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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.
Tsuji, Hirokazu; Miya, Kenzo*
Nucl. Eng. Des., 155, p.527 - 546, 1995/00
Times Cited Count：1 Percentile：16.76(Nuclear Science & Technology)no abstracts in English
Tsuji, Hirokazu; Miya, Kenzo*
Nucl. Eng. Des., 155, p.547 - 557, 1995/00
Times Cited Count：2 Percentile：28.65(Nuclear Science & Technology)no abstracts in English
Tsuji, Hirokazu; *; Nakanishi, Tsuneo*; *; Nakajima, Hajime
JAERI-M 93-209, 64 Pages, 1993/10
no abstracts in English
Tsuji, Hirokazu; Miya, Kenzo*
Preprints of the Post SMiRT Seminar No. ll on Construction Codes and Engineering, p.3.4-1 - 3.4-19, 1991/08
no abstracts in English
Tsuji, Hirokazu; Miya, Kenzo*
Preprints of the Post SMiRT Seminar No. ll on Construction Codes and Engineering, p.3.4-20 - 3.4-39, 1991/08
no abstracts in English
Tsuji, Hirokazu; ; Nakajima, Hajime
JAERI-M 90-191, 126 Pages, 1990/11
no abstracts in English
Hada, Kazuhiko; Motoki, Yasuo; Baba, Osamu
JAERI-M 90-148, 231 Pages, 1990/09
no abstracts in English
Onizawa, Takashi
no journal, ,
no abstracts in English
Onizawa, Takashi; Toyota, Kodai; Imagawa, Yuya; Okajima, Satoshi; Ando, Masanori
no journal, ,
In order to realize a fast reactor that achieves both safety and economic efficiency at a high level, Japan Atomic Energy Agency (JAEA) is developing the material strength standard for fast reactor design. JAEA has developed the material strength standard based on the acquired data and its evaluation results, and the standard have been incorporated in the Japan Society of Mechanical Engineers (JSME) code, Rules on the Design and Construction of Nuclear Power Plants, Section II, Fast Reactors (JSME D&C FRs Code). This paper describes the standard that recently incorporated in the JSME D&C FRs code and ongoing studies for improvements in the near future.
