About this Abstract |
| Meeting |
MS&T24: Materials Science & Technology
|
| Symposium
|
Ceramic Materials for Nuclear Energy Systems
|
| Presentation Title |
Ion Irradiation of UC and UN and Their Surrogates |
| Author(s) |
Rashed Almasri, Wei-Ying Chen, Adrian Wagner, Jian Gan, Lingfeng He |
| On-Site Speaker (Planned) |
Lingfeng He |
| Abstract Scope |
Uranium monocarbide (UC) and uranium mononitride (UN) are ceramic fuel materials that offer several advantages over traditional oxide fuels, such as UO2. UC and UN have a higher uranium density, higher thermal conductivity, and greater compatibility with carbon-based high-temperature matrix materials, such as SiC, graphite and ZrC, compared to oxide fuels. However, information on the irradiation performance of single-phase UC and UN is limited in the open literature. In this study we examine the impact of fission products on fuel swelling as well as radiation damage using Kr and Xe ion irradiations. In situ and ex situ electron microscopy techniques are used to visualize the size and density of gas bubbles and dislocation loops/lines in UC and UN and their surrogate materials, ZrC and ZrN. |