STRESS CORROSION CRACKING OF STEAM GENERATOR TUBE AND PRIMARY PIPE IN PWR TYPE NUCLEAR POWER PLANTS

The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), con-stant load test (CLT) and low frequency cyclic loading test(LFCLT). The purpose of these tests is to getthe test data for evaluating the integrity of pressurized boundary of pipe in Qinshan and Guangdongnuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes inwelded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy--800, Inconel--600,Inconel--690 and 321 SS which are used for steam generator in PWR. The effects of material metallur-gy, shot--peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior ofSCC are investigated.