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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Metallic Fuels - Design, Fabrication, and Characterization
Presentation Title Influence of Heterogeneous Microstructures and Large Deformation on Blister Formation in Monolithic U–Mo Nuclear Fuels
Author(s) Shenyang Hu, Benjamin Beeler
On-Site Speaker (Planned) Shenyang Hu
Abstract Scope Blister formation in monolithic U–Mo nuclear fuels is governed by complex, interrelated phenomena, including radiation-induced defect accumulation, phase-specific swelling behaviors, evolving mechanical properties of Zr, UZr₂, and U–Mo phases, and heterogeneous microstructures. These factors collectively influence local stress states, enhance defect transport, and drive gas bubble evolution. In this work, we develop a mesoscale finite-deformation model focused on a representative volume near the Zr/U–Mo interface. The model incorporates key irradiation-induced phenomena: large volumetric swelling, swelling-dependent mechanical softening in U–Mo, and stress-free strains in Zr, UZr₂, and U–Mo phases. To investigate gas-driven blistering, the model introduces structural defects and accounts for evolving thermo-mechanical properties and defect mobility. We systematically examine the coupling between stress fields, defect transport, and gas bubble pressure under irradiation. The results elucidate critical conditions for blister nucleation and growth, demonstrating the model’s capability to predict fuel performance risks related to structural defects and irradiation effects.
Proceedings Inclusion? Planned:
Keywords Modeling and Simulation, Nuclear Materials, Other

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advancements in Metallic Fuel Development and Qualification: Materials Challenges
Advances in the Understanding of Fuel-Cladding Chemical Interaction from FFTF MFF HT9 Clad U-Zr Metal Fuels
An Innovative Technique for Rapid Elemental and Isotopic Characterization of Nuclear Fuels
Analysis of Higher Burnup Sodium Free Annular U-10Zr Fuel
Assessing Thermal Conductivity in U-Zr Fuels: A Comparison of Modeling and Experimental Results
Atomic Mobility Assessments in the U-C-X Ternary Systems
Capturing fission products in metallic fuels using nanostructured interfaces
Comparison of FAST and Historical Irradiation Testing of U-Zr Fuels
Component Assembly and Fuel Fabrication for the IMPACT Advanced Test Reactor Irradiation Experiment
Compositional Effects on Phonon Dispersion and Lifetimes of UxTh1-xO2 using Inelastic X-Ray Scattering
Database-Driven Validation of BISON Metallic Fuel Performance Models for Sodium-Cooled Fast Reactors
Effects of ZrN coating and heat treatment on U-Mo dispersion fuel systems under irradiation
Fabrication and Properties of Metal Fuels with Controlled Porosities by Spark Plasma Sintering
Faulted and Perfect Loop Evolution with Dose and Temperature in Proton Irradiated Uranium Mononitride (UN)
First-Principles Assessment of Thermal Conductivity, in UCO TRISO Fuel with Varying Carbon Stoichiometry
First-Principles Evaluation of Heat Capacity in Disordered Uranium Oxycarbide Phases for TRISO Fuel Application
High-Resolution Characterization of U-Zr Metallic Fuel for Elucidating Precipitation and Fission Products Distribution Using Atom Probe Tomography
Influence of Heterogeneous Microstructures and Large Deformation on Blister Formation in Monolithic U–Mo Nuclear Fuels
Interactions between U-10Mo fuel with a burnable matrix
Investigating Phonon Transport in Uranium Nitride Using Neutron scattering
Investigating the structural and chemical properties in the oxidation of U-Th MOx fuel
Metal Fuels Opportunities Beyond Sodium Fast Reactors
Microalloying Metallic Fuels for Tracking and Traceability
Microstructural Investigation into Two Variants of Fuel-Cladding Chemical Interaction Observed in an Irradiated HT-9 Clad U-10Zr Metallic Fuel
Microstructure Analysis of Uranium-Molybdenum Fuel Alloy from Accelerated Irradiation at Various Temperatures
Novel Phase Identification and Characterization: Experimental insights in The U-Tc Binary System for Metallic Fuel Modeling
Scaling Irradiation Behaviors: Examining Steady-State Swelling, Redistribution, and Fission Gas Release in Larger Diameter U-Zr Fuel Pins
Site-specific porosity and high temperature behavior of U-10Zr fuel
Smaller and faster: conventional vs. nanocalorimetry techniques for determining thermophysical properties of nuclear fuels
Thermal Expansion and Neutron Cross-section of U-Mo fuel to 1000C
Thermal transport of uranium nitride (UN) after irradiation
Thermodynamic and Kinetic Pathways of Impurity-Induced Degradation in U-Ti-Nb-Mo-C Alloys
U-Zr Alloy Properties Review and Applicability to Lightbridge Corporation Fuel Performance Activities

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