Microstructural evolution and mechanical behavior of austenitic stainless steels are evaluated for neutron and heavy-ion irradiated materials. Radiation hardening in austenitic stainless steels is shown to result from the evolution of small interstitial dislocation loops during light-water-reactor (LWR) irradiation. The presence of these loops modifies deformation characteristics and prompts a significant increase in yield strength. Available data on stainless steels irradiated under LWR conditions have been analyzed and microstructural characteristics assessed for the critical fluence range (0.5 to 5x1021 n/cm2) where irradiation-assisted stress corrosion cracking susceptibility is observed. Heavy-ion irradiations are used to produce similar defect microstructure enabling the investigation of hardening and deformation mechanisms. Scanning electron, atomic force and transmission electron microcopies are employed to examine tensile test strain rate and temperature effects on deformation characteristics. Dislocation loop microstructure promote inhomogeneous planar deformation within the matrix and regularly spaced steps at the surface during plastic deformation. Twinning is the dominant deformation mechanism at rapid strain rates and at low temperatures, while dislocation channeling is favored at slower strain rates and at higher temperatures. Both mechanisms produce highly localized deformation and large surface slip steps. Channeling, in particular, is capable of creating extensive dislocation pileups and high stresses at internal grain boundaries which may promote intergranular cracking.
Irradiation-assisted stress corrosion cracking (IASCC) remains an important concern in austenitic stainless steel core components of nuclear power reactors. Although most observations of IASCC have been reported in oxidizing boiling-water reactor (BWR) environments, it is now quite clear that cracking will also occur in nonoxidizing environments. Recent work has documented IASCC in neutron-irradiated stainless steels under hydrogen-water chemistryl-3 and pressurized water reactor (PWR)M conditions. The critical neutron fluence for the onset of intergranular cracking appears to greater for the nonoxidizing environments. However, the exact differences are a function of the assessment method (e.g., comparing service observations, in-situ swelling mandrel tests or ex-situ slow-strain-rate tests) as illustrated in Figure 1.
It is likely that the cracking mechanism and some controlling parameters are altered when comparing response in oxidizing versus nonoxidizing environments, i.e., at high versus low electrochemical potentials. For example, radiation-induced chromium depletion which is believed to play an important role in oxidizing environments,7,8 should not control cracking in nonoxidizing environments. Another radiation-induced material change that will influence SCC susceptibility is the defect microstructure evolution and increase in strength that occurs over this same critical fluence range. A general tendency has been established in many alloy systems for enhanced cracking as the material yield strength increases. This is particularly true for hydrogen-induced cracking. The inhomogeneous deformation that often occurs in hardened microstructure can promote dislocation pileups and high local stresses leading to enhanced environmental cracking. However, our current understanding of radiation hardening effects on deformation and SCC of LWR-irradiated stainless steels is limited. Very little characterization of pertinent microstructure have been reported and no direct investigations of deformation behavior and SCC resis
