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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Presentation Title Impact of Grain Boundary and Surface Diffusion on Predicted Fission Gas Bubble Microstructural Evolution Behavior and Release in UO2 Nuclear Fuel
Author(s) Md Ali Muntaha, Sourav Chatterjee, Michael Tonks, Larry Aagesen, David Andersson, Brian Wirth, Sophie Blondel
On-Site Speaker (Planned) Md Ali Muntaha
Abstract Scope In this work, we quantify the impact of grain boundary (GB) and surface diffusion on fission gas bubble morphological evolution and release in UO2 nuclear fuel using 2D and 3D simulations with a hybrid phase field/cluster dynamics model. We represent fast GB and surface diffusion using a heterogeneous diffusivity that is a function of the order parameters representing bubbles and grains. A free surface boundary condition is applied to predict the gas release. We find that the GB diffusivity directly impacts the rate of gas release via GB transport and that the GB diffusivity is likely below 103 times the lower value from Olander and van Uffelen (2001). We also find that surface diffusivity impacts bubble coalescence and mobility, and we can observe the coalescence behavior more prominently using 3D simulation. We conclude that the bubble surface diffusivity is likely below 10-4 times the value from Zhou and Olander (1984).
Proceedings Inclusion? Planned:
Keywords Ceramics, Modeling and Simulation, Nuclear Materials

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From First-principles, Modeling the Effect of Point-defect Phonon Scattering on the Thermal Conductivity of Oxide Fuels
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In Situ High-temperature 3D Imaging of the Damage Evolution in a SiC Nuclear Fuel Cladding Material
Mechanical and Thermophysical Properties of ZrC, NbC, and TaC Binary Carbide Surrogate Fuels for Nuclear Thermal Propulsion Systems
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Multiphysics and Multiscale Modeling of the Mechanical Properties of the Porous Pyrocarbon Buffer Layer in TRISO Particle Fuel
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Radially Resolved Thermo-physical Modelling in High Burnup Oxide Nuclear Fuel
Radiation Effects and Corrosion of Silicon Carbide in Nuclear Reactor Environments
Radiation Effects in Ceramics for Immobilization of Actinide-containing Nuclear Waste
Reduced Processing Temperature of Advanced Ceramic Composites
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Uranium Enrichment Homogeneity Study on HALEU/LEU UO2 Fuel Pellets
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