Characteristics of lead induced stress corrosion cracking of alloy 690 in high temperature

Book · Tue Oct 01 04:00:00 EDT 1996

OSTI ID:367702

Chung, K K ^[1]; Lim, J K ^[2]; Watanabe, Yutaka; Shoji, Tetsuo ^[3]

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1. Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)
2. Chonbuk National Univ., Chonju (Korea, Republic of)
3. Tohoku Univ., Sendai (Japan). Research Inst. for Fracture Technology

Slow strain rate tests (SSRT) were conducted on alloy 690 in various lead chloride solutions and metal lead added to 100 ppm chloride solution at 288 C. The corrosion potential (rest potential) for the alloy was measured with SSRT tests. The cracking was observed by metallographic examination and electron probe micro analyzer. Also, the corrosion behavior of the alloy was evaluated by anodic polarized measurement at 30 C. Resulting from the tests, cracking was characterized by cracking behavior, crack length and crack growth rate, and lead effects on cracking. The cracking was mainly intergranular in mode, approximately from 60 um to 450 um in crack length, and approximately 10{sup {minus}6} to 10{sup {minus}7} mmS-1 in crack velocity. The cracking was evaluated through the variation the corrosion potential in potential-time and lead behavior during SSRTs. The lead effect in corrosion was evaluated through active to passive transition behavior in anodic polarized curves. The corrosion reactions in the cracking region were confirmed by electron probe microanalysis. Alloy 690 is used for steam generation tubes in pressurized water reactors.

• OSTI ID: 367702
• Report Number(s): CONF-960389--
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
CORROSIVE EFFECTS
CRACK PROPAGATION
EXPERIMENTAL DATA
INCONEL 690
INTERGRANULAR CORROSION
LEAD CHLORIDES
PWR TYPE REACTORS
REACTOR MATERIALS
STRESS CORROSION