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Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.
Tetsu To Hagane, 111(5), p.246 - 262, 2025/04
Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Sakasegawa, Hideo; Nakajima, Motoki*; Kato, Taichiro*; Nozawa, Takashi*; Ando, Masami*
Materials Today Communications (Internet), 40, p.109659_1 - 109659_8, 2024/08
Times Cited Count：1 Percentile：48.91(Materials Science, Multidisciplinary)Nanometric oxide particles play an important role in improving the creep property of Oxide Dispersion Strengthened (ODS) steels. In our previous research, we examined a microstructural feature known as prior particle boundary (PPB). PPB refers to the surface of mechanically alloyed (MA) powders before consolidation. We revealed that the ODS steel with fine PPBs produced from smaller MA powders, exhibited shorter creep rupture times, compared to that with coarse PPBs produced from larger MA powders. The size of MA powders had an impact on the creep property. In this study, we examined the shape of MA powders, which were non-spherical shapes. Such shapes have the potential to induce anisotropic creep behavior. We conducted small punch creep tests on specimens with two different orientations to study the possible anisotropy. The results revealed that the creep rupture times varied depending on the orientation of specimen, thus indicating anisotropic creep property.
Oka, Hiroshi; Tanno, Takashi; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Tachi, Yoshiaki
Journal of Nuclear Materials, 547, p.152833_1 - 152833_7, 2021/04
Times Cited Count：10 Percentile：73.48(Materials Science, Multidisciplinary)In order to evaluate the stability of nano-sized oxide particles and matrix structure of ODS cladding tube, which are the determinants of their high temperature strength, the microstructural observation was carried out after internal pressurized creep test at 700C for over 45,000 hours. The specimens were the as-received and crept specimens of 9Cr-ODS steel with tempered martensite and 12Cr-ODS steel with recrystallized ferrite. Small platelet was cut out from the crept pressurized tube, then thinned to foil. Microstructural observation was conducted with TEM JEOL 2010F. As a result of the observation, it was confirmed that the size and number density of the nano-sized particles were almost unchanged even after the creep test. In addition, the tempered martensite structure, which is one of the determinants of the creep strength of 9Cr-ODS steel, was not significantly different between the as-received and crept specimen, indicating the stability of their matrix structure.
Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi
Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12
Times Cited Count：1 Percentile：9.26(Nuclear Science & Technology)The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.
Shibata, Taiju; Baba, Shinichi; Yamaji, Masatoshi*; Sumita, Junya; Ishihara, Masahiro
Nihon Kikai Gakkai M&M 2004 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.407 - 408, 2004/00
no abstracts in English
Kurata, Yuji; Saito, T.*; Tsuji, Hirokazu; Takatsu, T.*; Shindo, Masami; Nakajima, Hajime
JSME International Journal, Series A, 45(1), p.104 - 109, 2002/01
no abstracts in English
Kurata, Yuji; Saito, T.*; Tsuji, Hirokazu; Takatsu, T.*; Shindo, Masami; Nakajima, Hajime
Proceedings of the 7th International Conference on Creep and Fatigue at Elevated Temperatures (CREEP7), p.93 - 99, 2001/06
no abstracts in English
Kurata, Yuji; Saito, T.*; Tsuji, Hirokazu; Takatsu, T.*; Shindo, Masami; Nakajima, Hajime
Nihon Gakujutsu Shinkokai Genshiro Zairyo Dai-122-Iinkai Shiryoshu, p.279 - 282, 2000/11
no abstracts in English
; Saito, T.*; Tsuji, Hirokazu; Takatsu, T.*; Shindo, Masami; Nakajima, Hajime
JAERI-Research 97-032, 20 Pages, 1997/05
JAERI-Research-97-032.pdf:1.78MB
no abstracts in English
; Tsuji, Hirokazu; Shindo, Masami; Nakajima, Hajime
Journal of Nuclear Materials, 246(2-3), p.196 - 205, 1997/00
Times Cited Count：19 Percentile：79.53(Materials Science, Multidisciplinary)no abstracts in English
Kurata, Yuji; ; ; Shindo, Masami; Nakajima, Hajime;
Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 36(2), p.149 - 156, 1995/07
no abstracts in English
Nakajima, Hajime; ; Watanabe, Katsutoshi; Kondo, Tatsuo
Computer Aided Innovation of New Materials, p.827 - 830, 1991/00
no abstracts in English
; ; ;
Nihon Kikai Gakkai Rombunshu, A, 52(477), p.1228 - 1231, 1986/00
no abstracts in English
Oka, Hiroshi; Ikusawa, Yoshihisa; Otsuka, Satoshi; Kaito, Takeji
no journal, ,
