Electrochemical study on the PbSCC of SG tube material at a high temperature

In the development of steam generator tube materials, Alloy 600 is expected to endure the high temperature corrosive environment of nuclear power plants. However, the occurrence of a stress corrosion cracking (SCC) of steam generator tubes is not unusual as nuclear power plants have been operated for a long time. Since Copson and Dean reported that the SCC susceptibility of steam generator tubes could be affected by lead, many other researchers have investigated a SCC of steam generator tube materials in a lead-containing test solution, and they have revealed that lead is one of the detrimental contaminants for the damage of tube materials. Some of the plants in the world have reported that the steam generator was contaminated by lead with a high concentration. This study was performed to observe a lead induced SCC (PbSCC) in relation to the electrochemical property on the surface of an Alloy 600 MA tubing material. The Steam generator tube materials which were used in nuclear power plants were named (heat-treatment) Alloy 600 MA (meal annealed), 600 TT (thermally treated), 600 HTMA (high temperature meal annealed), Alloy 800 and Alloy 690 TT (thermally treated)