Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Saito, Shigeru; Kikuchi, Kenji*; Hamaguchi, Dai; Tezuka, Masao*; Miyagi, Masanori*; Kokawa, Hiroyuki*; Watanabe, Seiichi*
Journal of Nuclear Materials, 431(1-3), p.91 - 96, 2012/12
Times Cited Count：16 Percentile：75.52(Materials Science, Multidisciplinary)To evaluate lifetime of structural materials for ADS, corrosion tests in LBE have been done at JAEA. The corrosion test was performed by using JAEA lead-bismuth flowing loop (JLBL-1). Experimental condition was as follows; The temperature of high and low temperature parts of the loop were 450
C and 350C, respectively. Flowing velocity at the test specimens was about 1m/s. Plate type SS316L-BM and SS316L-GBEM were used as a specimens. After the 3,000 hours operation, the test specimens were cut and macroscopic observation was carried out. The result showed that both materials were intensively eroded. Corrosion depth and LBE penetration through grain boundaries of GBEM were smaller than these of 316SS-BM.
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.
Yano, Yasuhide; Yoshitake, Tsunemitsu; Yamashita, Shinichiro; Akasaka, Naoaki; Onose, Shoji; Watanabe, Seiichi*; Takahashi, Heishichiro
Journal of Nuclear Science and Technology, 44(12), p.1535 - 1542, 2007/12
Times Cited Count：12 Percentile：63.69(Nuclear Science & Technology)The effects of fast neutron irradiation have been investigated on the mechanical properties of 11Cr-0.5Mo-2W, Nb, V ferritic/martensitic (F/M) stainless steel (PNC-FMS) and 10.5Cr-1.5Mo, Nb, V F/M stainless steel (HT9M) claddings, especially tensile and transient burst properties. These two F/M claddings were irradiated in the experimental fast reactor JOYO using the PFB090 fuel test assembly. Post irradiation tensile and temperature-transient-to-burst tests were carried out for defueled cladding specimens. The results of mechanical tests for PNC-FMS cladding showed that there was no significant degradation in tensile and transient burst strengths even after fast neutron irradiation. However, these strengths for HT9M cladding tended to shift to lower values than those of as-received specimens. This different behavior of tensile and transient burst strengths was attributed to martensite structural stability which was related to the stable solid solution elements.
Yano, Yasuhide; Yamashita, Shinichiro; Akasaka, Naoaki; Watanabe, Seiichi*; Takahashi, Heishichiro
Proceedings of 9th China-Japan Symposium on Materials for Advanced Energy Systems and Fission & Fusion Engineering jointed with CAS-JSPS Core-university Program Seminar on Fusion Materials, System and Design Integration, p.2 - 5, 2007/10
Ferritic/martensitic (F/M) steels are expected to be prospective not only for the long life core material of fast reactors but also for the blanket materials of fusion reactor because of their superior swelling resistance. Japan Atomic Energy Agency has developed a 11Cr-0.5Mo-2W-V, Nb F/M steel (PNC-FMS) for core materials of next fast reactor. In order to evaluate the effect of structural change due to irradiation on mechanical properties of PNC-FMS, neutron irradiations were carried out between 773 and 1013 K to doses of from 11 to 102 dpa in JOYO reactor. Post irradiation tensile tests were performed at 773-1013 K with a strain rate of 0.5
10
/s. The stability of microstructures under irradiation was also compared with those of electron irradiation using HVEM.
Akasaka, Naoaki; Tanaka, Kosuke; Onuki, Somei*; Takahashi, Heishichiro*; Watanabe, Seiichi*
Effects of Radiation on Materials (ASTM STP 1447), 516- Pages, 2003/00
Radiation-enhanced diffusion is one of the most important parameters for the understanding and modeling the effect of radiation on materials. In this study, the average effective interdiffusion coefficients of Ni and Cr in pure iron, an Fe-15Cr-15ni model
Ishino, Masahiko; Yoda, Osamu; Haishi, Yasuyuki*; Arimoto, Fumiko*; Takeda, Mitsuhiro*; Watanabe, Seiichi*; Onuki, Somei*; Abe, Hiroaki
Japanese Journal of Applied Physics, Part 1, 41(5A), p.3052 - 3056, 2002/05
Times Cited Count：12 Percentile：45.92(Physics, Applied)no abstracts in English
Takeda, Mitsuhiro*; Onuki, Somei*; Watanabe, Seiichi*; Abe, Hiroaki; Naramoto, Hiroshi; P.R.Okamoto*; N.Q.Lam*
Mat. Res. Soc. Symp. Proc., 540, p.37 - 42, 1999/00
no abstracts in English
Namba, Hideki; Aoki, Yasushi; Shibata, Hiromi*; Yoshida, Yoichi*; Side, Y.*; Tagawa, S.*; Nagai, Shiro; Watanabe, Hiroshi
Tokyo Daigaku Genshiryoku Kenkyusho Sogo Senta Heisei-Gan-Nendo Jushosha Kanribu Nempo, p.21 - 23, 1989/00
no abstracts in English
Tanikawa, Ryusuke*; Sakaguchi, Norihito*; Watanabe, Seiichi*; Kinoshita, Hiroshi*; Kokawa, Hiroyuki*; Kawai, Masayoshi*; Yamashita, Shinichiro
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Yano, Yasuhide; Endo, Masaki*; Sakaguchi, Norihito*; Watanabe, Seiichi*; Miyagi, Masanori*; Sato, Shinya*; Sato, Yutaka*; Kokawa, Hiroyuki*; Kawai, Masayoshi*
no journal, ,
no abstracts in English
Yano, Yasuhide; Yamashita, Shinichiro; Endo, Masaki*; Sakaguchi, Norihito*; Watanabe, Seiichi*; Miyagi, Masanori*; Oyamada, Tetsuya*; Sato, Shinya*; Sato, Yutaka*; Kokawa, Hiroyuki*; et al.
no journal, ,
no abstracts in English
Endo, Masaki*; Sakaguchi, Norihito*; Kinoshita, Hiroshi*; Watanabe, Seiichi*; Yamashita, Shinichiro; Yano, Yasuhide; Kawai, Masayoshi*
no journal, ,
no abstracts in English
Endo, Masaki*; Sakaguchi, Norihito*; Kinoshita, Hiroshi*; Watanabe, Seiichi*; Kokawa, Hiroyuki*; Yamashita, Shinichiro; Yano, Yasuhide; Kawai, Masayoshi*
no journal, ,
no abstracts in English
Yano, Yasuhide; Yamashita, Shinichiro; Akasaka, Naoaki; Takahashi, Heishichiro; Watanabe, Seiichi*
no journal, ,
no abstracts in English
Sakaguchi, Norihito*; Tanikawa, Ryusuke*; Endo, Masaki*; Watanabe, Seiichi*; Kokawa, Hiroyuki*; Kawai, Masayoshi*; Yamashita, Shinichiro; Yano, Yasuhide
no journal, ,
no abstracts in English
Sakaguchi, Norihito*; Endo, Masaki*; Kinoshita, Hiroshi*; Watanabe, Seiichi*; Kokawa, Hiroyuki*; Yano, Yasuhide; Yamashita, Shinichiro; Kawai, Masayoshi*
no journal, ,
no abstracts in English
Yamashita, Shinichiro; Yano, Yasuhide; Sakaguchi, Norihito*; Watanabe, Seiichi*; Kokawa, Hiroyuki*; Kawai, Masayoshi*
no journal, ,
In this study, grain boundary character distribution (GBCD)-optimized Type 316 corresponding austenitic stainless steel has been developed as a nuclear material for next generation energy systems. Some of steel sheets were cold-rolled additionally for making the GBCD-optimized and cold-worked (GBCD+CW) specimens. These specimens, including as GBCD-optimized, were examined tensile strength property, phase stability during high temperature thermal ageing and irradiation property. The GBCD was assessed by orientation imaging microscopy (OIM) and also the microstructure by TEM. The OIM results showed that the average grain sizes and the frequencies of CSL boundaries in the typical specimens were 40-47 
m and more than 70%. Totally considered all experimental results, it was indicated that it is highly possible to develop high performance FBR core material by means of grain boundary engineering.
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.
