Stress corrosion cracking aspects of nuclear steam generator tubing materials in the water containing lead at high temperature

Slow strain rate tests (SSRT) were performed to investigate the effect of lead concentration on stress corrosion cracking (SCC) of a nuclear steam generator tubing material (Alloy 600 MA) in mildly alkaline solution at 340 degrees C. Lead concentration as low as 100ppm caused extensive SCC of the specimens. C-ring tests were carried out at controlled potentials to evaluate stress corrosion cracking tendency and morphology of nuclear steam generator tubing materials (Alloy 600 MA, 600 TT and 690 TT) in a solution of 1M NaOH with Pb at 340 degrees C. The number of cracks for Alloy 690 TT was fewer than the value for Alloy 600 TT, however, a much longer maximum crack length was observed for Alloy 690 TT than that for Alloy 600 TT, while Alloy 600MA showed the highest number of cracks and the longest maximun crack length, in the solution containing 5,000ppm Pb at a controlled potential (95mV vs. Ni). In addition, different cracking morphologies were found for these materials, depending on the Pb concentration. Alloy 600 TT and 690 TT show transgranular SCC morphology in the solution containing 5,000 ppm Pb but intergranular or no SCC with 100ppm Pb.