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Meeting 2025 TMS Annual Meeting & Exhibition
Symposium Composite Materials for Nuclear Applications III
Presentation Title Understanding fission product behaviour in ion-implanted graphite
Author(s) Alex Theodosiou, Philippe Ouzilleau, Ben Spencer, Abbie Jones
On-Site Speaker (Planned) Alex Theodosiou
Abstract Scope A range of nuclear grade graphites were implanted with fission product (FP) ions in order to determine the effect of implantation on the graphite structure as well as investigate the physio-chemical behaviour of the embedded FP ions. The study involves the use of surface analytical techniques such as x-ray photoelectron spectroscopy (XPS) to understand the chemical bonding that occurs and provide insight into the transport/retention mechanisms. In some instances, carbonate formation has been identified through interactions between the host carbon and oxygen impurities. Thermodynamics equilibrium simulations, supported by simplified density functional theory (DFT) calculations, support these findings.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials, Surface Modification and Coatings, Other

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Machine Learning Approach for Predicting Nuclear Fuel Performance with Solid Fission Products
Additively manufactured transition layer design for fusion reactor components
Advancing SiC-SiC Cladding Technology To Support Nuclear Power Generation
Alloying Yttrium Hydride via Powder Metallurgy
Analyzing Advanced Composite Shield Materials for Fusion and Space Reactor Applications
Characterization and Post Irradiation Examination of Rapid Laser Chemical Vapor Deposited SiC Fibers
Computational Simulation on Irradiation Damage in GaAs-Based Betavoltaic Batteries
Corrosion Resistance of Amorphous Fe- and Ni-based Thermal Spray Coatings Exposed to Molten FLiNaK Salt Nuclear Reactor Coolant at 700 °C
Densification of 3D Printed Composite Ceramics via Spark Plasma Sintering
Determination of interface properties of W/EUROFER coating on steel substrate by phased array ultrasonic and fracture mechanical testing
Developing and testing silicon carbide composites for fusion-relevant conditions
Developing methods to predict failure and crack growth using small angle scattering techniques
Development of high-temperature-steam resistant UN via the addition of UB2
Development of Tungsten Composites as Plasma-Facing Materials by Doping Rare-Earth Boride Particulates
Development of tungsten fiber-reinforced tungsten composites for fusion application
Effects of Neutron Irradiation on the Three-Parameter Weibull Analysis of Graphite
Equivariant neural network force fields for 11-cation chloride molten salts system
Framatome’s SiCf/SiC LWR Cladding Design: Developments and Irradiation Programs
Impact of (U,Zr)C Carbon Stoichiometry on Thermal Properties
Irradiation tolerance and molten salt compatibility of beryllium carbide – a candidate high temperature moderator material.
Material bonding layered metallic and ceramic composites using continuous electric-field assisted sintering
Mechanical and irradiation behaviors in low-textured pyrolytic carbon
Microstructural evolution of tungsten boride neutron shielding materials under radiation
On the reduced damage tolerance of fine-grained nuclear graphite at elevated temperatures using in situ 4D tomographic imaging
Perspectives on Raman spectroscopy for carbon-based nuclear materials
Radiation resistance of nanostructured ferritic alloys produced via various methods
SiC-SiC composite cladding materials: effect of the manufacturing process on microstructure and physical properties
Silicon carbide composites for fusion applications
The Development of High-Temperature Continuous Fibre-Reinforced Ceramic Composite Materials for Molten Salt-Facing Environments
Thermal stability of powder metallurgically manufactured tungsten fiber-reinforced tungsten composites at 1450 °C
TRISO Coating Layer Failure Analysis
Tritium Breeder Composites for Fusion Applications
Understanding fission product behaviour in ion-implanted graphite

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