Miao Song

The intergranular attack near both stagnant and active stress corrosion crack (SCC) tips of solution annealed alloy 690 were characterized after constant extension rate tensile tests in simulated pressurized water reactor (PWR) primary water containing 18 cc H2/kg H2O. In both cracks, an intergranular oxide, composed of NiO and Cr2O3, was observed beyond the crack tip with an adjacent Cr depleted grain boundary migration zone. The stagnant crack has a compact Cr2O3 covering the crack tip and the adjacent grain boundary migration zone is deep and free of oxidation, while the active crack has porous penetrative oxide extending into the migration zone. The ability to form a compact Cr2O3 film at the crack tip is important to the IGSCC resistance and may be dictated by the efficiency of Cr diffusion up the migrated grain boundary. When a compact Cr2O3 film does not form at the crack tip, the Cr depleted grain boundary migration zone is subject to penetrative oxidation via the inward diffusion of oxygen, making the zone susceptible to crack propagation. Thus, even a high Cr content cannot guarantee the immunity of a Ni base alloy to SCC during dynamic straining.

Thirteen austenitic stainless steels, nickel-base alloys, and ferritic alloys were irradiated using 2 MeV protons at 360 °C to a damage level of 2.5 displacements per atom (dpa). Comprehensive microstructural characterization was performed for irradiation-induced features, including dislocation loops, voids, precipitates, and radiation induced segregation (RIS). Dislocation loops formed in all alloys except 14YWT, while voids were observed in alloys 316 L, 310, C22, and 14YWT. Irradiation-induced formation of γ′ precipitates was observed in alloys 316 L, 310, 800, and 690; the irradiation-enhanced, long-range ordered Ni2Cr phase (Pt2Mo-type) was observed in alloys 690, C22, 625, 625Plus, 625DA, and 725; and G-phase was observed in alloy T92. No irradiation-induced precipitates were observed in alloys X750, 718 or 14YWT. Precipitation of the γ′ phase can be understood through segregation and clustering of Si, Al, and Ti. Overall, austenitic stainless steels are generally susceptible to irradiation damage in the form of loops, voids, precipitates, and RIS. Ni-base alloys have this same type of dislocation loops and RIS behaviors but are more resistant to void swelling. Ferritic alloys showed better resistance to loop formation, void swelling and irradiation-induced precipitation. From the degree of irradiation-induced microstructural change, alloy T92 was identified as the most radiation resistant among these alloys.