Corrosion test qualification for in situ stress relief of recirculating steam generators' U-bends

Highly stressed alloy 600 is susceptible to intergranular stress corrosion cracking (IGSCC) in high-purity water at nuclear steam generator (NSG) operating temperatures. Two regions in recirculating steam generators (RSG) are particularly prone to primary-side-initiated SCC: tube expansion transitions of the tube in the tubesheet and tight radii tube bends. One remedial measure to improve IGSCC in these regions is to heat the tubes and thus relieve the residual stresses that contribute significantly to the cracking problem. This article describes a corrosion test program using the accelerated SCC environments of sodium tetrathionate and sodium hydroxide to qualify an in situ stress-relief process for the most SCC-susceptible U-bends in an RSG