|About this Abstract
||2022 TMS Annual Meeting & Exhibition
||Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
||A Comprehensive Study of Responses of SiC Materials to Neutron Irradiation for ATF Cladding Application
||Takaaki Koyanagi, Christian Petrie, Jose' Arregui Mena, Hsin Wang, Yutai Katoh
|On-Site Speaker (Planned)
Development of SiC-based composites continues for use in accident-tolerant fuel cladding for light water reactors (LWRs) because of its inherent advantages of low neutron absorption, retention of strength following irradiation, and high-temperature capability. Although responses of SiC materials to neutron irradiation have been extensively studied, there remains a knowledge gap in predicting the performance of SiC cladding under irradiation. This paper presents recent findings from evaluations of SiC materials irradiated under LWR-relevant temperature and dose conditions. The research subjects include (1) high-dose irradiation effects, (2) irradiation-induced stress and dimensional changes under temperature and neutron flux gradients, and (3) stored energy release during post-irradiation annealing. These studies reach a common conclusion that mechanistic understanding and modeling of irradiation-induced lattice swelling is important to assessing the performance of SiC-based cladding. This study was supported by US Department of Energy–Nuclear Energy and Westinghouse Electric Company/General Atomics FOA projects.
||Composites, Hydrometallurgy, Nuclear Materials