Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research & development study for the aging management & maintenance of the nuclear power plants. Based on the JNES project results, JNES proposed “IASCC evaluation guide for BWR core internals”. The purpose of this paper is to describe the background of the guide, especially crack growth rate (CGR) tests for irradiated stainless steels. The CGR tests had been carrying out with neutron irradiated compact tension (CT) specimens under constant load and crack growth was measured using the reversing dc potential drop (DCPD). Irradiation had been conducting in the core region of the Japan Material Testing Reactor (JMTR) in simulated BWR water environments. The specimens were irradiated to fluence ranging from 5×1024 n/m2 (0.7 dpa) to 1 × 1026n/m2 (E>1MeV)(14 dpa). CGR of SUS304L and SUS316L were formulated as a function of fluence and stress intensity factor. Outline of IASCC evaluation guide for irradiated core internals was described in this paper. The method requires time to time evaluation, that is, residual stress and CGR are calculated for the neutron fluence of the evaluating time, and then, structural integrity of core internal is evaluated by fracture mechanics analysis. The guide is tentative one and going to be modified by reflecting new knowledge and discussion.

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