CRACK GROWTH RATE TESTING OF FAST REACTOR IRRADIATED TYPE 304L AND 316 SS IN BWR AND PWR ENVIRONMENTS

摘要

Crack growth rate testing of various stainless steels irradiated in a fast reactor was carried out with the objectives to quantify the susceptibility to irradiation assisted stress corrosion cracking (IASCC), and to improve the understanding of IASCC. One heat of Type 304Lfrom a core shroud, irradiated to 5 and 10 dpa, was tested at different stress intensity (K) levels in BWR normal water chemistry (NWC) as well as hydrogen water chemistry (HWC). Testing in PWR primary water was performed on two specimens made from cold worked (CW) Type 316 and CW Type 316+Ti baffle bolt material, at various K levels and at temperatures ranging from 288 to 340 "C. The dose of the specimens tested in PWR environments was 25 dpa. The paper will discuss the effects ofK, corrosion potential, dose and temperature on the crack growth rate.