Similar to the fire tetrahedron, all elements must be present to meet the conditions for the event, and therefore, removing any of the three conditions will result in the prevention of SCC. Nuclear reactor components are often engineered from austenitic stainless steel, sadly leading nuclear reactor systems one step closer to meeting the conditions for SCC. .
Nuclear reactor systems are designed and manufactured to endure the arduous, full life of a nuclear reactor, and as such, the materials and the systems that they are comprised of are subjected to an extraordinary volume of neutron radiation over core life. Material degradation is inevitable with the microstructural properties beginning to change over time. The probability then of SCC increases as a function of core life. Aside from neutron irradiation material degradation, other factors can negatively affect the probability of SCC such as the electrochemical corrosion potential.
Intergranular stress corrosion cracking can occur and propagate for all grades of stainless steels if under the appropriate environmental condition. One such condition is known as the electrochemical potential (ECP) and the threshold of potential occurrence is at ECP > -230mV (IAEA, 2011, pg. 9). Factors potentially effecting the ECP include flow conditions, temperature or electrochemical impurities. Additionally, the ECP can be effected by the presence of oxygen and hydrogen peroxide, which are generated at extremely high temperatures during the oxidizing radiolysis process. The high temperature condition requirement also could significantly affect crack growth rate in the event IGSCC were to occur. Testing of pressurized water reactor alloys has shed light on the identifiable thresholds for various alloys. Alloys such as alloy 600 base metal, alloy 690 base metal and alloy 132/82 (weld metals) showed results for thresholds stresses but alloy 690 and other high chromium weld metals were inconclusive due to no cracking having occurred.
