|About this Abstract
||2022 TMS Annual Meeting & Exhibition
||Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
||Advanced Characterization and Multiscale Testing for SiC Ceramic Matrix Composite Cladding as Accident Tolerant Fuel Candidate Materials for LWR Applications
||Peng Xu, David Frazer, Tsvetoslav Pavlov, Nikolaus Cordes, Fabiola Cappia, David Kamerman, Sean Gonderman, Christian Deck, Jack Gazza
|On-Site Speaker (Planned)
SiC cladding is pursued by the nuclear industry as a candidate material for accident tolerant fuels for the light water reactors (LWRs). It offers superior oxidation resistance and strength retention in high temperature steam during severe accidents. Although the irradiation behavior of SiC has been well studied in the past, the irradiation data is still lacking for SiC ceramic matrix composite (CMC) as fuel cladding material at prototypic LWR conditions. The Idaho National Laboratory is partnering with General Atomics to perform test reactor irradiation at the Advanced Test Reactor (ATR) at the prototypical Pressurized Water Reactor conditions to evaluate its in-pile performance. The SiC composite cladding materials will be examined before and after irradiation. Multiscale characterization and testing will be conducted to correlate radiation induced microstructure changes to its thermophysical and mechanical properties. Cladding integrity and corrosion will be evaluated non-destructively and destructively.
||Ceramics, Composites, Nuclear Materials