|About this Abstract
||2022 TMS Annual Meeting & Exhibition
||Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
||In Situ Study of High Temperature Mechanical Behavior of Irradiated FeCrAl Alloys
||Tianyi Sun, Tongjun Niu, Dongyue Xie, Adam Gabriel, Lin Shao, Jian Wang, Haiyan Wang, Xinghang Zhang
|On-Site Speaker (Planned)
FeCrAl alloy is one of the near-term cladding materials in accident tolerant fuel systems. The improved high-temperature steam oxidation resistance of FeCrAl alloy compared with Zr alloy leads to a longer responding time and less hydrogen generation under accident conditions. With growing interest in FeCrAl alloy for nuclear applications, the mechanical behavior of FeCrAl alloy at high temperatures and irradiation states is of great importance. In this study, we used in situ SEM micropillar compression technique to investigate the mechanical behavior of both annealed and irradiated FeCrAl alloy at various temperatures, ranging from 200 – 500 °C. The temperature-dependent deformation behavior is quantified and discussed.
||Nuclear Materials, Mechanical Properties,