About this Abstract |
Meeting |
MS&T22: Materials Science & Technology
|
Symposium
|
Ceramics for a New Generation of Nuclear Energy Systems and Applications
|
Presentation Title |
SiC Oxidation and Irradiation Resistance in Advanced Nuclear Reactor TRISO Fuel |
Author(s) |
Kathy Lu, Yi Je Cho |
On-Site Speaker (Planned) |
Kathy Lu |
Abstract Scope |
Under accidental conditions for high temperature gas-cooled reactors (HTGR), the SiC layer in tri-structural-isotropic (TRISO) fuel particles can be exposed to water vapor. In this study, oxidation behaviors of surrogate TRISO fuel particles were investigated in a He-20 vol% water vapor mixed atmosphere at temperatures up to 1600 °C. Volatilization of the oxide layer was analyzed using a mechanistic model. The prediction indicates that the oxidized SiC layer should retain fission products. In addition, microstructure and defect evolution in the oxidized SiC layer of surrogate TRISO fuel particles under ion irradiation were observed by in-situ transmission electron microscopy. The defect number density at 800 °C was an order of magnitude lower than that in the sample irradiated at room temperature. Also, crystalline SiO2 had higher radiation resistance compared to SiC. A defect reaction rate theory was utilized to understand the fundamental defect evolution process and irradiation resistance difference. |