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Meeting 2020 TMS Annual Meeting & Exhibition
Symposium Computational Materials Science and Engineering of Materials in Nuclear Reactors
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
Organizer(s) Dilpuneet S. Aidhy, Clemson University
Michael R. Tonks, University of Florida
Mahmood Mamivand
Giovanni Bonny, Belgian Nuclear Research Center
Scope This symposium will focus on the current computational materials science and engineering efforts towards understanding the materials behaviors and microstructure evolutions in nuclear reactors. High neutron flux, thermal and chemical gradients, and corrosive environments cause significant degradation in the microstructural and mechanical properties of materials. Enhanced radiation resistance of structural materials, nuclear fuels, cladding materials and armor materials for plasma facing components are needed to overcome technological challenges necessary for future nuclear systems such as Generation IV fission and fusion reactors. This symposium seeks abstracts that apply electronic, atomistic, mesoscale and multiscale simulations to discover, understand, and engineer the macroscale performance of fission/fusion reactor 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. New computational methods including machine learning approaches, to predict materials behaviors and develop new interatomic potentials for molecular dynamics simulations are also included. 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.
• Developing improved material models for LWR fuel and cladding.
• Modeling and Simulation of TRISO fuel
• Modelling and simulation of radiation damage and their interaction with plasma components in plasma facing materials
• Modeling and simulation of new fuel materials including metal, silicide, and nitride fuels.
• New methods to develop interatomic potentials for molecular dynamics simulations.
• Modeling and simulation of new cladding materials, such as advanced carbides, 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 07/15/2019
Proceedings Plan Planned: Supplemental Proceedings volume

A First-principles Investigation on the Co-segregation Energetics of Chromium-helium at Grain Boundaries in α-Fe
A Machine Learning Approach to Thermal Conductivity Modelling of Irradiated Nuclear Fuels
A Micromechanics-based Modeling Approach to Predict the Mechanical Properties of Zircaloy with Hydride Precipitates
A Physical Model of Zircaloy Corrosion in Water for Simulating Nuclear Reactor Clad Response
Ab-initio Molecular Dynamics Simulations of bcc U and U-Zr Alloys
Amorphous Zirconia: a Host for Excess Oxygen in Cladding Barrier Oxides?
Analyzing U-Zr Experimental Data Using Quantitative Phase-field Simulation and Sensitivity Analysis
Application of Variational Bayesian Monte Carlo Method for Improved Prediction of Doped UO2 Fuel Performance
Atomistic Studies of Nuclear Materials with Temperature: Uranium Nitride and Thermocouples
Atypical Melting Behaviour of (Th,U)O2, (Th,Pu)O2 and (Pu,U)O2 Mixed Oxides
Corrosion of Silicon Carbide in Nuclear Environments
Density Functional Theory Study of He/H Effect in W-Ni-Fe Composite for Plasma Facing Material
Developing Capabilities to Investigate the Effect of Curvature on the Radiation Response of Solid-state Interfaces
Development and Testing of Machine Learning Interatomic Potentials for Radiation Damage Calculations
DFT Calculations for Modeling Point Defect and Fission Gas Behavior in Nuclear Fuels
DFT+U Point Defect Calculations of Uranium Mononitride
Diffusion and Interaction of Prismatic Dislocation Loops in Stochastic Dislocation Dynamics
E-33 (Invited): Development of a New Thermochemistry Solver for Multiphysics Simulations of Nuclear Materials
E-34: Ab-initio Modelling of Iodine Defects in Strained Zirconium and Ordered Zirconium-oxygen Suboxides
E-36: ICME Modeling of U-10%wt Mo Alloys: A Linkage between Microstructure Evolution and Process Modeling
E-37: Machine Learning-assisted Risk-informed Sensitivity Analysis for ATF Under SBO
E-38: Mesoscale Modeling of High Burn-up Structure (HBS) Formation and Evolution in U-Mo Alloys
E-39: Molecular Dynamics and Phase-field Study of Anisotropic Grain Growth Behavior in UO2
E-40: Origin of Hardening in Spinodally-decomposed Fe-Cr Binary Alloys
E-41: Recrystallization and Grain Growth Simulations for Multiple-pass Rolling and Annealing of U-10Mo
E-44: The Contribution of Li Vacancies to the Evolution of Thermal Conductivity in Irradiate LiAlO2
E-45: Thermodynamics of Hydrogen Pickup in Zr Alloys
Electron-phonon Coupling Effects in Ion Irradiation of Metallic Systems
Exploration of Fundamental Radiation Effects Phenomena in Materials
First-principles Cluster Expansion Study of Fe and Mo Effects on Atomic Ordering in Ni-Cr Alloys
First Principle Studies of Effects of Solute Segregation on Grain Boundary Strength in Ni-based X-750 Alloy
First Principles Modeling of Ion Ranges in Self-irradiated Tungsten
First Principles Modelling of the Role of Electrons in Collision Cascades in Solids
Influence of Coordination Numbers on Representing Molten Salts for Nuclear Reactor Applications Using the Modified Quasi-Chemical Model (MQM)
Mesoscale Modeling and Experiments for Predicting the Thermal Conductivity of UZr Fuels
Microstructure-based Finite Element Model to Investigate the Effect of Grain Size and Homogenization on Hot-rolled U-10Mo
Modeling of Interface Evolution during Zirconium Alloy Corrosion
Modeling the Fracture of Zirconium at an Atomic Level and Analyzing the Effects of Temperature and Strain Rate on the Deformation Mechanisms
Molecular Dynamics Simulations of Mixed Materials in Tungsten
Molecular Dynamics Simulations of Phosphorus Migration in a Grain Boundary of α-iron
Molecular Dynamics Studies of Thermal Conductivity Degradation of UO2 due to Dispersed Xe Atoms and Xe Bubbles
Phase-field Simulation of Intergranular Fission Gas Bubble Growth in Uranium Silicide
Plasticity of Zirconium Hydrides: an Edge and Screw Planar Discrete Dislocation Model
Recent Development of Thermochimica for Simulations of Nuclear Materials
Reduced Order Modeling of Thermal Creep in 316H Stainless Steel
Shape and Stability of Voids and Fission Gas Bubbles in UO2
Stabilizing Gamma Hydrides in Zr through Mechanical Stress
The Effect of Minor Additives on Radiation Induced Segregation in Austenitic Steel Alloys
The Use of Molecular Dynamics Simulations for Modeling Gas - Point Defect Interaction Behavior in Nuclear Materials
Thermochemical and Phase Equilibria (CALPHAD) Modeling of Nuclear Fuel Materials: A Constant in Reactor Development
Thermodynamic Properties at the Rim in High Burnup UO2 Fuels
Zirconium Alloy Cladding Burst Mechanisms under LOCA with Burnup Extension

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