Stress corrosion cracking (SCC) is a pervasive failure mode for austenitic stainless steels core
internals components in nuclear light water reactors. Some cases of SCC may be exacerbated by
irradiation. Irradiation may change in the dislocation distribution and increase the yield stress
(hardness) of the materials. Irradiation may also alter the local chemistry of the austenitic alloys,
for example causing chromium depletion and silicon enrichment at the grain boundaries. The
objective of the present work was to perform laboratory tests in order to better understand the
role of Si on the microstructure, electrochemical behavior and susceptibility to SCC of austenitic
stainless steels. Little or no effect was found on the effect of Si on the electrochemical behavior
in high temperature water of type 304 SS containing from less than 1% Si up to 5% Si in the
bulk. Similarly, current SCC crack growth rate results are not conclusive regarding a consistent
effect of the bulk concentration of Si on the SCC resistance of the stainless steels.
Keywords: Austenitic stainless steels, silicon, electrochemical behavior, repassivation rate, stress
corrosion cracking