| Abstract Scope |
This study investigated the behavior of niobium-stabilized austenitic stainless steel as an alternative material for nuclear fuel rod cladding under simulated LOCA (Loss of Coolant Accident) conditions, focusing on its oxidation resistance, ductility, and mechanical integrity at high temperatures (up to 1200 °C). Using ring compression test (RCT) and microstructural characterization techniques (i.e. SEM, EDS, and XRD), the research examined metal-water oxidation reaction after quenching, and the kinects of the oxide layer formation to address its limitations as a clad material. The niobium-stabilized austenitic stainless steel demonstrated superior post-exposure ductility and potential for PWR reactor applications, contributing to accident tolerant fuels and reliable energy technologies. The outcomes include quantifying oxidation rates, oxide layer thickness, post-quench mechanical strength, and direct performance comparisons with zirconium based alloys. This empirical research showed that the material had a superior perfomance during LOCA scenarios when compared with the traditional zirconium based alloys. |