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Meeting 2026 TMS Annual Meeting & Exhibition
Symposium Interrelated Extremes in Materials Degradation for Fission and Fusion Environments
Presentation Title Development of In-Situ Irradiation Creep Testing and Application to a Ferritic-Martensitic Steel
Author(s) Dave Lunt, Benjamin Poole, Thomas Hughes, Philipp Frankel, Ed Pickering, Samir de Moraes Shubeita, Allan Harte, Cory Hamelin, Chris Hardie
On-Site Speaker (Planned) Dave Lunt
Abstract Scope Understanding of material degradation mechanisms in a nuclear environment, such as STEP, is currently focussed on testing parameters (mechanical, thermal or irradiation) in isolation. This lack of a realistic assessment of component performance leads to materials for fusion being considered for longer than necessary. Here, we detail the commissioning and implementation of synergistic irradiation-thermomechanical testing capability at Dalton Cumbrian Facility, which has been developed as part of the materials validation strategy for STEP. This capability has been used to compare the thermal and irradiation creep response of a ferritic-martensitic steel under the anticipated fusion environment. Developing capability to improve the understanding of irradiation-enhanced creep is important for timely qualification of high temperature materials for STEP, given that long-term performance phenomena is not well understood. Furthermore, deformation has been quantified at multiple length-scales using in-situ optical digital image correlation and deformation mechanisms determined from ex-situ high-resolution strain mapping at the grain-scale.
Proceedings Inclusion? Planned:
Keywords Characterization, Iron and Steel, Nuclear Materials

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
Design Kinetic Parameters for Improved Resilience of Materials Under Irradiation
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
He Ion Irradiation Induced Defect Evolution and Micromechanical Response of W
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
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
Predicting Fracture Toughness Degradation in Irradiated Duplex Structure Stainless Steels Using Data-Driven Methods
Review of Point Defect Structures in Hexagonal Close Packed Metals and Across the Periodic Table
Simulated Ex-Service SFR Fuel Cladding for Characterization of Degradation Under DGR Groundwater Conditions
Simultaneous Irradiation and Corrosion in High Temperature Coolants - The Plot Thickens
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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