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Meeting Materials Science & Technology 2020
Symposium Ceramics in the Nuclear Fuel Cycle
Presentation Title Evaluation of the Corrosion of High Purity CVD SiC in Light Water Reactor Environments
Author(s) Peter Doyle, Stephen Raiman, Steven Zinkle
On-Site Speaker (Planned) Peter Doyle
Abstract Scope 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 288C and 350C 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 4m/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.
Proceedings Inclusion? Planned: At-meeting proceedings

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A First-principles Database Approach to Predicting Trans-Uranic Waste Forms
Evaluation of the Corrosion of High Purity CVD SiC in Light Water Reactor Environments
First-principles Study on the Trapping and Recombination of Tritium in Lithium Vacancy of the γ-LiAlO2 (100) Surface
Multi-scale Cs Sorbents Easily Transformable into Waste Confinement Matrices
Nb and Ti Alloying Effects on the Phase and Oxidation of U3Si2
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces
Thermophysical Properties of Sintered Yttrium Dihydride

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