A state of the art on primary side stress corrosion cracking in nuclear power plant

A state of art on primary water stress corrosion cracking (PWSCC) of alloy 600 used as steam generator tubing of nuclear power plant and remedial action on the PWSCC were reviewed and analyzed. One of the major metallurgical factors which have effect on PWSCC is Cr carbide distribution. A semicontinuous intergranular Cr carbide distribution enhance PWSCC of alloy 600. PWSCC rate is reported to be reported to be proportional to exp(-50 cal/RT) {sigma}{sup 4}. PWSCC rate also increase with increase in hydrogen partial pressure from 0 to 150 ppm and then decreased with further increase in hydrogen partial pressure to 757 ppm. Development of PWSCC prediction technology which takes into account tubing material, fabrication process and operating history of steam generator is needed to manage PWSCC of domestic nuclear power plant. PWSCC has mainly occurred at expansion irregularities within tubesheet, expansion transitions, dented tube support plate intersections and transition and apex of U bend. Remedial actions to PWSCC are sleeving, plugging, temperature reduction, Ni plating, Ni sleeving, shot peening and steam generator replacement in worst case. Option to remedial actions depend on plant specific such as plant age, leak rate from primary to secondary, density and progression of PWSCC. Ni sleeving developed in Framatome seems to be a powerful method because it never subject to PWSCC. Remedial action should be developed and evaluated for possible PWSCC of domestic nuclear power plant. (author)