A. S. Rao

Crack Growth Behavior of Irradiated Type 316 SS in Low Dissolved Oxygen Environment

Cracking susceptibility of austenitic stainless steels is known to be affected by dissolved oxygen (DO) or corrosion potential. In low-DO environments, crack growth rate (CGR) is significantly lower than that in high-DO environments. A strong dependence of CGR on corrosion potential has also been seen in irradiated stainless steels. While it has been shown that reducing the potential can reduce the CGRs of irradiated SSs, some high-dose specimens have shown elevated CGRs even in low potential environments. Thus, it is not clear how irradiation affects the dependence of CGR on corrosion potential. In the present study, a disk-shaped compact tension specimen of Type 316 SS was tested in low-DO environment. The specimen had been irradiated in the BOR-60 reactor to 5 dpa at 320°C. Post-irradiation CGR tests were performed in a low-DO environment. The effect of unloading on crack growth behavior in low-DO environment is discussed.

Crack Growth Rate and Fracture Toughness of CF3 Cast Stainless Steels at ~3 DPA

Cast austenitic stainless steels (CASS) used in reactor core internals are subject to high-temperature coolant and energetic neutron irradiation during power operations. Due to both thermal aging and irradiation embrittlement, the long-term performance of CASS materials is of concern. To assess the cracking behavior of irradiated CASS alloys, crack growth rate (CGR) and fracture toughness J-R curve tests were performed on two CF3 alloys. Miniature compact tension specimens were irradiated to ~3 dpa, and were tested at ~315 °C in simulated LWR coolant environments with low corrosion potentials. No elevated cracking susceptibility was observed at this dose in the test environments. The power exponents of the 3 dpa J-R curves were much lower than that of unirradiated or irradiated specimens at lower doses, indicating a significant decline in fracture resistance. A preliminary microstructural study revealed irradiation-induced microstructural changes in both austenite and ferrite, suggesting an embrittlement mechanism involving both phases at this dose level.

Microstructure and Deformation Behavior of Thermally Aged Cast Austenitic Stainless Steels

Cast austenitic stainless steels (CASS) consist of a dual-phase microstructure of delta ferrite and austenite. The ferrite phase is critical for the service performance of CASS alloys, but can also undergo significant microstructural changes at elevated temperatures, leading to severe embrittlement. To understand thermal aging embrittlement, fracture toughness J-R curve tests were performed on unaged and aged CF8 specimens at 315 ℃. The microstructure of CF8 was also examined before and after thermal aging with transmission electron microscopy and atom probe tomography. While no microstructural change was observed in the austenite after thermal aging, a high density of G-phase precipitates and a phase separation of alpha/alpha prime were detected in ferrite. To study the deformation behavior, tensile tests were performed at room temperature with in situ wide-angle X-ray scattering measurements. The differences in lattice strains between ferrite and austenite were much higher in the aged than in the unaged samples, suggesting a higher degree of incompatible deformation between ferrite and austenite in the aged samples.

Cyclic Crack Growth Rate of Irradiated Austenitic Stainless Steel Welds in Simulated BWR Environment

Reactor core internal components in light water reactors are subjected to neutron irradiation. It has been shown that the austenitic stainless steels used in reactor core internals are susceptible to stress corrosion cracking after extended neutron exposure. This form of material degradation is a complex phenomenon that involves concomitant conditions of irradiation, stress, and corrosion. Interacting with fatigue damage, irradiation-enhanced environmental effects could also contribute to cyclic crack growth. In this paper, the effects of neutron irradiation on cyclic cracking behavior were investigated for austenitic stainless steel welds. Post-irradiation cracking growth tests were performed on weld heat-affected zone specimens in a simulated boiling water reactor environment, and cyclic crack growth rates were obtained at two doses. Environmentally enhanced cracking was readily established in irradiated specimens. Crack growth rates of irradiated specimens were significantly higher than those of nonirradiated specimens. The impact of neutron irradiation on environmentally enhanced cyclic cracking behavior is discussed for different load ratios.Copyright