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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Composite Materials for Nuclear Applications II
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
TMS: Composite Materials Committee
TMS: Mechanical Behavior of Materials Committee
TMS: Advanced Characterization, Testing, and Simulation Committee
Organizer(s) Anne A. Campbell, Oak Ridge National Laboratory
Dong Liu, University of Bristol
Rick Ubic, Boise State University
Lauren M. Garrison, Oak Ridge National Laboratory
Peng Xu, Idaho National Laboratory
Johann Riesch, Max-Planck-Insitut Fuer Plasmaphysik
Scope Composite materials are of growing interest for nuclear fusion and fission due to their combined excellent physical and mechanical properties that are compatible with extreme radiation and high temperature environments. With the development of next-generation fission reactors and fusion power, materials that can withstand higher neutron flux/thermal load/thermal mechanical stresses and more aggressive environments in terms of oxidation, corrosion/erosion, and tolerance to transmutation elements are required. This requirement makes it necessary to (i) understand the operational limits and degradation mechanisms of existing composite materials and (ii) develop and qualify new materials designs. There is a strong overlap in materials research between fission and fusion in terms materials design, processing, characterization, and modelling. This symposium aims to bring scientists and engineers together to share ideas and so join the effort in both fields at an international level for the development of these crucial composite materials and to enable collaborations across groups and countries. The design/processing/modelling/joining of the following materials, as well as their physical/mechanical characterization using ex situ and/or in situ techniques, are encouraged:

• Graphite/carbon-based composites for fission and/or fusion (e.g., nuclear graphite, C/C, and novel designs)
• Ceramic-based composites for fusion and/or for nuclear cladding (e,g., SiC-SiC, C/SiC, and novel designs)
• Metal-based composites (e.g., ODS steels, components with protective single- or bi-layer coatings including diamond on fusion components and/or Cr or Cr/Nb on accident-tolerant fuel cladding, tungsten/tungsten composites, laminate systems)
• TRISO fuel (e.g., particles, compacts, and FCM fuel)
Presentations on SiC-related topics will be coordinated with concurrent symposia on ceramics to minimize overlap.

Abstracts Due 07/17/2022
Proceedings Plan Planned:
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

Advanced Modeling for Use in Accelerate Fuel Qualification of Silicon Carbide Composite Cladding
An Innovative Additive Manufacturing Route for Metal Matrix Composites for Nuclear Applications
Characterization of Al/B4C composites fabricated by hot pressing and spark plasma sintering processes
Characterization of defects generated from thermal stresses in SiC/SiC composites
Characterization of the effects of intermediate temperature neutron irradiation on model Fe-Cr alloys
Correlating heterogeneous pore distribution with stochastic fracture in the pyrocarbon buffer layer in TRISO fuel particles
Development and additive manufacturing of ODS IN-718 alloys for nuclear applications
Development and evaluation of dual-purpose coating to SiC/SiC composite accident-tolerant fuel cladding for light water reactors
Development of SiCf/SiC composite materials for fusion applications
Effect of copper fiber in RAFM steel composite on improving the thermal conductivity
Effect of Hot Rolling and High Temperature Annealing on the Microstructure and Mechanical Properties of Hot-Rolled 90W7Ni3Fe WHA
Effect of interfacial features on the strengthening behavior of B4C/Al composites
ENHANCED Shield: A Critical Materials Technology Enabling Compact Superconducting Tokamaks
Ion Beam Synthesis of Nano-Oxides in FeCr: Towards an Understanding of Precipitation in Oxide Dispersion Strengthened Steels
Irradiation Effects in the Composite Phases of Graphite and Carbon-Based Materials
Is There Residual Stress in Tungsten Fiber Reinforced Tungsten Composites?
Mechanistic understanding of hydrothermal corrosion of SiC under irradiation
Microstructure and Mechanical Behavior of Cr coatings for Mitigating Hydrothermal Corrosion of SiC-SiCf Fuel Cladding
Microstructure and Thermophysical Property Characterization of U-ZrHx Fuel Fabricated by Powder Metallurgy
Next-Generation Nuclear Grade Composite Components
Nuclear Graphite as a Core Composite Material
Oxidation effects on the microstructure of nuclear graphite
Oxidation response of irradiated and unirradiated TRISO fuel
Predictive Models for Mechanical Properties of Oxide Dispersion Strengthened Alloys
Progress in the development of tungsten fibre-reinforced copper composites for heat sink applications in plasma-facing components
Recent progress in the development of tungsten fibre-reinforced tungsten composites
Role and structure of HTGR matrix material
Ruthenium and Silver Diffusion in Nuclear Graphite
Status update on Framatome PROtect ATF solutions: Cr-coated M5Framatome and SiCf/SiC cladding designs
Stress rupture of SiC/SiC composite tubes under high-temperature steam: implications for resistance to light water reactor accident
Study of of Thermal Oxidation to Helium Implantation in 316L Stainless Steel
Thermal Properties of Dispersoid-Strengthened Tungsten Alloys for Fusion Applications
TRISO Fuel and Matrix Modeling
Use of Carbon Fibre-Reinforced Carbon in Wendelstein 7-X
W2C-reinforced tungsten: a promising candidate for EU DEMO divertor material


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