ProgramMaster Logo
Conference Tools for 2022 TMS Annual Meeting & Exhibition
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Symposium
Meeting 2022 TMS Annual Meeting & Exhibition
Symposium Materials Systems for the Future of Fusion Energy
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
TMS: Additive Manufacturing Committee
TMS: Computational Materials Science and Engineering Committee
TMS: Mechanical Behavior of Materials Committee
Organizer(s) Jason Trelewicz, Stony Brook University
Kevin Field, University of Michigan
Takaaki Koyanagi, Oak Ridge National Laboratory
Yuanyuan Zhu, University Of Connecticut
Dalong Zhang, Pacific Northwest National Laboratory
Scope The recent National Academies report on “Bringing Fusion to the U.S. Grid” has established an ambitious goal of producing net electricity in a nuclear fusion pilot plant by 2040. Scientific and technical innovations in materials for plasma facing components, structural and functional materials, high temperature superconducting magnets, and the tritium fuel cycle were recognized as essential components to achieving this critical milestone in carbon-free energy production. Exposure of materials to fusion plasmas and their ensuing property degradation has long been recognized as one of the most important challenges facing the future of fusion energy. Bulk damage in structural and blanket materials from aggressive neutron fluxes combined with surface damage on plasma facing components from the high flux, high fluence plasma conditions will ultimately limit material stability, and in turn, component and sub-system lifetimes. A fundamental understanding of the coupling between near-surface phenomena and bulk microstructure evolution under fusion relevant conditions, and its implications for materials performance and sustainability, are thus key scientific drivers for future innovations in fusion materials.

This symposium aims to broadly discuss fusion materials research and the fundamental physics of materials degradation under separate effects testing and coupled extremes involving elevated temperature, stress, irradiation (ion and neutron), plasma exposure, and oxidizing environments. Talks are solicited that cover new structural and functional materials systems, fusion specific applications of materials, fundamentals of radiation damage, novel in situ techniques and other testing approaches, and advances in modeling and theory for fusion materials. Topics of interest include but are not limited to:

• Reduced activation ferritic/martensitic steels, tungsten and refractory alloys, composites and functionally graded materials, and novel radiation-resistant materials including compositionally complex alloys and interface engineered materials. Findings using novel model material analogues for fundamental mechanism exploration are also of interest.
• Advanced manufacturing methods that enable scalable, cost-effective fabrication of fusion reactor components, including but not limited to novel powder metallurgy, additive manufacturing, solid state processing, simulations-informed alloy design and processing, etc.
• Irradiation effects and synergistic effects under coupled extremes using neutron sources, accelerators, multi-ion beams, environmental test cells, and other tests systems including in situ and in operando techniques and innovative algorithms for high-throughput characterization.
• Consideration of off-normal events and associated safety hazards such as the aggressive thermal oxidation and decomposition of plasma facing components in case of air ingress accidents.
• Multiscale modeling and simulation of radiation effects including the fundamentals of gas behavior, design of radiation-resistant materials, and integrated studies on materials performance.
• Cross-cutting materials science for fusion and fission including fusion prototypic neutron experiments for probing materials degradation toward fusion conditions.

Abstracts Due 07/19/2021
Proceedings Plan Planned:
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

Alternative ODS Steel Manufacturing with Gas Atomization Reaction Synthesis (GARS) and Friction-based Processing
Anomalous Precipitation of Cr in Fe-rich Ferritic Steels under Irradiation in Presence of C and N Impurities: First Principles Modeling and Experimental Observations
Characterization of Atomic-scale Defects in Neutron Irradiated Silicon Carbide
Composite Shielding for Advanced Fusion Systems
Deuterium Trapping and Release from Irradiation-induced Voids in Tungsten: Theory and Experimental Validation
Development of Tungsten Heavy Alloy Composites for Fusion Applications
Development, Production, and Qualification of Berrylide Neutron Multiplier Parts for the He Cooled Breeding Blanket
Direct Visualization of Tungsten Oxidation by In-situ Environmental TEM
Effect of Cr and He on Cavity Swelling in Dual-Ion Irradiated High Purity Fe-Cr Alloys
Effect of He Plasma Exposure on Recrystallization and Properties of W
Evaluating the Temperature Dependence of Bubble Bursting Rate for Low Energy Helium Plasma-exposed Tungsten
First-principles Calculations of Tungsten-based Alloys under Fusion Power Plant Conditions
Grain Boundary Softening in Helium Implanted Fine-grained Tungsten
High-performance Superconductors for High Field Magnets for Compact Fusion
In-situ TEM of Quantum De-trapping and Transport of SIA Clusters in Tungsten
In-situ TEM of the Microstructure and He Behavior of AM W Alloys
In-situ Thermal Diffusivity Recovery and Defect Annealing Kinetics in Self-ion Implanted Tungsten Using Transient Grating Spectroscopy
Integrated Multi-physics Modeling of Impurity Migration, Surface Morphology, and Material Evolution in Present and Future Tokamaks
Liquid Metal Compatibility Evaluations for Fusion Applications
Macroscopic Elastic Stress and Strain Produced by Irradiation
Microstructural Examination of Radiation Damage in Tungsten
Microstructural Transitions during Powder Metallurgical Processing of Solute Stabilized Nanostructured Tungsten Alloys
Modelling and Experimental Study of Yttrium Segregation in Smart Alloys as Plasma Facing Materials
Molecular Dynamics Simulations of Hydrogen and Nitrogen Implantation in Tungsten
Multi-scale Model for Segregation of Transmutation-generated Solutes in Neutron Irradiated Tungsten
Multiscale Materials Modeling of Structural Materials and Plasma Facing Components in the Extreme Fusion Environment
N-21: Dislocation Loop Formation in Self-ion Irradiated Ultra-high Purity Fe-Cr Alloys
N-23: Heavy Ion Irradiation Studies on an Additively Manufactured 316LN Stainless Steel at Elevated Temperatures
N-24: Microstructure Deformation and Possible Densification of Tungsten in High Heat Flux Conditions
N-26: Oxide-dispersion-strengthened Steel Processing by Additive Manufacturing of Gas Atomization Reaction Synthesis (GARS) Powders
N-27: Polycrystal Homogenization Modelling Accounting for Channeling in Irradiated Metals and Alloys
N-28: Promoting Radiation Resistance in Metallic Solid Solutions via the Use of Multiple Synergistic Solutes
N-29: Thermal and Mechanical Characterization of W-Cu Composites for Next Generation Fusion Devices
Neutron Irradiated Tungsten Defect, Surface Chemistry, and Microstructural Characterization
Neutron Radiation Enhanced Grain Growth in Tungsten and Tungsten Alloys under Mixed Spectrum Neutron Irradiation
Novel Transitional Layer Structure between Reduced Activation Ferritic Martensitic Steels and Tungsten for Fusion Reactors
NOW ON-DEMAND ONLY - Fabrication of an Oxide Dispersion Strengthened Ferritic Steel Using SolidStir Technology
NOW ON-DEMAND ONLY - Low Temperature Hardening-embrittlement in Neutron Irradiated ODS Steels
NOW ON-DEMAND ONLY - Post-irradiation Examination of High-dose Ion Irradiated MA956 ODS Alloy
Paving the Way for a Fusion Pilot Plant
Promoting Oxide Dispersion Strengthening in Ferritic Steels Made with GARS Powder for High-shear Powder Consolidation and for L-PBF
Recent Progress in Understanding Fundamental Radiation Degradation Processes
Reduced Interstitial Mobility in W Based Transition Metal Ternary Systems
Self-passivating SMART Alloys for a Fusion Power Plant
Strain and Thermal Gradient Effects on the Transport Properties of Intrinsic Defects and Impurities in Tungsten
Synergies between H and He in Radiation-induced Swelling of Candidate Fusion Blanket Materials


Questions about ProgramMaster? Contact programming@programmaster.org