Abstract
Austenitic stainless steels (SSs) core internal components in nuclear light water reactors (LWRs) are susceptible to irradiation-assisted stress corrosion cracking (IASCC). One of the effects of irradiation is the hardening of the SS and a change in the dislocation distribution in the alloy. Irradiation may also alter the local chemistry of the austenitic alloys; for example, silicon may segregate and chromium may deplete at the grain boundaries. The segregation or depletion phenomena at near-grain boundaries may enhance the susceptibility of these alloys to environmentally assisted cracking (EAC). The objective of the present work was to perform laboratory tests in order to better understand the role of Si in the microstructure, properties, electrochemical behavior, and susceptibility to EAC of austenitic SSs. Type 304 SS can dissolve up to 2 pct Si in the bulk while maintaining a single austenite microstructure. Stainless steels containing 12 pct Cr can dissolve up to 5 pct bulk Si while maintaining an austenite structure. The crack growth rate (CGR) results are not conclusive about the effect of the bulk concentration of Si on the EAC behavior of SSs.
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