Pre-immersion effect on IASCC susceptibility in austenitic stainless steel

Book · Fri Sep 01 04:00:00 EDT 1995

OSTI ID:99569

Katsura, R; Morisawa, J; Nishimura, S ^[1]

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1. Nippon Nuclear Fuel Development Co., Ltd., Ibaraki (Japan)

Irradiation assisted stress corrosion cracking (IASCC) is now recognized as one of the major engineering concerns for extending the operating life of light water reactors (LWRs). IASCC appears in the form of intergranular stress corrosion cracking (IGSCC) and the threshold fluence level has been reported to be about 5 {times} 10 {sup 24}n/m{sup 2} (E>1MeV) in austenitic stainless steels such as types 304 and 316 SSs. In order to investigate the effect of oxide film on the initiation of irradiation assisted stress corrosion (IASCC), UCL (uniaxial constant load) tests of highly irradiated type 304 stainless steel (neutron fluence: 1.4 {times} 10{sup 26}n/m{sup 2}, E > 1MeV) with oxide film by pre-immersion were conducted in high temperature oxygenated pure water. IASCC susceptibilities were not observed even over the IASCC threshold stress level at 463--523 MPa during 2,000 hour tests. Analysis of TTF (times to failure) results in polished specimen of highly irradiated steels also supported the idea that the surface oxide film on irradiated specimens would affect the IASCC initiation mechanism, but not for IGSCC in unirradiated sensitized stainless steel. The mechanism of IASCC behavior was discussed from the viewpoint of the relationship between chromium depletion at a grain boundary and the surface oxide film.

• OSTI ID: 99569
• Report Number(s): CONF-950304--
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
CRACK PROPAGATION
EXPERIMENTAL DATA
INTERGRANULAR CORROSION
PHYSICAL RADIATION EFFECTS
REACTOR MATERIALS
SERVICE LIFE
STAINLESS STEEL-304
STRESS CORROSION
WATER COOLED REACTORS