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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Accelerated Discovery and Qualification of Nuclear Materials for Energy Applications
Sponsorship TMS Structural Materials Division
TMS Materials Processing and Manufacturing Division
TMS: Integrated Computational Materials Engineering Committee
TMS: Nuclear Materials Committee
TMS: Additive Manufacturing Committee
Organizer(s) Yongfeng Zhang, University of Wisconsin-Madison
Adrien Couet, University of Wisconsin-Madison
Michael R. Tonks, University of Florida
Jeffery A. Aguiar, Idaho National Laboratory
Andrea Jokisaari, Idaho National Laboratory
Karim Ahmed, Texas A&M University
Scope Materials used in nuclear energy applications usually operate in harsh operating conditions combining high temperature, irradiation, stress, and corrosive environments, with long in-cycle service lives lasting from years to decades. Nuclear materials are purposely processed for controlled chemistries and microstructures to mitigate physical degradation caused by exposure to extreme environments. The requirements for a purposely designed nuclear material must carefully consider a number of functional and safety concerns that exceed the demands for general structural bearing materials.

As the demands on materials are even higher in advanced nuclear reactors, including high temperatures and fluences, the acceleration of nuclear materials development becomes a critical path in the readiness of future nuclear technology. At the bottleneck of developing and qualifying nuclear materials, however, is addressing the traditional materials development for nuclear-grade materials.

Successful stories of accelerated material design have emerged in many other fields other than nuclear energy, and the experiences and knowledge may be transferrable to nuclear materials. In line with the Nuclear Materials Discovery and Qualification Initiative (NMDQI) established by the Nuclear Science User Facilities (NSUF), this symposium focuses on novel tools and approaches that accelerate our understanding of nuclear material behaviors and the development of advanced materials for nuclear energy applications. In particular, we look for tools and approaches that can be used to reduce the time and cost for discovery, advanced manufacturing, testing, and qualification, including both fuels and structural materials. The topics of interest include but are not limited to:

• Modeling and experimental tools for accelerated discovery and optimization of nuclear materials by constructing the processing-structure-property-performance links.
• First-to-learn modeling approaches and strategies that can reduce the number of needed steps to enhance the efficiency and utility of in-pile irradiation tests.
• Physics-based and reduced-order modeling of in reactor materials behavior.
• Advanced manufacturing of nuclear materials with controlled chemistry and selective microstructures.
• Higher throughput characterization techniques that can maximize the efficiency of in-pile and out-of-pile testing.
Abstracts Due 07/20/2020
Proceedings Plan Planned:

A Computer Vision Application for Materials Production Quality Control Screening
A rapid turnaround approach studying helium effects in materials
A Standards Perspective on Nanomechanical Testing to Accelerate Nuclear Materials Development & Qualification
A Superb Void Swelling Resistant Type 316L Stainless Steel Developed by Additive Manufacturing Enabled High Throughput Microalloying
Accelerated Development and Evaluation of Complex Alloys for Nuclear Applications
Accelerated Study of Thermal and Irradiation Creep in Fe-based Multi-principal Element Alloys
Accelerating Separate Effects Irradiation Testing of Nuclear Fuel with Reduced Order Models
An Integrated Approach for Coupling Experimental Data, Physics-based Models, and Machine Learning Algorithms for Predicting the Effective Thermal Conductivity of U-based Fuels
Anisotropic Biaxial Creep Behavior of Textured Nb-modified Zircaloy Cladding
Characterization of As-Fabricated Additively Manufactured Alloy 718 Enhanced by Modern Tools and Machine Learning
Comparison of Void Swelling in Conventional and Novel HT9 Alloys after High Damage Level Ion Irradiation
Compositionally Graded Bulk Specimen: A High-throughput Approach for Nuclear Alloy Development and Qualification
Constructing Multi-Component Diffusion Under Irradiation in U-Mo alloys
Deep Learning Algorithms to Inform and Accelerate Fuels Investigations
Deep Learning for Automated Analysis of Cavities in Transmission Electron Microscopy Images
Defect Cluster Mobilities and Preferred Configurations in α-zirconium: A Comparison of Two Interatomic Potentials
Defect Evolution Mechanism in an Accident Tolerant Fuel U3Si2
Development and Qualification of Ultrafine-grained and Nanocrystalline Steels for Nuclear Applications
Development of Assembly Technique for Fuel Specimens for the MARCH-SERTTA TREAT Irradiation Testing Platform
Development of Sintered High Strength and Thermal Conductivity Cu-Cr-Nb-Zr Alloy for Fusion Components
Dislocation Loop Characterization Using STEM-Contrast Techniques in an Irradiated FCC Alloy
Dislocation Loop Formation in Self-ion Irradiated Ultra-high Purity FeCr Alloys
Effect of Distributed Gas Bubbles on Elastic-Plastic Deformation Behavior in Polycrystalline UMo
Effect of Microstructure and Rolling Treatment on Static Recrystallization Behavior in Monolithic U-10Mo Fuel Foils
Effective Bias for Interstitial Clusters to Cavities in BCC Fe
Effects of Order and Disorder on Primary Damage Production
Enabling Actinide Materials Discovery through Combinatorial Deposition
Evaluation of Creep Deformation of Ferritic/Martensitic (FM) Grade 91 Steel Fabricated Using Wire Arc Additive Manufacturing (WAAM)
Evaluation of the Anisotropic Grain Boundaries and Surfaces of αU via Molecular Dynamics
Experimental Verification of Thermodynamic Predictions of MPEA Compatibility with Fuel and Fission Products for Advanced Reactor Cladding Applications
Helium Effect on Cavity Swelling in Dual-ion Irradiated Fe and Fe-Cr Alloys
High-temperature, High-throughput Ion Irradiation Enabled by Additive Technologies
High-throughput Heavy Ion Irradiation of CrFeMnNi Magnetron-sputtered Combinatorial Thin Film
High power irradiation testing of TRISO MiniFuel-Compacts in HFIR
Improving Irradiation Resistance of Alloys by Controlling Defect Diffusion: A Modeling Perspective
In-situ TEM Heating Chip Experiments to Study Thermal Behavior of U-Zr Metallic Fuel
In-situ Thermal Conductivity Measurement of SiC Composite
Innovative Wet Chemical Approaches for Novel Nuclear Materials
Machine Learning and Atomistic Modeling of Defect Diffusion in Concentrated Ni-Fe Alloys
Machine Learning for Accelerating Property Prediction and Materials Characterization in Irradiated Materials
Machine Learning Perovskites in the Quest for Improved Scintillators
Manufacturing Process Optimization of High-density LEU Targets for Mo-99 Production
Materials selection in nuclear applications a challenge and an opportunity for advanced material design, fabrication and testing.
Mesoscale Modeling of the Effect of Interfaces on Segregation of Point Defects and Solutes and the Patterning of Extended Defects
Microscale Measurement of Elastic Constants in Ceramics Using Picosecond Ultrasonics for High Throughput Characterization and Atomic Model Validations
Modeling and Analysis of the Effects of the Microstructure on U-10Mo Fuel Thickness Variation during Hot Rolling
Molecular Dynamics Study of Cascade Overlap Effects in FCC Ni
Multiscale Characterization of Defects in Ion Irradiated Ceramics for Validation of Atomistic Models
Overview of Advanced Fuels and Materials R&D within the US DOE-NE NEAMS Program
Overview of Nuclear Materials Discovery and Qualification Initiative (NMDQi)
Point Defect Capture Characteristics and Stress States of Dislocation Loops in α-zirconium
Point defect energies in concentrated alloys using ab initio calculations and machine learning
Properties of a Helium Ion Beam Degrader for Implanting SSJ2 Tensile Specimens at the LBL 88-Inch Cyclotron
Proton Irradiation Induced Microstructural Evolution in Compositionally Graded Type 316L Stainless Steel
Qualification of 316L Stainless Steel Components for ASME Pressure Retaining Applications
Role of Composition and Thermal Aging on Corrosion Behavior of Duplex Stainless Steels in Pressurized Water Reactors
Sink Strength Effect on Bubble Formation in Helium-implanted Nanostructured Ferritic Alloys
Synergistic Irradiation and Ageing Effect on Microstructure and Mechanical Properties of Grade 92 at ~700C
Ultrasonic Additive Manufacturing of Zircaloy-4 Plates: Exploring Viability Through Microstructure and Mechanical Property Evaluation
Understanding the Mechanisms of Porosity Development in α-uranium for Metallic Fuel Applications
Using Ion Beam Irradiation for Accelerated Aging of Radiation Effects in Metal Hydrides
What Can We Learn About Nepheline Formation in High-level Nuclear Waste Glasses from Exploratory Data Analysis of Heterogeneous Datasets?

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