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Harjo, S.; Gong, W.; Kawasaki, Takuro; Morooka, Satoshi; Yamashita, Takayuki*
ISIJ International, 62(10), p.1990 - 1999, 2022/10
Times Cited Count：1 Percentile：0(Metallurgy & Metallurgical Engineering)Nishimura, Hayato*; Hojo, Tomohiko*; Ajita, Saya*; Shibayama, Yuki*; Koyama, Motomichi*; Saito, Hiroyuki*; Shiro, Ayumi*; Yasuda, Ryo*; Shobu, Takahisa; Akiyama, Eiji*
Tetsu To Hagane, 107(9), p.760 - 768, 2021/09
Times Cited Count：0 Percentile：0(Metallurgy & Metallurgical Engineering)no abstracts in English
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.
Saito, Kei*; Hirade, Tetsuya; Takai, Kenichi*
Metallurgical and Materials Transactions A, 50(11), p.5091 - 5102, 2019/11
Times Cited Count：29 Percentile：84.98(Materials Science, Multidisciplinary)An attempt was made to separate and identify hydrogen peaks desorbed from lattice defects formed by plastic-strain in the presence of hydrogen in tempered martensitic steel showing quasi-cleavage fracture using thermal desorption spectroscopy from a low temperature (L-TDS) and positron annihilation spectroscopy (PAS). The L-TDS results made it possible to separate two peaks, namely, that of the original desorption and also that of new desorption. The PAS results revealed that the new desorption obtained by L-TDS corresponded to vacancy-type defects. Hydrogen enhanced vacancy-type defect concentration, approximately 10
order in terms of atomic ratio, formed within 1.5 mm from the fracture surface, These results indicate that the accumulation of excess vacancy-type defects enhanced by hydrogen in the local region can lead to nanovoid nucleation and coalescence in plastic deformation, resulting in quasi-cleavage fracture of tempered martensitic steel.
Ebihara, Kenichi; Saito, Kei*; Takai, Kenichi*
"Suiso Zeika No Kihon Yoin To Tokusei Hyoka" Kenkyukai Hokokusho, p.57 - 61, 2018/09
no abstracts in English
Tsuchida, Yutaka*; Ebihara, Kenichi
Tetsu To Hagane, 103(11), p.653 - 659, 2017/11
Times Cited Count：2 Percentile：11.25(Metallurgy & Metallurgical Engineering)A single peak in thermal desorption profiles of hydrogen, which are measured in low-temperature thermal desorption spectrometry (L-TDS) for a very thin plate specimen of tempered martensitic steel, was reproduced successfully by the superposition of two Gaussian distributions. Then, the parameters concerning the detrapping rate constants for both peaks, which are trap energy and pre-exponential factor, were calculated using the Choo-Lee plot. We confirmed that Kissinger model incorporating the obtained parameters could simulate the two peaks. In addition, we reproduced the single peak well using the reaction-diffusion equation incorporating the obtained parameters and the appropriate trap site concentration. From the results, we interpreted that the one peak corresponds to dislocation and the other to grain-boundary.
Harjo, S.; Kawasaki, Takuro; Gong, W.; Aizawa, Kazuya
Journal of Physics; Conference Series, 746(1), p.012046_1 - 012046_7, 2016/10
Times Cited Count：5 Percentile：88.68(Physics, Nuclear)Ebihara, Kenichi; Saito, Kei*; Takai, Kenichi*
"Suiso Zeika No Kihon Yoin To Tokusei Hyoka Kenkyukai Chukan Hokokukai" Shimposium Yokoshu (USB Flash Drive), p.30 - 35, 2016/09
no abstracts in English
C in JMTRWakai, Eiichi; Jitsukawa, Shiro; Tomita, Hideki*; Furuya, Kazuyuki; Sato, Michitaka*; Oka, Keiichiro*; Tanaka, Teruyuki*; Takada, Fumiki; Yamamoto, Toshio*; Kato, Yoshiaki; et al.
Journal of Nuclear Materials, 343(1-3), p.285 - 296, 2005/08
Times Cited Count：48 Percentile：93.78(Materials Science, Multidisciplinary)The dependence of helium production on radiation-hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel doped with 
B, 
B and B+B irradiated at 250C to 2.2 dpa. The total amounts of doping boron were about 60 massppm. The range of He concentration produced in the specimens was from about 5 to about 300 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He
irradiation was also performed to implant about 85 appm He atoms at 120C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain DBTT. Radiation-hardening of the neutron-irradiated specimens increased slightly with increasing He production. The 100 MPam
DBTT for the F82H+B, F82H+B+B, and F82H+B were 40, 110, and 155C, respectively. The shifts of DBTT due to He production were evaluated as about 70C by 150 appmHe and 115C by 300 appmHe. The DBTT shift in the small punch testing was evaluated as 50C.
Miwa, Yukio; Jitsukawa, Shiro; Yonekawa, Minoru
Journal of Nuclear Materials, 329-333(Part2), p.1098 - 1102, 2004/08
Times Cited Count：11 Percentile：58.46(Materials Science, Multidisciplinary)Fatigue properties were examined on a reduced activation ferritic/martensitic steel, and preliminary results were presented. F82H steel was irradiated at 523 K to 3.8 dpa, and then fatigue-tested at 298-573 K in vacuum with total strain range of 0.4-1.0%. Effect of irradiation on fatigue lives was observed on test at 298 K with total strain range of 0.4%. The fatigue life of irradiated specimen was reduced to about 1/7 of unirradiated specimen. The reduction of the fatigue life was attributed to the occurrence of channel fracture. Effect of test temperature was discussed.
Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro
Purazuma, Kaku Yugo Gakkai-Shi, 80(7), p.551 - 557, 2004/07
Environmental assisted cracking (EAC) is one of the materials issues for the reactor core components of light water power reactors (LWRs). Much experience and knowledge have been obtained about EAC in LWR field. They will be useful to manage the EAC of water-cooled blanket systems of the fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC of a water-cooled blanket does not seem to be critical issues. However some uncertainties about influences of water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations investigating for such the uncertainties are discussed.
Wakai, Eiichi; Miwa, Yukio; Hashimoto, Naoyuki*; Robertson, J. P.*; Klueh, R. L.*; Shiba, Kiyoyuki; Abiko, Kenji*; Furuno, Shigemi*; Jitsukawa, Shiro
Journal of Nuclear Materials, 307-311(Part.1), p.203 - 211, 2002/12
Times Cited Count：26 Percentile：82.29(Materials Science, Multidisciplinary)no abstracts in English
Sawai, Tomotsugu; Shiba, Kiyoyuki; Hishinuma, Akimichi
Journal of Nuclear Materials, 283-287(Part.1), p.657 - 661, 2000/12
Times Cited Count：43 Percentile：90.73(Materials Science, Multidisciplinary)no abstracts in English
Shiba, Kiyoyuki; Klueh, R. L.*; Miwa, Yukio; Igawa, Naoki; Robertson, J. P.*
Fusion Materials Semiannual Progress Report (DOE/ER-0313/28), p.131 - 135, 2000/06
no abstracts in English
Mizuta, Shunji; ;
JNC TN9400 2000-048, 28 Pages, 2000/04
JNC-TN9400-2000-048.pdf:0.64MB
ODS (Oxide Dispersion Strengthened) ferritic-martainsitic steels are one of the most prospective cladding materials for advanced fast breeder reactors, since they are expected to have excellent swelling resistance and superior high temperature strength due to the finely distributed stable oxide particles(Y
O
). Properties and the tentative strength equations for ODS ferritic-martainsitic were proposed on the basis of the latest data to apply to the feasibility study of the sodium coolant MOX fuel plant. The items of equations are follows. (1)creep rupture strength (2)correction factor of creep rupture strength (in Na and in reactor) (3)outer surface eorrosion (Na) (4)inner surface corrosion (in MOX fuel pin) (5)thermal conductivity
;
JNC TN9400 2000-035, 164 Pages, 2000/03
JNC-TN9400-2000-035.pdf:3.67MB
High Strength Ferritic/Martensitic Steel (PNC-FMS : 0.12C-11Cr-0,5Mo-2W-0.2V-0.05Nb), developed by JNC, is one of the candidate materials for the long-life core of large-scale fast breeder reactor. Ductile brittle transition temperature (DBTT) was tentatively determined in 1992 in material design base standard of PNC-FMS. Howevcr, specimen size effect on impact property and upper shelf energy (USE) have not been evaluated. ln this study, effects of specimen size, thermal aging and neutron irradiation on the charpy impact property of PNC-FMS were evaluated, using together with recently obtained data. The design value of USE and DBTT as fabricated and each correlation of aging and irradiation effects were determined. The results are summarized as follows. (1)lt was found that USE is related to (Bb) as USE=m(Bb)
, where B is specimen width, b is ligament size and both m and n are constant. For PNC-FMS, n value is equal to 1.4. It's possible to determine n value from USE (J) for full size specimen using the correlation: n=1.38
10
USE + 1.20. (2)lt was clarified that DBTT is correlated with (BKt) as DBTT=p(log
BKt)+q, where Kt is elastic stress concentration factor and both p and q are constant. For PNC-FMS, the correlation is as follows: DBTT=119(logBKt)-160. (3)DBTT estimated at the irradiation temperature from 350 to 650 C for sub size specimen (width and height are 3 and 10 mm, respectively), was below 180 C, based on the design value of DBTT as fabricated and each correlation of aging and irradiation effects.
Hishinuma, Akimichi; Koyama, Akira*; R.L.Klueh*; D.S.Gelles*; Ehrlich, K.*; W.Dietz*
Journal of Nuclear Materials, 258-263, p.193 - 204, 1998/00
Times Cited Count：210 Percentile：99.81(Materials Science, Multidisciplinary)no abstracts in English
; ; *; *; Yoshida, Eiichi;
PNC TN9450 92-004, 37 Pages, 1992/06
High Strength Ferritic/Martensitic Steel is one of the cardidate core materials for largescale FBR because of excellent resistance to swelling. This report are presented about the internal pressure creep of High Strength Ferritic/Martensitic Steel based on the R&D results obtained through the activities of material tests. Contents of the data sheet are as follows; (1) Material: High Strength Ferritic/Martensitic Steel Fuel cladding tube (
6.50.47.mm
) (2) Environment: In Air and In Sodium (3) Test temperature: 600 and 650C (4) Hoop stress: 9.48
32.43 kgf/㎜
(5) Number of data: 13 points
; *; Seki, Masayuki; Tobita, Noriyuki; Nagai, Shuichiro; ;
PNC TN8410 91-221, 67 Pages, 1991/08
None
Tsutagi, Koichi; Seki, Masayuki; Tobita, Noriyuki; Nagai, Shuichiro; ; *; *
PNC TN8410 91-256, 64 Pages, 1991/05
None
