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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Composite Materials for Nuclear Applications
Presentation Title Overview of the Westinghouse Accident Tolerant and High Burnup Fuel Program
Author(s) Edward Lahoda, Zeses Karoutas, Luke Olson, Luther Hallman, Kathryn Metzger, Jorie Walters, Michael Sivack, John Lyons, Luke Czerniak, Allan Jaworski, Ben Maier, Robert Terry, Zachary McDaniel, Frank Boylan, Jeffrey Kobelak, Michael Shockling, Magnus Limback, Antoine Claisse, Jonathan Wright, John Ghergurovich
On-Site Speaker (Planned) Edward Lahoda
Abstract Scope Westinghouse is commercializing a portfolio of four EnCore accident tolerant fuel (ATF) designs; chromium-coated zirconium alloy cladding, SiGA silicon carbide (SiC) cladding, ADOPT fuel and UN fuel. This program has been expanded to include development of high burnup, high energy (HBHE) fuel including advanced Zr alloy cladding (AXIOM cladding) and greater than 5% U235 UO2 and ADOPT fuel. Lead test rods have been inserted into the commercial reactor Byron Unit 2 as part of the Spring 2019 and into Doel-4 in summer of 2020 fuel reloads. Cr coated cladding samples are continuing irradiation at the Massachusetts Institute of Technology Reactor. Out-of-reactor testing on UN fuel has revealed a much better response to a potential leaking fuel rod during operation and has replaced U3Si2 as the advanced fuel of choice. Licensing is proceeding in parallel on ATF and HBHE products incorporating the advanced technologies of in-rod sensors and atomic scale modeling.
Proceedings Inclusion? Planned:

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

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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