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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Interrelated Extremes in Materials Degradation for Fission and Fusion Environments
Presentation Title Vessel material selection and design for the ARC fusion power plant
Author(s) Trevor Clark, Amit Behera, Andrea Lehn-Bradshaw, Lauren Garrison, Claire Saunders, Cody A. Dennett
On-Site Speaker (Planned) Trevor Clark
Abstract Scope Commonwealth Fusion Systems (CFS) is developing the ARC fusion power plant, which features a molten salt breeding blanket and a replaceable integrated vacuum vessel. CFS is undertaking a multifaceted material development roadmap to design materials that will be ready for the first ARC deployment in the early 2030s. The liquid FLiBe blanket motivates a desired steady-state operating temperature of approximately 600-850°C for the structural materials of the vessel. Thus, these components must withstand significant heat and neutron flux while operating at elevated temperatures in a potentially corrosive environment. ARC’s architecture is designed such that this integrated vessel will be replaced over a plant’s lifetime, allowing on-line learning and materials optimization. CFS has developed a suite of technical solutions viable for initial fusion power deployment to target this operational regime where few materials are viable and significant gaps in performance data currently exist.
Proceedings Inclusion? Planned:

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Accelerating Nuclear Materials Development Through Thermal Gradient Ion Irradiations
Analysis of Attenuation Data from the Beltline of the Reactor Pressure Vessel from the Decommissioned ZION 1 Nuclear Power Plant
Assessment of Brittle Fracture of 304L Austenitic Stainless Steel – Material of Light Water Reactor Core Internals
Atomic-scale characterisation of the in-pile corrosion of ZIRLO and X2 Zr cladding alloys
Combined experimental, simulation and computer vision analysis of irradiated microstructure
Detailed post-irradiation examination of harvested PWR baffle-former bolts
Developing dispersion-strengthened tungsten to withstand coupled extremes in fusion reactors
Development of in-situ irradiation creep testing and application to a ferritic-martensitic steel
Dislocation effects on high temperature He embrittlement in iron-based alloys
Disordering and Defect Evolution Processes at Epitaxial Fe3O4 / Cr2O3 and Fe2O3 / Cr2O3 Interfaces under Irradiation
Evaluation of SiC-Based Composite Tubes Under Multi-Physics Environments for Accident-Tolerant Cladding Development
Ghosts in the Mechanism: Corrosion Happening at the Same Time as Other Stuff
High-throughput synchrotron methods for fusion materials research
Impact of Oxidation temperatures and different Helium Irradiation doses-Induced Defects in Fe-18Cr
Integrated Model of Grain Growth in Tungsten Armor Materials Under ARC Plasma Edge Operation Conditions
Integrating high-throughput ex-situ, and in-situ, corrosion testing to investigate irradiation effects on molten salt corrosion
Interphase characterisation and testing on SiC-based CMCs for fusion applications
Investigating the effect of high-temperature ion irradiation on the microstructures and mechanical properties of (Cr,Hf,Ta,Ti,Zr)C and (Hf,Ta,Ti,W,Zr)C compositionally complex carbides
Investigation of Neutron Irradiation Effects on T91 Ferritic/Martensitic Steel for Fusion Reactor Applications
Irradiation temperature and mechanical properties in neutron irradiated ferritic steel
Large-scale finite element models informed by neutron transport and atomistic simulations
Mechanistic understanding of irradiation damage in fuel assemblies: an overview of the MIDAS research programme
Predicting Fracture Toughness Degradation in Irradiated Duplex Structure Stainless Steels Using Data-Driven Methods
Revealing deformation behaviour of tungsten-copper composites for fusion applications
Review of point defect structures in hexagonal close packed metals and across the Periodic Table
Self-ion irradiation induced defect evolution and micromechanical response of W
Simultaneous Irradiation and Corrosion in High Temperature Coolants - The Plot Thickens
Simultaneous or Sequential? Evaluating irradiation-corrosion approaches for Molten Salt Reactor materials
Study on Corrosion Properties of F82H/SUS316L Dissimilar Joints Produced by Fiber Laser Welding or Friction Stir Welding
Three Dimensional Characterization of the Microstructures of PWR Baffle Former Bolts After 40 years in Service
Transmission Electron Microscopy of Second Phase Precipitates in Zirconium Alloys Exposed to Neutron Irradiation in the Advanced Test Reactor
Understanding coupled environments Radiation +
Understanding Synergistic Degradation Mechanisms in Nuclear Materials through Coupled Environment Testing
Vessel material selection and design for the ARC fusion power plant

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