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Journal Articles

Strength anisotropy of rolled 11Cr-ODS steel

Tanno, Takashi; Yano, Yasuhide; Oka, Hiroshi; Otsuka, Satoshi; Uwaba, Tomoyuki; Kaito, Takeji

Nuclear Materials and Energy (Internet), 9, p.353 - 359, 2016/12

BB2015-1727.pdf:6.74MB

 Times Cited Count:4 Percentile:47.23(Nuclear Science & Technology)

Materials for core components of fusion reactors and fast reactors, such as blankets and fuel cladding tubes, must be excellent in high temperature strength and irradiation resistance because they will be exposed to high heat flux and heavy neutron irradiation. Oxide dispersion strengthened (ODS) steels have been developing as the candidate material. Japan Atomic Energy Agency (JAEA) have been developing 9 and 11 Chromium (Cr) ODS steels for advanced fast reactor cladding tubes. The JAEA 11Cr-ODS steels were rolled in order to evaluate their anisotropy. Tensile tests and creep tests of them were carried out at 700 $$^{circ}$$C in longitudinal and transverse orientation. The anisotropy of tensile strength was negligible, though that of creep strength was distinct. The observation results and chemical composition analysis suggested that the cause of the anisotropy in creep strength was prior powder boundary including Ti-rich precipitates.

JAEA Reports

The evaluation of material base standard of ODS ferritic stainless steel core component for fast breeder reactors

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$$_{2}$$O$$_{3}$$). 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

JAEA Reports

Creep strength of Hastelloy XR welded joints

Tachibana, Katsumi; Nishi, Hiroshi; Eto, Motokuni;

JAERI-Tech 99-024, 65 Pages, 1999/03

JAERI-Tech-99-024.pdf:3.01MB

no abstracts in English

JAEA Reports

Creep characteristics of Alloy 800H

Tachibana, Katsumi; Nishi, Hiroshi; Eto, Motokuni;

JAERI-Tech 98-010, 107 Pages, 1998/03

JAERI-Tech-98-010.pdf:3.52MB

no abstracts in English

Journal Articles

Attractive characteristics of high-chromium iron-based alloys for nuclear reactor application

Hishinuma, Akimichi; *; *; *

Physica Status Solidi, 160(2), p.431 - 440, 1997/00

no abstracts in English

Journal Articles

Structural design of pressure vessel of High Temperature Engineering Test Reactor

Kurihara, Ryoichi; Tachibana, Yukio; Nishihara, Tetsuo; Maruyama, So; Shiozawa, Shusaku; *

Atsuryoku Gijutsu, 32(3), p.154 - 165, 1994/00

no abstracts in English

JAEA Reports

None

Tsutagi, Koichi; ; Tobita, Noriyuki; Nagai, Shuichiro; ; *; *

PNC-TN8410 91-256, 64 Pages, 1991/05

PNC-TN8410-91-256.pdf:4.7MB

None

Journal Articles

High temperature strength of hastelloy XR under biaxial stress states

; Hada, Kazuhiko; Koikegami, Hajime*; *

Nippon Genshiryoku Gakkai-Shi, 33(5), p.475 - 481, 1991/05

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Filler metal development for Hastelloy alloy XR

Watanabe, Katsutoshi; Nakajima, Hajime; Sahira, Kensho*; *; Takeiri, Toshiki*; Saito, Teiichiro*; Takatsu, Tamao*; Nakanishi, Tsuneo*

JAERI-M 89-206, 120 Pages, 1989/12

JAERI-M-89-206.pdf:6.81MB

no abstracts in English

Journal Articles

Effect of Cr phase on high temperature creep resistance of a 30Cr-50Ni-2Mo alloy

*; *; *; *; *; *

Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku 29(2), p.265 - 273, 1988/00

no abstracts in English

Journal Articles

Mechanical and metallurgical characteristics of iron-base austenitic alloys designed for LMFBR cladding

Kondo, Tatsuo; Nakajima, Hajime; *; ; *; *

Proc.of Int.Conf.on Irradiation Behaviour of Metallic Materials for Fast Reactor Core Components, 6 Pages, 1979/00

no abstracts in English

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