Effects of Pb on SCC of Alloy 600 and Alloy 690 in Prototypical Steam Generator Chemistries

Abstract:

Intergranular attack/stress corrosion cracking of Alloy 600 continues to be an issue in the tube/tube support plate crevices and top of tubesheet locations of recirculating steam generators and in the upper bundle of free span superheated regions of once through steam generators (OTSG). Recent examinations of degraded pulled tubes from several plants suggest possible lead involvement in the degradation. Laboratory investigations have been performed to determine the factors influencing lead cracking in Alloy 600 and Alloy 690 steam generator tubes. The test environment is believed to be prototypical, with the addition of lead oxide, of a concentrated liquid phase existing in the pores of thin deposits on upper bundle tubes of an OTSG. Highly strained reverse U-bend specimens were tested at controlled electrochemical potentials. Maximum susceptibility was at open circuit potential, unlike cracking of Alloy 600 in caustic and acid sulfate environments where maximum susceptibility occurs when specimens are polarized above the open circuit potential. Transgranular, intergranular and mixed mode cracking was observed and in all Alloy 600 conditions tested (mill annealed, sensitized, thermally treated) while thermally treated Alloy 690 has so far resisted cracking. A film rupture/anodic dissolution model with displacement plating of Pb preceding passive film formation is consistent with the experimental observations

• Periodical: Materials Science Forum (Volumes 475-479)
• Pages: 1387-1392
• DOI: https://doi.org/10.4028/www.scientific.net/MSF.475-479.1387
• Online since: January 2005
• Authors: Jesse Lumsden, Allan McIlree, Richard Eaker, Rocky Thompson, Steve Slosnerick
• Keywords: Alloy 600, Alloy 690, Corrosion Potential, Intergranular Corrosion (IGC), Lead Cracking, Once Through Steam Generators, Steam Generator (SG), Stress Corrosion Cracking (SCC)

References

[1] R. W. Staehle: Assessment of and Proposal for a Mechanistic Interpretation of the SCC of High Nickel Alloys in Lead-Containing Environments, 11th Environ. Degrad. ANS, 2003, p.381.

[2] D. P. Rochester and R. W. Eaker: Laboratory Examination Results from Oconee Nuclear Station Once through Steam Generator Tubes, 9th Environ. Degrad., TMS, 2000, p.639.

DOI: 10.1002/9781118787618.ch66