Kawamura, Hiroshi; Tsuchiya, Kunihiko; Mishima, Yoshinao*; Yoshida, Naoaki*; Munakata, Kenzo*; Ishida, Kiyohito*; Hatano, Yuji*; Shibayama, Tamaki*; Sato, Yoshiyuki*; Uchida, Munenori*; et al.
INL/EXT-06-01222, p.1 - 7, 2006/02
no abstracts in English
TiShibayama, Tamaki*; Nakamichi, Masaru*; Uchida, Munenori*; Kawamura, Hiroshi; Kinoshita, Hiroshi*; Kiyanagi, Yoshiaki*; Takahashi, Heishichiro*; Nomura, Naoyuki*
JAERI-Conf 2004-006, p.216 - 219, 2004/03
no abstracts in English
Uchida, Yosuke*; Nagai, Yoshiyasu*; Suda, Takanori*; Hashimoto, Naoyuki*; Onuki, Somei*; Shibayama, Tamaki*; Yamashita, Shinichiro; Akasaka, Naoaki
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Otsuka, Satoshi; Watanabe, Masashi*; Uchida, Yosuke*; Suda, Takanori*; Hashimoto, Naoyuki*; Onuki, Somei*; Shibayama, Tamaki*
no journal, ,
no abstracts in English
Oka, Hiroshi*; Watanabe, Masashi*; Hashimoto, Naoyuki*; Kinoshita, Hiroshi*; Shibayama, Tamaki*; Onuki, Somei*; Yamashita, Shinichiro; Otsuka, Satoshi
no journal, ,
In this study, ODS austenitic stainless steels based on an advanced SUS316 steel has been developed by mechanically alloying (MA) and hot extrusion with the addition of minor alloying elements. The chemical composition of ODS316 is Fe-16Cr-13Ni-0.35Y
O
-0.1Ti-0.6Hf. Thin foils for transmission electron microscope (TEM) examination were prepared with electro-polishing. HRTEM and EDS were used for characterization of oxide particles in ODS-316, especially precipitation behavior and chemical composition before and after heat treatment. Microstructural observation revealed that the ODS316 has grains of 0.5-1
m in diameter and complex oxides (Y-Ti-Hf-O) distributed in matrix. The strain contrast was observed around oxide particles, suggesting coherency of oxide particles. HR-TEM observation revealed that a part of faceted particles have coherency. Interface between matrix and oxide particles after heat treatment was also carefully investigated.
Sekio, Yoshihiro; Yamashita, Shinichiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Tokita, Shun*; Fujii, Hiromichi*; Sato, Yutaka*; Kokawa, Hiroyuki*
no journal, ,
In order to improve ductility loss by helium embrittlement (or grain boundary embrittlement) induced under high temperature and neutron irradiation dose in nickel alloys which are expected to have high-temperature phase stability under non-irradiation, the grain boundary engineering was applied for NIMONIC PE16 to enhance the grain boundary strength. And, its high temperature tensile properties under non-irradiation were investigated as the first approach. As a result, the temperature dependence of the yield stress in the grain boundary engineered (GBE) PE16 was similar to that in NIMINIC PE16, but the yield stress was slightly lower and the uniform elongation was slightly higher at each temperature in GBE PE16 comparing to NIMINIC PE16. This would be caused by grain coarsening due to some heat treatments. If the gain size of GBE PE16 is optimized, tensile properties of GBE PE16 would be the same or more than that of NIMONIC PE16.
Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*
no journal, ,
Recent grain boundary structure studies have shown that optimal distribution of a high frequency of coincidence site lattice boundaries and consequent discontinuity of random boundary network in the material is one of very effective methods to enhance the intergranular corrosion resistance. This advantageous property, one of important ones for structural material of nuclear reactor, can be obtained through simple thermomechanical treatment process without any change of original chemical composition. In this study, grain boundary character distribution(GBCD)-optimized Ni-based alloy (PE16) has been developed as a prospective high-performance nuclear reactor material by grain boundary engineering processing, and then tensile behavior of GBCD-optimized Ni-based alloy was investigated to evaluate the effect of grain boundary engineering processing on mechanical property. The result of tensile test at room temperature showed that tensile strength of GBCD-optimized PE16 was somewhat lower than that of as-received PE16. However, no significant change was confirmed in elongation property. Details on tensile behavior analyses would be discussed in the conference.
Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Tokita, Shun*; Fujii, Hiromichi*; Sato, Yutaka*; Kokawa, Hiroyuki*
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Sekio, Yoshihiro; Sakaguchi, Norihito*; Shibayama, Tamaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*
no journal, ,
Wakai, Eiichi; Noto, Hiroyuki*; Kano, Sho*; Makimura, Shunsuke*; Ishida, Taku*; Shibayama, Tamaki*
no journal, ,
It is important for materials and devices that are subjected to high-energy particle beams to be able to withstand high thermal loads and irradiation. In this study, we investigated the irradiation resistance of nanoparticle-dispersed W (tungsten) and 1.1 wt% TiC-doped nanoparticle-dispersed W materials with a grain size of 1-2 m and high strength, which were fabricated by mechanical alloying and high-temperature isostatic sintering. The irradiation was carried out at the HIT ion irradiation facility at the University of Tokyo at 773 K to about 0.7 dpa using tungsten ions, which are self ions. After the irradiation, the irradiation resistance of the materials was measured using a nanoindenter, and the results showed that the former material changed, while the latter material did not show any irradiation hardening.
Iwamoto, Yosuke; Wakai, Eiichi; Nakagawa, Yuki*; Shibayama, Tamaki*
no journal, ,
In order to develop a new non-destructive inspection technique to accurately measure defects inside materials in radiation fields such as nuclear power plants, accelerators, and aerospace, we have measured the time evolution of electrical resistivity of copper, aluminum, and niobium metal wires and silicon plates by irradiating a pulsed laser beam (20 Hz) at Hokkaido University. From the obtained increase in electrical resistivity, the number of atomic vacancies and Frenkel pairs (FPs) of interstitial atoms and the dislocation density of the formed FPs were estimated, assuming that the atoms were ejected by the laser due to electronic excitation. From this experiment, it was estimated that for metals, the amount of defects formed inside the material increased with the increase in the irradiation dose. On the other hand, for silicon, the electrical resistivity was found to decrease due to the electronic transition to the valence band caused by laser irradiation.
Toyota, Kodai; Wakai, Eiichi; Onizawa, Takashi; Shibayama, Tamaki*; Nakagawa, Yuki*
no journal, ,
no abstracts in English
