Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Otsuka, Satoshi; Shizukawa, Yuta; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Onizawa, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.
Journal of Nuclear Science and Technology, 60(3), p.288 - 298, 2023/03
Times Cited Count：4 Percentile：78.52(Nuclear Science & Technology)JAEA has been developing 9Cr-oxide dispersion strengthened (ODS) tempered martensitic steel(TMS) as a candidate material for the fuel cladding tubes of sodium-cooled fast reactors(SFRs). The reliable prediction of in-reactor creep-rupture strength is critical for implementing the 9Cr-ODS TMS cladding tube in the SFR. This study investigated the quantitative correlation between the creep properties of 9Cr-ODS TMS at 700 
C and the dispersions of nanosized oxides by analyzing the creep data and the material's nanostructure. The possibility of deriving a formula for estimating the in-reactor creep properties of 9Cr-ODS TMSs based on an analysis of the nanostructure of neutron-irradiated 9Cr-ODS TMSs was also discussed. The creep properties of 9Cr-ODS TMS at 700 C closely correlated with the dispersion of nanosized oxide particles. The correlation between creep-rupture lives and nanosized oxide particle dispersion was determined using existing creep models. The elucidation of correlation between the stress exponent of secondary creep rate and the nanostructure is essential to enhance future modeling reliability and formulation.
Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi
Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08
Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Kato, Shoichi; Furukawa, Tomohiro; Kaito, Takeji
Journal of Nuclear Materials, 505, p.44 - 53, 2018/07
Times Cited Count：2 Percentile：20.93(Materials Science, Multidisciplinary)A calculation model was constructed to systematically study the effects of environmental conditions (i.e. Cr concentration in sodium, test temperature, axial temperature gradient of fuel pin, and sodium flow velocity) on Cr dissolution behavior. Chromium dissolution was largely influenced by small changes in Cr concentration (i.e. chemical potential of Cr) in liquid sodium in the model calculation. Chromium concentration in sodium coolant, therefore, should be recognized as a critical parameter for the prediction and management of Cr dissolution behavior in the sodium-cooled fast reactor (SFR) core. Because the fuel column length showed no impact on dissolution behavior in the model calculation, no significant downstream effects possibly take place in the SFR fuel cladding tube due to the much shorter length compared with sodium loops in the SFR plant and the large axial temperature gradient. The calculated profile of Cr concentration along the wall-thickness direction was consistent with that measured in BOR-60 irradiation test where Cr concentration in sodium bulk flow was set at 0.07 wt ppm in the calculation.
Ukai, Shigeharu*; Otsuka, Satoshi; Kaito, Takeji; de Carlan, Y.*; Ribis, J.*; Malaplate, J.*
Structural Materials for Generation IV Nuclear Reactors, p.357 - 414, 2017/00
Times Cited Count：70 Percentile：99.33(Energy & Fuels)Oxide dispersion strengthened (ODS) steels are the most promising candidate materials for fuel cladding of generation IV nuclear reactors. The progress and current status for development of ODS/FM(ferrite-martensite) steels conducted mainly in Japan and France are overviewed. The chemical compositions of ODS/FM steels are listed. Fabrication routes of cladding tube are mentioned for ferrite-type ODS steels using recrystallized process and martensite-type one using 
-
phase transformation. The optimized process is identical for both countries. Joining process between cladding and end-plug has been also developed by using the pressurized resistance upset welding method. The improvements brought by ODS/FM steels in high-temperature strength and irradiation resistance are verified.
Nishi, Hiroshi
Fusion Engineering and Design, 81(1-7), p.269 - 274, 2006/02
Times Cited Count：4 Percentile：30.68(Nuclear Science & Technology)Tensile strength of the diffusion bonded joint was as large as that of Alumina dispersion-strengthened copper (DS Cu), however, the Charpy absorbed energy of the joint was considerably lower than that of DS Cu. Instrumented Charpy impact test and slow-bend Charpy test of diffusion bonded joints were performed to clarify the degradation of Charpy absorbed energy. Elasto-plastic analyses were also carried out in order to study the deformation behavior of the tensile and V-notched specimens for joints. As the results, the fracture behaviors of the impact and slow-bend tests were almost the same. Elasto-plastic analyses showed that the maximum strain occurred at the DS Cu apart from the interface for tensile specimen, however, the strain concentrated at the DS Cu near the interface for the notched specimen. This strain concentration arose from the mechanical heterogeneity between stainless steel and DS Cu in the bonded zone and attributed to the degradation of the absorbed energy of the joints.
Yamada, Hirokazu*; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, G.*; Nagao, Yoshiharu; Sato, Satoshi; Mori, Kensuke*
Journal of Nuclear Materials, 335(1), p.33 - 38, 2004/10
Times Cited Count：8 Percentile：48.81(Materials Science, Multidisciplinary)no abstracts in English
Saito, Shigeru; Fukaya, Kiyoshi*; Ishiyama, Shintaro; Sato, Kazuyoshi
Journal of Nuclear Materials, 307-311(2), p.1542 - 1546, 2002/12
Times Cited Count：37 Percentile：89.18(Materials Science, Multidisciplinary)no abstracts in English
Hatano, Toshihisa; Kuroda, Toshimasa*; Barabash, V.*; Enoeda, Mikio
Journal of Nuclear Materials, 307-311(2), p.1537 - 1541, 2002/12
Times Cited Count：4 Percentile：29.25(Materials Science, Multidisciplinary)no abstracts in English
Uchida, Munenori*; Ishitsuka, Etsuo; Hatano, Toshihisa; Barabash, V.*; Kawamura, Hiroshi
Journal of Nuclear Materials, 307-311(Part2), p.1533 - 1536, 2002/12
Times Cited Count：3 Percentile：23.41(Materials Science, Multidisciplinary)no abstracts in English
Nishi, Hiroshi; Araki, Toshimitsu*
Journal of Nuclear Materials, 283-287(Part.2), p.1234 - 1237, 2000/12
Times Cited Count：17 Percentile：72(Materials Science, Multidisciplinary)no abstracts in English
Suzuki, Satoshi; Ezato, Koichiro; Sato, Kazuyoshi; Nakamura, Kazuyuki; Akiba, Masato
Fusion Engineering and Design, 49-50, p.343 - 348, 2000/11
Times Cited Count：5 Percentile：37.66(Nuclear Science & Technology)no abstracts in English
Hatano, Toshihisa; Goto, Masahiro*; Yamada, Tetsuji*; Nomura, Yuichiro*; Saito, Masakatsu*
Fusion Engineering and Design, 49-50, p.207 - 212, 2000/11
Times Cited Count：3 Percentile：26.42(Nuclear Science & Technology)no abstracts in English
Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Eto, Motokuni; Akiba, Masato
JAERI-Research 2000-006, p.57 - 0, 2000/02
JAERI-Research-2000-006.pdf:20.86MB
no abstracts in English
Tsuchiya, Kunihiko; Nakamichi, Masaru; Kawamura, Hiroshi
Effects of Radiation on Materials (ASTM STP 1366), p.988 - 999, 2000/00
no abstracts in English
Hatano, Toshihisa; Sato, Satoshi; Hashimoto, T.*; Kitamura, Kazunori*; Furuya, Kazuyuki; Kuroda, Toshimasa*; Enoeda, Mikio; Takatsu, Hideyuki
Journal of Nuclear Science and Technology, 35(10), p.705 - 711, 1998/10
Times Cited Count：1 Percentile：15.03(Nuclear Science & Technology)no abstracts in English
Sato, Satoshi; Kuroda, Toshimasa*; Hatano, Toshihisa; Furuya, Kazuyuki; *; Takatsu, Hideyuki
Fusion Engineering and Design, 39-40, p.609 - 614, 1998/00
Times Cited Count：10 Percentile：63.67(Nuclear Science & Technology)no abstracts in English
Nishi, Hiroshi; Araki, Toshimitsu*; Eto, Motokuni
Fusion Engineering and Design, 39-40, p.505 - 511, 1998/00
Times Cited Count：27 Percentile：87.41(Nuclear Science & Technology)no abstracts in English
Hatano, Toshihisa; Kuroda, Toshimasa*; Iwadachi, Takaharu*; Osaki, Toshio*; Enoeda, Mikio; Takatsu, Hideyuki
Fusion Technology 1998, 1, p.97 - 100, 1998/00
no abstracts in English
Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Eto, Motokuni; Sato, Kazuyoshi; Akiba, Masato
Fusion Technology 1998, Vol.1, p.113 - 116, 1998/00
no abstracts in English
