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About this Symposium
Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Mechanical Behavior of Nuclear Reactor Components
Sponsorship TMS Materials Processing and Manufacturing Division
TMS Structural Materials Division
TMS: Nanomechanical Materials Behavior Committee
TMS: Nuclear Materials Committee
Organizer(s) Clarissa A. Yablinsky, Los Alamos National Laboratory
Assel Aitkaliyeva, University of Florida
Eda Aydogan, Middle East Technical University
Laurent Capolungo, Los Alamos National Laboratory
Khalid Hattar, University of Tennessee Knoxville
Kayla Yano, Pacific Northwest National Laboratory
Caleb Massey, Oak Ridge National Laboratory
Scope Current and future generation nuclear reactors require improved structural materials that improve efficiency during in-service conditions, allow for long reactor lifetimes, and increase safety during accidents. Given the increasingly large number of reactor design being considered (e.g. fusion, molten salt, LWRs, etc.), a series of distinct material concepts have been proposed to address these needs. Effects of reactor environments on mechanical behavior will be a key component to predicting strength and performance of materials in the aforementioned circumstances.

This symposium aims to take a closer look at the mechanical behavior of reactor components across length scales. With recent advancements and increased use of in-situ techniques, more is known about irradiation effects on strength than ever before. Simultaneously, ex-situ techniques are critical to probe component-sized parts, and validate the use of a material for inclusion within a reactor. Furthermore, synergy with materials modeling is advancing the prediction of material performance under normal and accident conditions, as well as reactor lifetimes.

Topics of interest include, but are not limited to:
• Mechanical behavior testing, including tension, compression, bend, bulge, creep, fatigue, and fracture
• Effects of environment on strength, including dose, dose rate, temperature, and corrosion
• Hardness testing, including nanohardness and microhardness
• Development of microstructure sensitive material strength models
• Modeling and simulation of irradiation defect interactions during mechanical testing
• Macroscopic component modeling for full scale performance
• In-situ mechanical testing, including micromechanical and nanomechanical compression and tension
• Novel techniques to probe material strength under reactor conditions

Abstracts Due 07/20/2020
Proceedings Plan Planned:
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A Model for Dislocation Climb and Precipitate Interactions Applied to Creep in Ferritic Steel
A Novel Displacement Cascade Driven Irradiation Creep Mechanism in Pure Copper
Atom-probe Study of Nano-hardening Features in Neutron Irradiated RAFM Steels
Atomistic Simulations and Theoretical Modelling of the Yield Behavior of Industrial Tantalum Alloys
Bridging the Length Scales via Femtosecond Laser Machining of Micro-mesoscale Tensile Specimens
Burst Behavior of Accident Tolerant Fuel Cladding Concepts under Simulated Loss-of-coolant Conditions
C-ring Compression of SiC-SiC Cladding at 1200C with In-situ X-ray Computed Micro-tomography
Challenges to Accurate Evaluation of Bulk Hardness with Nanoindentation Testing at Low Indent Depths
Controlling Helium Morphology in Pure Metals: Dislocation-helium Interactions
Correlating the Neutron-irradiation Induced Hardening and Solute Nano-clustering in Oxide Dispersion Strengthened Alloys
Creep Behavior of Helium Implanted Submicron Films under Irradiation
Creep Crack Growth Behaviour of Austenitic Stainless Steels Alloy 709 and 316H
Development of Modified 3Cr-3WVTa Base Bainitic Steels for Fusion Structural Applications
Dose and Temperature Dependence of Microstructure and Mechanical Properties in Ion-Irradiated PM-HIP Inconel 625
Effect of Cr Concentration On <111> and <100> Dislocation Loop Formation in Fe-Cr Alloys
Effects of Low-temperature Neutron Irradiation and Post-weld Heat Treatment on Tensile Properties of Welded Zircaloy-4
Enabling In-situ Crack Growth Testing and Monitoring in VTR Cartridge Loop Environments
He Ion Irradiation Response of a Gradient T91 Steel
High Temperature Strength of Additively Manufactured Gr91 Steel
High Throughput Assessment of Creep Behavior of Advanced Alloys for Model Development and Validation
In-situ Micro-tensile Studies on the Effects of Ion Irradiation on the Mechanical Properties of Small-grained Alloys
In-situ Observations of the Failure Mechanisms of Hydrided Zircaloy-4 under Different Stress-States
In-situ Scanning Electron Microscopic Observation of Creep and Creep-fatigue of Alloy 709
Irradiation Resistance in Several Multi-principal Element Alloys
Low Temperature Neutron Irradiation and Mechanical Properties of Welded AISI 347
Mechanical Behavior and Radiation Effect in Additively Manufactured 316L Stainless Steel
Mechanical Characterization of Neutron Irradiated HT-9 Heats (ORNL, LANL and EBR II) at LWR and Fast Reactor Relevant Temperatures
Mechanical Properties of Additively Manufactured 316L Stainless Steel before and after Neutron Irradiation
Microstructural Effects on the Mechanical Behavior of FeCrAl Alloys
Modeling the Effect of Helium Bubbles, Rigid Inclusions, and Grain Boundaries on Crack Initiation in Nickel
Multiscale Modeling of Creep and Transient Conditions in Steels: Application to HT9 Steel Alloy
Nanomechanical Assessment of a Neutron Irradiated U-10Zr Fuel
Neutron Irradiation Response of SA508 Pressure Vessel Steel Prepared by Powder Metallurgy and Hot Isostatic Pressing
Novel Small Scale Mechanical Testing Techniques for Nuclear Materials
On the Role of Material Pedigree to Predict Engineering Material Properties
Probing the Mechanical Behavior of Irradiated Materials through Micromechanical Testing
Quantifying Zirconium Embrittlement Due to Hydride Microstructure Using Image Analysis
Simulating the Effects of Neutron Irradiation on Zirconium Alloys: A Crystal Plasticity Finite Element Approach
Simulation of Intergranular Void Growth under the Combined Effects of Surface Diffusion, Grain Boundary Diffusion, and Bulk Creep
Small Scale Mechanical Testing of Nuclear Fuel and Cladding
Stress Corrosion Cracking Resistance of FeCrAl Alloys in Light Water Reactor Environments
The Merit of In-situ Environmental TEM for the Study of Tungsten under Fusion-relevant Conditions
The Thermo-mechanical Fracture of Chromium-zirconium Systems
Void Swelling and Transmutation in Tungsten Metals and Alloys after Fusion Relevant Neutron Irradiation
Wear Behavior of Incoloy™ 800HT and Inconel™ 617 for High-Temperature Gas-cooled Reactor (HTGR) Applications
α’ Precipitation and Hardness Change in Ion Irradiated High Purity FeCr Alloys


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