Abstract
Plant life management is one of the most important issues to improve safety of LWR in Japan. It is known that austenitic stainless steels, which are the construction materials of reactor internals, has susceptibility to irradiation assisted stress corrosion cracking (IASCC) due to high neutron irradiation, and some incidents of IASCC in core shroud of Japanese BWR plants have been reported. Recently, "IASCC evaluation guide for BWR core internals" was proposed by Japan Nuclear Energy Safety Organization (JNES). In this study, fracture mechanics investigation for IASCC crack growth was performed for applicability assessment of the guide. It is found that it is impossible to satisfy acceptable standard of the Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code for 60 years. It is considered that the re-evaluation of IASCC crack growth rate curves by obtaining crack growth data of irradiation material in the region where the stress intensity factor is above 30 MPa√<m> and below 10 MPa√<m> and fracture toughness of irradiation material are needed.
