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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Microstructure and Chemical States of Fission Products in Irradiated AGR-1 and AGR-2 TRISO Particle UCO Fuel Kernels
Author(s) Yong Yang, Isabella van Rooyen, Zhenyu Fu, Boopathy Kombaiah
On-Site Speaker (Planned) Yong Yang
Abstract Scope Over the past three years, extensive microstructural characterizations were conducted on irradiated AGR-1 and AGR-2 TIRSO fuel UCO kernels. It was found that the irradiated fuel kernel mainly consists of UO2 and UC, while the UO2 is the dominating fuel phase. UC phase contains a significant amount of fission products including Zr, Nb, Mo, Ru, Tc and Rh. The post irradiation safety testing significantly promotes the precipitations of U2Ru(Rh)C2 and UMo(Tc)C2 phases. Fission gas bubbles were only observed in UC or UMoC2 phases. Results from EDS analysis and APT study show that the lanthanides, e.g., Nd, are more likely to enrich within the UO2 phase. Comparison between the AGR-1 and AGR-2 fuel particle kernels shows the differences in irradiated microstructures and chemical states of fission products may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Characterization,


A Model of Grain Boundary Energy Anisotropy in Uranium Dioxide Nuclear Fuel
A Thermo-mechanical Coupled Phase Field Dynamic Fracture Model and Its Application in UO2
Comprehensive Treatment of Thermal Transport Under Irradiation in ThO2
Development of Hydrothermal Corrosion Barrier Coatings for High-density Nuclear Fuels
Development of Yttrium Hydride for High Temperature Moderator Application
Effect of UV and Gamma Irradiation on the Hydrothermal Corrosion of Ion-irradiated SiC
Electron Microscopy Characterization of the Fuel-cladding Interaction in Annular Fast Reactor MOX
Evolution of Ion Irradiated Nitride Ceramics Properties for Coated Particle Fuel Systems
Exotic Magneto-elastic Properties in Uranium Dioxide
Hydrothermal Corrosion of Silicon Carbide
Hydrothermal Corrosion Study of Additive Manufactured SiC Fibers
Impact of Dislocation Loops on Thermal Conductivity of CeO2
In-situ Measurement of Tritium Release from Lithium Aluminate Under Neutron Irradiation
Influence of Dose Rate and Temperature on Mass Transport in Hematite
Ionization Effects on Damage Accumulation Behavior in SiC
Irradiation Damage in High-entropy Carbide Ceramics
Irradiation Effects on Zirconium Alloy Oxides and Their Impacts on In-reactor Corrosion Rates
Microstructural Analysis and Micro-mechanical Testing on Xenon-irradiated Uranium Dioxide
Microstructural and Fission Products Analysis from Irradiated UO2 Fuel Using Atom Probe Tomography
Microstructural Characterization of Radiation Effects in 3D printed SiC
Microstructure and Chemical States of Fission Products in Irradiated AGR-1 and AGR-2 TRISO Particle UCO Fuel Kernels
Modeling of Pressure-driven Inter-granular Fracture in High Burnup Structure UO2 during LOCA Using A Phase-field Approach
Molecular Dynamics Investigations of AlN-based Piezoelectric Ceramics under Irradiation
Multiscale Modeling of High Burn-up Structure (HBS) Formation and Evolution in UO2
On the Role of Neutron Irradiation Damages on Fission Products Transport in the SiC Layer of TRISO Fuel Particles
Oxidation Behavior of TRISO Fuel Materials
Phase-field Modeling of Bubble Growth During High Burn-up Structure Formation in UO2
Radiation Tolerance of Nanoporous Gadolinium Titanate
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces
TEM Characterization of Dislocation Loops in Ion-irradiated Single Crystal ThO2
TMIST-3A Post-irradiation Examination
Towards a Model of Coupled Irradiation and Corrosion

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