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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Composite Materials for Nuclear Applications
Sponsorship TMS Structural Materials Division
TMS: Composite Materials Committee
TMS: Nuclear Materials Committee
Organizer(s) Anne A. Campbell, Oak Ridge National Laboratory
Dong (Lilly) Liu, University of Oxford
Rick Ubic, Boise State University
Lauren M. Garrison, Commonwealth Fusion Systems
Peng Xu, Idaho National Laboratory
Johann (Hans) Riesch, Max Planck Institute for Plasma Physics
Scope Composite materials are of growing interest for nuclear fissions and fusion 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, characterisation, 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 characterisation 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)

Abstracts Due 07/20/2020
Proceedings Plan Planned:

A Novel Processing Route for ODS Steel by Liquid Metallurgy
Competition between Void Evolution and Amorphization In Radiation-tolerant Nanocrystalline Cu-10at%Ta Alloy
Conformal Tungsten Coatings for Cermet Nuclear Fuel Elements
Corrosion and TEM Analysis of CVD and PVD Coatings for BWR Accident Tolerant Fuel Cladding
Coupled Primary and Secondary Recrystallization in Single Tungsten Fiber-reinforced Tungsten Composites
Development of PVD Cr Coatings for Hydrothermal Corrosion Mitigation of SiC-SiCf Fuel Cladding in LWRs
Development of UN/UO2 Composite Fuels for LWR Applications
Enhanced Microstructural Stability of ARB-processed Cu/Nb Nanolayers Under Heavy Dose Ion Irradiation at Elevated Temperatures
Evaluation and Irradiation of 14YWT Capacitive Discharge Resistance Welds
Experimental Characterisation of the Variation of Local Residual Stresses in TRISO Coatings
Fabrication, Characterisation and Oxidation Resistance of an Innovative Composite Fuel: UN Microspheres Embedded in UO2 Matrix
Improved Techniques for Determining Local Thermal Transport in Composite Nuclear Fuels
Irradiation Induced Forced Chemical Mixing and Local Hardening in Mechanically-processed Immiscible Zr/Nb Multilayers
Mechanical Strength of Explosion Welded Thin Stainless-steel Cladding on Carbon Steel
Novel Fiber Fretting Technique for Tribological Properties of Composite Interphases
Opportunities for Nanostructured Tungsten Alloys in Composite Fusion Materials
Overview of the Westinghouse Accident Tolerant and High Burnup Fuel Program
Post-irradiation Examinations of TRISO Particles Corroded in Molten FLiBe Salt under Neutron Irradiation
Radiation Tolerance and Microstructural Changes of Nanocrystalline Cu-Ta Alloy to High Dose Self-ion Irradiation
SiGA SiC-SiC Composites Development for Accident Tolerant Fuel
Solving the Brittleness Problem of Tungsten - Tungsten Fibre-reinforced Tungsten Composites
Sub-critical Crack Initiation, Coalescence and Propagation in Nuclear Graphite Studied by High-speed Pink Beam Synchrotron Tomography
Synthesis and Irradiation Response of Hetero FeCr - Fe2O3 Interfaces
Tristructural Isotropic (TRISO) Fuel for High-Temperature, Passively-Safe Nuclear Reactors
Tungsten-based High and Medium Entropy Alloys and Composites for Nuclear Applications
Tungsten Fibre-reinforced Copper – A High-Conductivity, High-Strength Composite Material for Plasma-facing Component Applications
Understanding Defect Recovery and Accommodation and Their Implications on Mechanical Performance in Irradiated Nanocomposite Materials
Uranium Nitride Advanced Fuel: An Evaluation of the Oxidation Resistance of Coated and Doped Grains
Use of Carbon Fibre-reinforced Carbon in Wendelstein 7-X
W2C-reinforced Tungsten: A Promising Candidate for DEMO Divertor Material

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