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Meeting 2022 TMS Annual Meeting & Exhibition
Symposium Materials Systems for the Future of Fusion Energy
Presentation Title Composite Shielding for Advanced Fusion Systems
Author(s) Lance Snead, Steven Zinkle, Jason Trelewicz, Ethan Peterson, David Sprouster, Bin Cheng
On-Site Speaker (Planned) Lance Snead
Abstract Scope With significant improvement in High Temperature Superconductors (HTS), a number of projects are adopting HTS technology for compact fusion systems. While HTS irradiation tolerance is unclear, the lack of available space for shielding in compact fusion machines threatens the attractiveness of HTS. Current shield solutions use combinations of high-Z, low-Z, and absorbing materials such as W, H2O, and 10B. Unfortunately, as H2O is avoided for compact reactors and B-compounds suffer from irradiation instability and burnout, there are currently no hi-performance shielding materials to enable the potential performance of HTS technology. This presentation will review a relatively new ARPA-E program aimed at developing specifically engineered metal-matrix and ceramic-matrix composite shielding. These composites are notable insofar as they possess hydrogen densities similar to water through entrainment of neutron-absorbing hydrides. Shield effectiveness, implied stability, and the processing routes to entrain the relatively volatile hydrides in low-diffusivity metal and ceramic matrices will be reviewed.
Proceedings Inclusion? Planned:
Keywords Composites, Nuclear Materials, Process Technology

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

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

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