Effect of temperature and cold work on the crack growth rate of alloy 600 in primary water

Journal Article · Fri Sep 01 04:00:00 EDT 1995 · Corrosion

DOI:https://doi.org/10.5006/1.3293632· OSTI ID:116517

Rebak, R B; Xia, Z; Szklarska-Smialowska, Z ^[1]

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1. Ohio State Univ., Columbus, OH (United States). Fontana Corrosion Center

Alloy 600 (UNS N06600) has been used to fabricate the tubing in steam generators (SG) of pressurized water reactors (PWR) in nuclear power plants. The susceptibility of alloy 600 to stress corrosion cracking (SCC) in hydrogenated high-temperature water was studied as a function of the degree of cold work (CW) of the alloy and the temperature of the solution. Cracks did not nucleate in smooth specimens during constant extension rate testing (CERT). However, cracking susceptibility increased as the degree of CW by bending increased. An extrapolation was made to predict cracking susceptibility in smooth non-CW conditions. The activation energy (Q) for the crack growth rate (CGR) was calculated for two CW conditions in the temperature ranges between 315 C and 350 C and 270 C to 350 C. Results showed there was no unique value of Q. Analysis of Q values suggested that the CGR depended primarily on events occurring inside the alloy, which were enhanced by processes occurring in the metal-electrolyte interface.

• Sponsoring Organization: USDOE
• OSTI ID: 116517
• Journal Information: Corrosion, Journal Name: Corrosion Journal Issue: 9 Vol. 51; ISSN 0010-9312; ISSN CORRAK
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
COLD WORKING
CORROSION RESISTANT ALLOYS
CRACK PROPAGATION
INCONEL 600
PWR TYPE REACTORS
STEAM GENERATORS
STRESS CORROSION
TEMPERATURE DEPENDENCE