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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Modeling of Pressure-driven Inter-granular Fracture in High Burnup Structure UO2 during LOCA Using A Phase-field Approach
Author(s) Wen Jiang, Larry Aagesen , Kyle Gamble
On-Site Speaker (Planned) Wen Jiang
Abstract Scope UO2 has been widely used as nuclear fuel material for commercial light water reactors (LWRs). One of the current objectives of LWR operators is to extend current UO2 fuels to high burnups and ensure them to operate safely during accident conditions such as Loss of Coolant Accidents (LOCA). During a LOCA, it has been observed in experiments that the UO2 fuel can finely fragment and axially relocate within the rod resulting in the possibility of fuel dispersal upon cladding rupture. To understand the fragmentation mechanism, a phase field fracture model is developed to model the pressure-driven inter-granular fracture. The bubble pressure inside the bubble due to temperature transient is obtained by a phase field bubble pressure model. The simulation results will be used to assess whether the bubble pressure is sufficient to drive fracture and provide fundamental understanding of the fragmentation mechanism for high burnup structure.
Proceedings Inclusion? Planned:
Keywords Computational Materials Science & Engineering, Nuclear Materials, Ceramics

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

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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