|About this Abstract
||2020 TMS Annual Meeting & Exhibition
||Characterization of Minerals, Metals and Materials
||The Effect of Radiation Damage and Radiolysis on the Corrosion of SiC with and without Corrosion-mitigation Coatings
||Peter Doyle, Takaaki Koyanagi, Caen Ang, Yutai Kato, Steven Zinkle, David Carpenter, Stephen S. Raiman
|On-Site Speaker (Planned)
Due to favorable high temperature strength and corrosion resistance, radiation damage resistance, and adequate mechanical properties, SiC is a leading candidate to replace Zr-based alloys as accident-tolerant fuel cladding material for light water reactors. However, SiC does not form a passivating oxide in liquid water due to the dissolution of SiO2. To evaluate the operational viability of SiC, monolithic and ceramic matrix composite (CMC) SiC was exposed to PWR water chemistry at 300°C in the MIT nuclear reactor water loop in the presence of neutron and gamma flux, gamma flux, and only coolant (~127 days, 0.5-1 dpa). Coatings of monolithic Cr, monolithic TiN, and multilayer Cr/CrN on SiC were also included to evaluate the effectiveness of corrosion-mitigation. SiC was found to corrode acceptably in all conditions, with elevated mass loss in the gamma and neutron flux conditions. Mixed effectiveness of the coatings in mitigating SiC corrosion was observed.