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Meeting MS&T25: Materials Science & Technology
Symposium Metallic Nuclear Fuel Design, Fabrication and Characterization
Presentation Title Advanced Characterization of Pu Oxidation with S/TEM
Author(s) Douglas Smith, Matthew Janish, Sarah Hernandez
On-Site Speaker (Planned) Douglas Smith
Abstract Scope Despite its use in mixed oxide fuels (MOX) – since the 1960s – and recent interests in metallic fuels, relatively little research has been carried out on the oxidation of plutonium using modern electron microscopy techniques. To understand differences in the oxidation of pure and alloyed plutonium, the metal/oxide interfaces of a range of samples have been characterized with a Scios 2 dual-beam focused ion beam (FIB) microscope and a probe-corrected Themis G3 80-300 scanning transmission electron microscope (S/TEM). Observations of the samples include phase identification and compositional mapping along the metal/oxide interfaces in addition to calculation of the inelastic mean free path of 300 keV electrons in plutonium and a discussion of the methods employed to mitigate hydriding during TEM liftout preparation. These results are discussed in terms of the bulk sample preparation techniques and storage conditions affecting plutonium microstructures and oxidation.

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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