ProgramMaster Logo
Conference Tools for 2022 TMS Annual Meeting & Exhibition
Register as a New User
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:

A Fusion Relevant Engineering Void Swelling Model for 9Cr Tempered Martensitic Steels
Advancing Materials for Fusion Energy: A Department of Energy Perspective
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
Applications of Machine Learning to Analytical Electron Microscopy for Fusion Materials
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
Dislocation loop formation in self-ion irradiated ultra-high purity Fe-Cr alloys
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
Effect of Hot-Rolling and Specimen Size on the Fracture Toughness of WNiFe Tungsten Heavy Metal Alloys
Evaluating the Temperature Dependence of Bubble Bursting Rate for Low Energy Helium Plasma-exposed Tungsten
Evaluation of major and minor solute effects in ferritic alloys through real-time quantification in TEM in situ ion irradiations
Expanding irradiation damage models to fusion conditions: Tackling the multispecies paradigm at high temperatures.
Fabrication of an oxide dispersion strengthened ferritic steel using SolidStir technology
First-principles calculations of tungsten-based alloys under fusion power plant conditions
Fusion Relevant Models of Irradiation Effects of Displacement Damage and He on the Constitutive Properties of Tempered Martensitic Steels
Grain Boundary Softening in Helium Implanted Fine-grained Tungsten
Heavy ion irradiation studies on an additively manufactured 316LN stainless steel at elevated temperatures
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
Ion-Irradiation-Induced Property Change in FeCr: Hardness, Thermal Diffusivity and Lattice Strain
Irradiation dose effects on silicon carbide fiber reinforced SiC matrix composites
Liquid Metal Compatibility Evaluations for Fusion Applications
Low temperature hardening-embrittlement in neutron irradiated ODS steels
Macroscopic elastic stress and strain produced by irradiation
Microstructural Transitions During Powder Metallurgical Processing of Solute Stabilized Nanostructured Tungsten Alloys
Microstructure deformation and possible densification of tungsten in high heat flux conditions
Molecular Dynamics Simulations of Hydrogen and Nitrogen Implantation in Tungsten
Multiscale Materials Modeling of Structural Materials and Plasma Facing Components in the Extreme Fusion Environment
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
Oxide-Dispersion-Strengthened steel processing by Additive Manufacturing of Gas Atomization Reaction Synthesis (GARS) powders
Paving the Way for a Fusion Pilot Plant
Polycrystal homogenization modelling accounting for channeling in irradiated metals and alloys
Post-Irradiation Examination of High-dose Ion Irradiated MA956 ODS Alloy
Promoting Oxide Dispersion Strengthening in Ferritic Steels Made with GARS Powder for High-shear Powder Consolidation and for L-PBF
Promoting radiation resistance in metallic solid solutions via the use of multiple synergistic solutes.
Radiation Effects and Thermal Stability in Ferritic Steels and High Entropy Alloys
Radiation response of MX-type precipitates at elevated temperatures in a novel laser powder blown additive nanostructured alloy
Recent progress in understanding fundamental radiation degradation processes
Reduced interstitial mobility in W based transition metal ternary systems
Response of advanced tungsten alloys to neutron irradiation
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
Temperature and dose effect on the phase separation of neutron irradiated oxide dispersion strengthened alloys
Tensile properties and microstructure of neutron irradiated tungsten fibers for fusion materials application
Thermal and Mechanical Characterization of W-Cu composites for next generation fusion devices
Unraveling the correlation between Fe–Cr nanoscale decomposition and its related magnetic properties.

Questions about ProgramMaster? Contact