IRRADIATION ASSISTED STRESS CORROSION CRACKING (IASCC) OF LOW STRENGTH AND HIGH STRENGTH ALLOYS IN LIGHT WATER REACTOR ENVIRONMENTS

摘要

Irradiation-assisted stress corrosion cracking (IASCC) is one of the primary degradation modes for structural materials in light water reactors. As reactors age, the accumulated damage generates new concerns for IASCC susceptibility of core materials. Seven austenitic alloys: 316L SS, 310 SS, 690, and C22 (low strength alloys), and X750, 718 and 725 (high strength alloys) were irradiated to nominal damage levels of 2.5 and 5 dpa, and two ferritic/martensitic (FM) steels: T92 and 14YWT were also irradiated to 2.5 dpa. Radiation induced microstructure including voids, loops and precipitates were characterized. The IASCC resistance was evaluated by constant extension rate tensile (CERT) tests to 4% strain in both simulated BWR normal water chemistry (288°C high purity water, 2000 ppb O_2) and PWR primary water (320°C, 1000 ppm B, 2 ppm Li, 35 cc/kg H_2). Cracking severity was evaluated by crack length per unit area, as it incorporates both the number of cracks and their length. Clear degradation in IASCC resistance was observed in these alloys as dose increased from 2.5 to 5 dpa. Despite significant microstructural changes, no single microstructural feature was identified that could account for the degradation in IASCC resistance in these alloys. Among these alloys, alloys 690 and X750 were identified as the most IASCC resistant materials for low strength and high strength application, respectively.