Materials data acquisition and life assessment for BWR reactor pressure vessel

Conference · Sun Sep 01 04:00:00 EDT 1996

OSTI ID:277810

Kato, Takahiko ^[1]; Kanazawa, Yasushi ^[2]; Suzuki, Isao; Koyama, Masakuni ^[3]; Akiyama, Mamoru ^[4]; Okubo, Tadatsune ^[5]; Mishima, Yoshitsugu ^[6]

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1. Hitachi Ltd., Ibaraki (Japan)
2. Toshiba Corp., Yokohama (Japan). Nuclear Energy Div.
3. Japan Power Engineering and Inspection Corp., Urayasu, Chiba (Japan)
4. Tokyo Univ., Hongo, Tokyo (Japan). System Engineering Faculty
5. Sophia Univ., Tokyo (Japan)
6. Tokyo Univ., Shinbashi, Tokyo (Japan)

The systematic testing of materials to obtain the data for life assessment on reactor pressure vessel (RPV) of boiling water reactor (BWR) is ongoing. The data of fatigue crack propagation in simulated BWR reactor water and fracture toughness of low alloy steels and nickel base alloys are summarized in this paper. As a case study, fatigue crack propagation and crack stability at the feed water nozzle corner, made of a low alloy steel were evaluated for RPV life assessment. Postulating an 80-year normal operation of an 800 MWe BWR plant, it was predicted that fatigue crack growth would not penetrate through the nozzle corner wall and the nozzle material has enough toughness to prevent unstable fracture.

• OSTI ID: 277810
• Report Number(s): CONF-960306--; ISBN 0-7918-1226-X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
BWR TYPE REACTORS
FRACTURE PROPERTIES
LOW ALLOY STEELS
MECHANICAL PROPERTIES
NICKEL BASE ALLOYS
PRESSURE VESSELS
REACTOR MATERIALS
SERVICE LIFE
STRESS INTENSITY FACTORS