Microstructural and Stress Corrosion Cracking Characteristics of Austenitic Stainless Steels Containing Silicon

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.

Atomic scale investigation of radiation-induced segregation in austenitic stainless steels

“…IASCC in austenitic stainless steels is most often associated with radiation-induced segregation (RIS) at grain boundaries (GB) [1]. Thus, RIS has been extensively studied for the two last decades [3][4][5][6][7][8][9][10] for both neutron and ion irradiated alloys. Ion irradiations cannot directly simulate neutron conditions, particularly due to large differences in the damage rates and, in some cases, bulk versus near surface behaviour.…”

Towards an integrated materials characterization toolbox

“…An example of this strategy is provided by the work of Lozano-Perez et al to understand the effects of cold work in SCC of 304 stainless steel; cold working is known to increase the crack growth rate in pressurized water nuclear reactors under primary water conditions, but the mechanisms that control crack propagation are unclear. [313][314][315][316][317][318][319] Cold-worked samples are harder, have measurably different corrosion properties, and exhibit a complex microstructure with a high density of so-called twin deformation bands, which have a complex internal structure where sub-cells and dislocation bundles are frequently observed. The density of bands is dependent on the level of cold work.…”

Irradiation-assisted stress corrosion cracking

“…Crack growth data on unirradiated structural materials in BWR and PWR environments have been acquired since the 1970s, and have been extensively discussed by Andresen and many other researchers (Andresen, 1993(Andresen, , 2002(Andresen, , 2005aAndresen, Chou, Morra, Nelson, & Rebak, 2009;Andresen, Emigh, & Morra, 2005;Andresen & Morra, 2005a, 2007a, 2007b, 2008a, 2008bAndresen et al, 2004a;. The fi rst crack growth rate measurements on irradiated stainless steels in the laboratory and in situ in an operating BWR (Andresen & Ford, 1995;Andresen et al, 1990) were performed by GE in the mid-1980s, and subsequently many other IASCC growth rate studies have been performed.…”