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About this Symposium
Meeting 2018 TMS Annual Meeting & Exhibition
Symposium Computational Materials Science and Engineering for Nuclear Energy
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
TMS: Chemistry and Physics of Materials Committee
Organizer(s) Haixuan Xu, University of Tennessee
Michael R Tonks, Pennsylvania State University
Blas Pedro Uberuaga, Los Alamos National Laboratory
James R. Morris, Oak Ridge National Laboratory
Scope This symposium will highlight current computational materials science and engineering efforts for nuclear reactors in the United States and abroad. High neutron flux, thermal and chemical gradients, and corrosive environments cause significant degradation in the chemical and mechanical properties of materials. Enhanced radiation resistance of structural materials and nuclear fuels are needed to overcome technological challenges necessary for future nuclear systems. This symposium seeks abstracts that apply atomistic and mesoscale simulations to discover, understand, and engineer the macroscale performance of fission/fusion reactor materials, including fuel, cladding, and structural materials.
This symposium will also consider multiscale modeling efforts that bridge length and time scales in order to better connect simulation results with experimental data for predictive model validation. It will also highlight validation of all relevant models, as well as uncertainty quantification. Finally, the application of ICME approaches to use modeling and simulation to better understand structure-property relationships, their associated links with performance, and their application to designing future reactor concepts and materials is also desired.

Some examples include:
• Modeling and simulation of materials behavior under extreme environments – radiation, corrosion, stress and temperature, including radiation effects, phase stability, fuel-clad interactions, fission product behavior.
• Modeling and simulation of model materials to uncover fundamental behavior that affects material performance in radiative environments.
• Developing improved material models for LWR fuel and cladding.
• Modeling and simulation of new fuel materials including metal, silicide, and nitride fuels.
• Modeling and simulation of new cladding materials, such as silicon carbide, coated zirconium alloys, or FeCrAl.
• Development and integration of computational tools, methods, and databases for reactor structural material design.
Uncertainty quantification and validation of all the applications listed above.
Abstracts Due 07/16/2017
Proceedings Plan Planned: Supplemental Proceedings volume
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A Computational Study of the Effect of Irradiation and Temperature on Fatigue Crack Growth in Zirconium
A Model Coupling Hydrides Formation and Mechanical Behavior of Zircaloy Cladding during Fuel Rod Lifecycle
A Phase Field Study of Void Superlattice Formation in Irradiated Materials
Ab Initio Modeling of Vacancy-type Defects in a High Entropy Alloy
Atomistic Modeling of Primary Damage in Fe-based Ferritic Alloys
Atomistic Study of Thermal Spike Response of Xe Bubbles in UO2
Breaking the Power Law: Multiscale Simulations of Self-ion Irradiated Tungsten
Calculating free energies of metal-He interfaces from atomic models
Cluster Dynamics in Irradiated Materials: A Hybrid Deterministic/Stochastic Coupling Algorithm
Competition of Deformation Modes in Irradiated Zr Alloys: A Micromechanical Approach
Computational Modelling of Thermal Transport in Uranium Dioxide
Computer Simulations of Dislocation-obstacle Interactions in the Hardening and Recovery of BWR-irradiated 304L SS
Density Functional Theory Simulations of Clusters in Reactor Pressure Vessel Steels
Density Functional Theory Study of the Magnetic Moment of Solute Mn in BCC Fe
Discrete Dislocation Sinks in Spatially Resolved Cluster Dynamics Simulations
Dislocation Loop Bias in BCC Fe
Effect of Dopants on Uranium-based Metallic Fuels to Mitigate Fuel-cladding Chemical Interactions
Effect of Post Fabrication Voids on Irradiation Performance of U-10Mo Monolithic Mini-plate
Effect of Stress on Hydrides Precipitation and Re-orientation in Zircaloy: A Phase Field Study
Effects of Oxygen on the Density of States and Elastic Properties of Hafnium—First Principles Calculations
Evolution of Grain Boundary Structure and Composition in Irradiated SiC
Experimentally Validated Computational Modeling of Advanced Alloys and Radiation Effects for Nuclear Energy Applications
Flux Effect on RIS in a Fe3%Ni Model Alloy: CD Modelling of the T Shift
Formation and Re-orientation of Multi-phase Zirconium Hydrides under Applied Strain
Fundamental Understanding of Corrosion of Nuclear Materials: Holistic Approach to Fuel Cladding Corrosion under Irradiation
Fundamentals of Energy Dissipation and Defect Energetics of Maximally Disordered Alloys
Grain Growth and Grain Subdivision in Triuranium Disilicide, a Potential Light Water Reactor Fuel
Hydrogen Transport and Trapping in Irradiation Damaged Zirconium Alloys
Interstitial-mediated Diffusion and Aggregation Mechanism for Transmutation Elements Rhenium and Ormium Precipitation in Tungsten
Kinetics of Point Defect Absorption by Sinks: Effect of Point Defect Properties and Surrounding Microstructure
Microstructure-sensitive Phase Field Fracture Model Including Anisotropic Elastic Properties
Modeling Inclusions with Surface Stresses in the Phase Field Framework
Molecular Dynamics Simulations of Effects of Stacking Fault Energies on Defect Formation Process in FCC Metals
Molecular Dynamics Study of Defect-grain Boundary Interactions in Irradiated PyC-like Configuration
Molecular Dynamics Study of Irradiation Damage in Nano-grain Sized Polycrystal
Morphological Study of Dispersion Phases in Heterogenous Waste Form Materials for Efficient Nuclear Waste Containment
Multiscale Simulations of Sequential Dislocation/Obstacle Interactions in FCC Metals
Off-stoichiometric Cluster Dynamics in Irradiated Oxides
Phase-field Modeling of Fission Rate Effect on the Gas Bubble Swelling in U-Mo Fuel
Phase Field Modeling of Grain Boundary Evolution in Porous Oxides: Grain Growth and Pore Mobility Effects
Phase Transformation in Zirconium Oxide – a Mesoscale Study
Quantitative Phase Field Modeling of Void Growth in Irradiated Solids
Radiation Damage in Carbon-based Materials
Rate-theory Modeling of Irradiation Damage Cascades and the Influence of the Underlying Microstructure using the MOOSE Framework
Rate Theory Modeling of Fission Gas Behavior in Ion Implantation Experiment
Real-space Diffusion-driven Models for Microstructural Evolution of Irradiated Materials
Residual Point Defects and their Evolution near Dislocation Loops and Grain Boundaries in α-zirconium: An Atomistic Study
Simulation of Phosphorous Migration to Grain-boundary by Molecular Dynamics
Sink Density Effect on Radiation-induced Segregation and Precipitation in Fe-Cr Alloys
The Thermodynamic and Kinetic Properties of Spinels as They Relate to CRUD
Thermomechanical Analysis of the Multi-metallic Layered Composite Fuel Cladding for Improved Accident Tolerance of LWRs
Thermophysical Properties of (U,Zr)O2 Pellet-cladding Interface though MD Simulations
Using Computational Modeling to Interpret Experimental Measurements of Irradiation Induced Hardening in Metals
Vacancy Clusters and Xenon Diffusion in UO2
Xe Bubble Behaviors in Single Crystal Molybdenum via Molecular Dynamics Simulation


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