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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Composite Materials for Nuclear Applications
Presentation Title Use of Carbon Fibre-reinforced Carbon in Wendelstein 7-X
Author(s) Jean Boscary, Henri Greuner, Boris Mendelevitch, Gunnar Ehrke, Patrick Junghanns, Reinhold Stadler
On-Site Speaker (Planned) Jean Boscary
Abstract Scope Wendelstein 7-X (W7-X) is an optimized stellarator which started operation in 2015 at Greifswald, Germany. The objective is to demonstrate steady state operation of fusion-relevant hydrogen and deuterium plasmas. The divertor function is to control the plasma power and particle exhaust and is subject to heat loads by convection or thermal radiation from the plasma. The 20mē W7-X divertor is made of 890 target elements (TEs) designed to remove a stationary heat flux of 10 MW/mē and cooled with pressurized water. The TEs are made of the Cu alloy CuCrZr heat sink armored with about 16,000 carbon fibre reinforced carbon (CFC) NB31 flat tiles as plasma facing material. The developed joining technology between tiles and heat sink is a Cu bi-layer technology. The manufacturing route of CFC and target elements will be presented.
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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