Austenitic stainless steels (SS) commonly used in the primary circuit of light water reactors (LWRs) have excellent corrosion resistance in demineralized (DM) water at high temperature (290°C–330°C) and pressure (7.4 MPa for boiling water reactor [BWR], 16 MPa for pressurized water reactor [PWR]). Radiolysis of primary DM water in BWR forms 200–300 ppb of oxidizing species (normal water chemistry [NWC]), which elevates the electrochemical potential of the SS from −300 to −600 mVSHE (mV with standard hydrogen electrode) when dissolved oxygen is 10–20 ppb to +100 to +200 mVSHE. Neutron irradiation of SS further induces metallurgical and microstructural changes, which compromises corrosion resistance. Thus, radiation makes BWR (NWC) environment hostile, causing extensive irradiation-assisted stress corrosion cracking (same as intergranular stress corrosion cracking [IGSCC]) in austenitic SS. IGSCC occurs in both sensitized (grain boundary chromium depletion) and nonsensitized conditions (strain-hardened region in base metal immediately adjacent to the weld fusion zone) and is a generic problem. IGSCC in BWR can be mitigated by hydrogen addition in water (hydrogen water chemistry [HWC]). BWR-HWC has limitations and is ineffective where boiling occurs. IGSCC of SS in PWR is limited because hydrogen addition in primary water suppresses radiolysis though the formation of aggressive environment in low-flow occluded regions can cause IGSCC in SS. Increasing demand for economic power has led to the primary environment in nuclear reactors to become hostile. Conceptual supercritical water reactor (SCWR) will use DM water at 500°C–600°C and 25 MPa, which is extreme for conventional LWR materials. IGSCC in austenitic SS and nickel-based alloys occurs in both oxidizing and reducing SCWR conditions, which can be further exacerbated by radiation. This article reviews how benign primary DM water becomes hostile in LWRs and extreme in SCWR conditions, causing SCC to be a generic problem in austenitic alloys.