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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Mechanical Behavior of Nuclear Reactor Materials and Components III
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
TMS: Mechanical Behavior of Materials Committee
Organizer(s) Assel Aitkaliyeva, University of Florida
Clarissa A. Yablinsky, Los Alamos National Laboratory
Osman Anderoglu, University of New Mexico
Eda Aydogan, Middle East Technical University
Kayla Yano, Pacific Northwest National Laboratory
Caleb Massey, Oak Ridge National Laboratory
Djamel Kaoumi, North Carolina State University
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
• 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 07/17/2022
Proceedings Plan Planned:

A Mesoscale Model of Creep in Monolithic UMo Fuels
A New Microcrack Healing Mechanism in an Annealed 14YWT Nanostructured Ferritic Alloy
Additively Manufactured Digital Image Correlation for Nuclear Materials
Anisotropic Compressive Strength of Single Crystal Zirconium Pillars and the Effects of Irradiation Hardening and Temperature Through Micro-Pillar Mechanical Testing
Castable Nanostructured Alloy Steels and the Graded Transition to Tungsten for Fusion Reactors
Cladding Coating Integrity Quantified by Ring Pull and Local Strain Analysis
Computer Vision-assisted Oxide Thickness Determination of 304 Stainless Steel in PWR Environments
Data-driven Surrogate Constitutive Modeling of Mechanical Creep Behavior under Extreme Conditions
Deformation Characteristics of Additively Manufactured 316L Stainless Steels after Neutron Irradiation
Deformation Mechanisms in Gen-IV Candidate Structural Steels Studied by In-situ Micromechanical Testing Techniques
Deformation Twinning Versus Slip in Ni-based Alloys, Containing Pt2Mo-structured, Ni2Cr-typed Precipitates
Effect of Irradiation on the Tensile Strength of Select Layers and Layer Interfaces of TRISO-coated Nuclear Fuel Particles
Effects of Helium Implantation on Mechanical Properties Near the Tungsten-carbide Interfaces of Dispersion Strengthened Tungsten Alloy
Elastic Field Solution of Helical Dislocation Loop in BCC Iron and its Interaction with Point Defects
Error in RUS Measurements Due to Geometric Uncertainties
Estimating the Strengthening Parameters for Irradiated Alloys using Atomic Scale
Evaluating ATF Cladding Mechanical Behavior
Evaluation of Size Effects in Small Scale Mechanical Testing Combining Multi-length Scale Models and Experiments
Examining Microstructural Effects on Tensile Properties in Irradiated Inconel 718 using Miniaturized Tensile Specimens
Examining Microstructures and Mechanical Properties of Neutron and Ion Irradiated T91, HT9 and 800H Alloys
Fabrication and Characterization of Oxide Dispersion Strengthened Nickel Alloys for Advanced Molten Salt Reactor Components
Fatigue Assessment of Metastable Austenitic AISI 347 Pipe Components for Nuclear Engineering
Fracture Toughness of Highly Irradiated RPV Steels
High Throughput Nanoindentation Creep Testing in Nuclear Reactor Steels
Hydride Reorientation Behavior in ZIRLO Using Ring Compression Tests
Hyper-localized Strengthening of Inconel 617 for Very High Temperature Reactor Applications
Impact of Electrolytic Hydrogen Charging on Fatigue Crack Propagation in Reactor Steels
Impact of Thermal Treatment and Irradiation on Mechanical Behavior of Cold Spray Cr Coatings on Zr-alloy Cladding
Investigating Environmentally-Assisted Cracking in 316 Stainless Steel U-Bend Specimens Exposed to Liquid Sodium
Ion Irradiation and Examination of Additive Friction Stir Manufactured 316 Stainless Steel Component
Irradiation and Nanomechanical Performance of Additively Manufactured, In Situ Tempered Grade 91 Steel
Liquid Lead Embrittlement: Experiments and Molecular Dynamics Simulations
Mechanical and Microstructural Characterization of Neutron Irradiated HT-9 Heats at LWR and Fast Reactor Relevant Temperatures
Mechanical Behavior of Bare and Cr Coated Zirconium Claddings During Burst Testing via In-situ Strain Measurements
Mechanical Martensites in Nuclear Steels
Mechanical Properties of Ultrasonically Welded Zircaloy-4 Following Low-temperature Neutron Irradiation
Mechanical Testing and Characterization of an Integrated Welding and Thermal Processing Method on Eurofer97
Micromechanical Aspects of Deformation and Failure of Advanced Iron-Chromium-Aluminum Alloys
Microstructure-aware Predictions of the Creep Response of Metals Subjected to Nuclear Environments
Microstructure and Mechanical Properties of Neutron Irradiated Tantalum-alloyed Ferritic Martensitic Steels
Migration of Intergranular He Gas Bubbles under a Thermal Gradient in Fe by Phase-field Modeling
Modeling Long-term Radiation Effects on the Concrete Biological Shield
Molecular Dynamics Studies of Helium Bubble Effects on Grain Boundary Fracture Vulnerabilities in an Fe70Ni11Cr19-1%H Austenitic Stainless Steel
Multi-scale Modeling of Defect Recombination in Collision Cascade with Molecular Dynamics and Binary Collision Monte Carlo
Musings on Advanced Cladding Qualification
Non-destructive Stress Evaluation in Nuclear Materials by Positron Annihilation Spectroscopy
ODS Cu Materials for Fusion Application Produced by Mechanical Alloying
Optimizing Nuclear Cladding Mechanical Property Output for Hot-cell Testing
Preliminary Studies on Creep Behavior of Commercial FeCrAl Alloy (APMT)
Probing Neutron Irradiation Simulated Damage with Ion Irradiation and In-situ Mechanical Testing
Robust Constitutive Modeling with Artificial Neural Networks
Simulating Irradiation Induced Creep with Coupled Rate Theory and Plasticity Models
Structure-property Evolution of PM-HIP Fabricated Ni-Alloys 625 and 690 Neutron Irradiated to 1 and 3dpa
The Influence of Nanoindentation Orientation on Deformation Mechanisms in Irradiated Fe – P and Fe – N
The Origin of Superior IASCC Resistance of Additively Manufactured 316L Stainless Steel after Hot Isostatic Pressing in Oxygenated BWR Water
The Role of Stress-State on the Failure Mechanism, Strain to Failure and Fatigue Resistance of Zircaloy-4
Understanding the Mechanisms Involved in Chlorine-Induced Stress Corrosion Cracking of Stainless Steel 304 under a Simulated Marine Environment
Welding Repair : Behavior Study of the Heat-affected Zone Regarding the Risk of HIC

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