|About this Abstract
||Materials Science & Technology 2020
||Ceramics in the Nuclear Fuel Cycle
||Evaluation of the Corrosion of High Purity CVD SiC in Light Water Reactor Environments
||Peter Doyle, Stephen Raiman, Steven Zinkle
|On-Site Speaker (Planned)
SiCf/SiC composites have been identified as a potential accident tolerant fuel cladding. Among the R&D topics requiring investigation, aqueous corrosion behavior under normal operating conditions requires improved understanding. In the present work, SiC was exposed to high purity pressurized water in a constantly recirculating autoclave environment. Exposures ranged between 288°C and 350°C for times up to 2000 h, with either 1-4ppm O2 or 0.15-3ppm H2 dissolved in the water. Oxygen reacted with SiC with a reaction order of 1 and was initially linear with time until grain fallout became prevalent. No localized attack was observed in the absence of oxygen and uniform dissolution is predicted to be below 4µm/5 years, an acceptable rate. A predictive equation is given and compared to other published data. Recommendations are made for future testing parameters, include sample preparation. Funding was provided by the U.S. Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign.
||Planned: At-meeting proceedings