Conference Logo ProgramMaster Logo
Conference Tools for MS&T25: Materials Science & Technology
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools

About this Abstract

Meeting MS&T25: Materials Science & Technology
Symposium Metallic Nuclear Fuel Design, Fabrication and Characterization
Presentation Title Impact of Dislocation Loops on the Thermal Transport in Nuclear Fuels: A First-Principles Atomistic Approach
Author(s) Saqeeb Adnan, Marat Khafizov
On-Site Speaker (Planned) Saqeeb Adnan
Abstract Scope Understanding how irradiation-induced defects influence thermal transport is essential for designing advanced nuclear fuels with enhanced performance and safety. We present a first-principles-based atomic-scale model that captures the impact of extended defects, specifically dislocation loops with long-range strain fields, on the thermal transport of nuclear fuel materials. The impact of these defects has traditionally been approximated through analytical formulation that grossly simplifies the impact of long-range strain fields. In this work, we incorporate a perturbative approach using real phonon dispersion calculated from density functional theory (DFT) and extend Klemens’ original formulation for extended defects. Our calculation of thermal conductivity is bench-marked against experimentally measured values from irradiated UO2 and ThO2. This work offers fundamental insights into defect–phonon interactions and establishes a predictive pathway for correlating microstructural evolution with fuel performance.

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced Characterization of Pu Oxidation with S/TEM
Characterization of High Uranium Density Compositionally Complex Refractory Alloys
Exploring the Complex Interplay Between Phases, Porosity, and Thermal Properties in Metallic Fuels
High Temperature Structures, Oxidation, and Thermodynamics of Uranium Fuels
Impact of Dislocation Loops on the Thermal Transport in Nuclear Fuels: A First-Principles Atomistic Approach
Metal Fuel Performance in Sodium Fast Reactors: Post-Irradiation Examination and Innovative Experiments
Mitigating FCCI in Metallic Fuels: Evaluating Cladding Liners Using Multiscale Modeling
Multi-scale modeling of wastage layer formation in metallic fuel cladding
Multiscale fuel performance modeling of U-Mo fuel for research reactors
Nitinol as Surrogate for Laser Additive Manufacturing of Uranium-10 Zirconium Metallic Nuclear Fuel
Perspectives on Accelerated Fuel Irradiation Testing in Uranium-Zirconium Alloys
Refractory Systems in Uranium-Containing Alloys for High Temperature Metallic Fuel
Thermal transport of uranium nitride (UN) after irradiation
Three-Dimensional Microstructural Evolution of Irradiated U-10Zr Nuclear Fuel Revealed by Synchrotron X-ray Micro-CT
Understanding grain refinement and gas bubble evolution in U-10Mo fuel using phase-field modeling
Volume fraction of porosity and distinct phases in irradiated FAST U-Zr fuel using manual point counting and automatic image analysis

Questions about ProgramMaster? Contact programming@programmaster.org | TMS Privacy Policy | Accessibility Statement