Surface Characterization of Accident Tolerant Fuel Claddings Candidates After Prolonged Exposure to Steam Oxidation

Accident Tolerant Fuel Claddings are designed to improve fuel performance and safety in abnormal reactor conditions. The development of these cladding materials has become a top priority for the nuclear community, with incidents such as the Three Mile Island accident and the Fukashima Dai-chi accident. The development of these claddings has resulted in 2 subsections: coating current claddings or replacing cladding materials with new materials. The claddings studied in this thesis are a Zircaloy-4 cladding coated with chromium through physical vapor deposition, Zircaloy-4, and a FeCrAl alloy, C26M. This thesis assessed the surface chemistry, roughness, and wettability after extended exposure to a high-temperature steam environment. The surfaces of the materials were studied to determine the impact of the flowing steam on the material’s oxidation. The results showed that both proposed cladding materials exhibited improved oxidation characteristics compared to the Zircaloy-4. The surface chemistry of the claddings evidenced the oxide layer with a thickness of 3 μm, 23 μm, and 150 nm for chromium-coated Zircaloy-4, substrate Zircaloy-4, and C26M, respectively. The results of the coated cladding were found to agree with previous research in where the chromia layer forms and protects the substrate from oxidation. The FeCrAl alloy, C26M, showed the oxide development of an external layer rich in iron, with a lower fraction of chromium and aluminum, and an internal layer rich in aluminum. The preliminary characterization of the materials’ mechanical properties results also showed a decrease in hardness for all materials