| About this Abstract |
| Meeting |
2025 TMS Annual Meeting & Exhibition
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| Symposium
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Microstructural, Mechanical, and Chemical Behavior of Solid Nuclear Fuel and Fuel-Cladding Interface II
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| Presentation Title |
Fission Product Distribution Analysis in Zr Lined U-Mo Fuels Using Transmission Electron Microscopy and Atom Probe Tomography |
| Author(s) |
Nicole Rodriguez Perez, Sobhan Patnaik, Shehab Shousha, Mukesh Bachhav, Luca Capriotti, Geoffrey Beausoleil, Benjamin Beeler, Maria A Okuniewski |
| On-Site Speaker (Planned) |
Nicole Rodriguez Perez |
| Abstract Scope |
Uranium-molybdenum (U-Mo) alloys are fuel candidates for power producing reactors, such as light water reactors. In reactor, cladding temperatures may exceed 500°C, temperatures at which U-Mo fuels exhibit significant fuel-cladding chemical interaction (FCCI). Zirconium (Zr) barriers have proven effective to prevent the FCCI in U-Mo fuels at temperatures below 150ºC and thus are proposed as an option to mitigate FCCI at higher temperatures. To demonstrate the efficacy of Zr liners with U-Mo fuel, a Zr lined U-10wt.%Mo specimen was irradiated to 4.29% FIMA at a peak inner cladding temperature of 440ºC. In this study, transmission electron microscopy is applied along with atom probe tomography to obtain localized burn-up relationships with fission product distributions within the fuel. Emphasis is placed on the fuel-liner interface with a focus on the lanthanide fission product penetration. Experimental results are compared to recent computational predictions based on diffusion coefficients derived from density functional theory calculations. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Characterization, Other |