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Meeting MS&T25: Materials Science & Technology
Symposium Metallic Nuclear Fuel Design, Fabrication and Characterization
Presentation Title Volume fraction of porosity and distinct phases in irradiated FAST U-Zr fuel using manual point counting and automatic image analysis
Author(s) Lily Cait Alberts, Nicole Rodriguez Perez, Sobhan Patnaik, Geoffrey Beausoleil, Maria A. Okuniewski
On-Site Speaker (Planned) Lily Cait Alberts
Abstract Scope Swelling in uranium-zirconium (U-Zr) fuels results from the accommodation of fission products (FP) within the fuel during irradiation. The formation of pores and FP precipitates affect the mechanical and thermal properties of the fuel. Quantifying porosity and phase volume fractions is essential for predictive modeling of the evolution of the fuel properties during irradiation. However, conventional post-irradiation examination (PIE) techniques are both time consuming and cost intensive. The Fission Accelerated Steady-State Testing (FAST), using reduced-geometry specimens, offers a more efficient alternative. In this study, scaled U-Zr fuel pins were irradiated to 3.9at. % burnup, at 430-530ºC. The specimens were analyzed using scanning electron microscopy (SEM), and then manual point counting and automatic image analysis were applied to obtain porosity and phase fractions. The results were compared to existing literature on conventionally irradiated U-Zr fuels to assess the capability of the FAST experiments to emulate conventional fuel microstructures.

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

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