Hm. Chung et al., Tensile and stress corrosion cracking properties of type 304 stainless steel irradiated to a very high dose, NUCL ENG DE, 208(3), 2001, pp. 221-234
Certain safety-related core internal structural components of light water r
eactors, usually fabricated from Type 304 or 316 austenitic stainless steel
s (SSs), accumulate very high levels of irradiation damage (20-100 displace
ment per atom or dpa) by the end of life. Our databases and mechanistic und
erstanding of the degradation of such highly irradiated components, however
, are not well established. A key question is the nature of irradiation-ass
isted intergranular cracking at very high doses, i.e. is it purely mechanic
al failure or is it stress-corrosion cracking? In this work, hot-cell tests
and microstructural characterization were performed on Type 304 SS from th
e hexagonal fuel can of the decommissioned EBR-II reactor after irradiation
to approximate to 50 dpa at approximate to 370 degreesC. Slow-strain-rate
tensile tests were conducted at 289 degreesC in air and in water at several
levels of electrochemical potential (ECP), and microstructural characteris
tics were analyzed by scanning and transmission electron microscopies. The
material deformed significantly by twinning and exhibited surprisingly high
ductility in air, but was susceptible to severe intergranular stress corro
sion cracking (IGSCC) at high ECP. Low levels of dissolved O and ECP were e
ffective in suppressing the susceptibility of the heavily irradiated materi
al to IGSCC, indicating that the stress corrosion process associated with i
rradiation-induced grain-boundary Cr depletion, rather than purely mechanic
al separation of grain boundaries, plays the dominant role. However, althou
gh IGSCC was suppressed, the material was susceptible to dislocation channe
ling at a low ECP, and this susceptibility led to a poor work-hardening cap
ability and low ductility. (C) 2001 Elsevier Science B.V. All rights reserv
ed.