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Meeting 2022 TMS Annual Meeting & Exhibition
Symposium Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials
Sponsorship TMS Structural Materials Division
TMS: Mechanical Behavior of Materials Committee
TMS: Nuclear Materials Committee
Organizer(s) Dong Liu, University of Bristol
Peng Xu, Idaho National Laboratory
Simon C. Middleburgh, Bangor University
Christian Deck, General Atomics
Erofili Kardoulaki, Los Alamos National Laboratory
Robert O. Ritchie, University of California, Berkeley
Scope Understanding the mechanical behaviour and performance of nuclear fuel and structural materials in harsh environments in a mechanistic and predictable manner is vital to ensure the safety and regulate nuclear energy systems, from current light water reactors to future advanced reactor systems. It is desirable to reduce the uncertainties in the margins to failure to enhance performance for current reactor fleets to improve their economics, safety, and reliability. The lack of property data of existing materials in new operating regimes often limits the potential to optimise performance (e.g. extend fuel burnup). For the next generation of reactors, novel nuclear fuels and materials need to be qualified and its irradiation performance needs to be assessed. The traditional new fuel qualification approach involves two to three iterations of irradiation and post irradiation examination, typically taking two to three decades to gather enough data for licensing. To meet the aggressive schedule for advanced reactor deployment within the next decade, the fuel qualification process needs to be expedited and requires a paradigm shift. Advanced testing methods and high throughput testing capabilities coupled with a mechanistic understanding from modelling and simulations (including machine learning and data analysis) are key to this shift. To date, a significant amount of research effort has been directed to this field but more is now necessary. It is time to bring together all the researchers from academia, national and international research institutes and nuclear industry to share and discuss the most recent advances in mechanical testing for advanced material systems for nuclear energy.

Topics of interest include, but not limited to the mechanical behaviour and properties of:
• Accident tolerant fuel and advanced technology fuel (ATF) systems (including both near term and advanced concepts – pellets and cladding).
• Materials produced by non-conventional and advanced manufacturing methods
• Novel alloy designs and metal composites (experimental and computational designs).
• Advanced ceramics and ceramic matrix composites concepts’ properties and in-reactor behaviour.
• Environmental separate effects and coupled behaviour such as irradiation and/or coupled with corrosion causing embrittlement, hardening, stress corrosion cracking etc.
• High throughput testing and rapid qualification methods including coupling to modelling methods.
• Advanced testing methods including micromechanical testing, in-situ techniques coupled with x-ray, synchrotron, and neutron imaging and diffraction
• Improve understanding via modelling and simulations, machining learning and data analysis

Abstracts Due 07/19/2021
Proceedings Plan Planned:

A Case Study on Radiation-induced Degradation of High-entropy Alloys
A Combined S/TEM and APT Study of the Effects of Sink Strength and Radiation Parameters on Microchemical Evolution in Irradiated Additively Manufactured HT9
A Comprehensive Study of Responses of SiC Materials to Neutron Irradiation for ATF Cladding Application
Accelerated Study of Proton Irradiation Creep and Hardening in Fe-Cr-Ni-based Multi-principal Element Alloys
Accelerating Advanced Fuel Development and Analysis by Combining Modelling and Experiment
Accelerating Separate Effects Irradiation Testing of Nuclear Fuel with Reduced Order Models
Advanced Characterization and Multiscale Testing for SiC Ceramic Matrix Composite Cladding as Accident Tolerant Fuel Candidate Materials for LWR Applications
Advanced Characterization of Material Degradation Processes
Advances in SiGAŽ Development for Nuclear Applications
Assessment of Local Deformation Behavior in Mesoscale Tensile Specimens via Digital-image Correlation
Atomistic Simulation Study of the Effect of Hydride Morphology on the Ductility of Polycrystalline Zirconium
Behavior of Compositionally Complex Ceramics in Molten Salt Environments Under Biaxial Stresses
Characterizing and Testing High Dose Neutron Irradiated Materials for Cladding Applications
Chemical Redistribution of Alloying Elements through Oxide/Metal Interface of Irradiated ZrNbFe Alloys and Its Implication on Corrosion Behavior
Comparison of Neutron Irradiation Effects in PM-HIP and Cast Grade 91 Steel
Compositionally Graded Specimen: A High-throughput Approach for Nuclear Material Development
Creep Performance of IronClad Accident Tolerant Fuel Cladding
Deconvoluting Properties of Additively Manufactured Alloy 718 Utilizing Coupled Microscopy and Machine Learning
Deformation Behavior of Helium Irradiated Nano-pillars Containing a Helium Gas Bubble Superlattice
Degradation Modes of Core Materials under Multiple Components of the Reactor Environment
Development of Advanced Nuclear Fuels for Current and Next Generation Reactors
Development of Direct and Indirect Ab Initio Radiation Damage Models
Development of PVD Cr Coatings for Hydrothermal Corrosion Resistance of SiC-SiCf Fuel Cladding in LWRs
Diffusion in Doped and Undoped Amorphous Zirconia
Dimensional Recovery of SiC Swelling in LWR Accident Scenarios
Dose Rate Effects on Irradiation-enhanced Precipitation in FeCrAl Alloys
Dynamics of Helium Bubbles during Thermal Annealing: A Data-driven Approach
Effect of Free Surface on Displacement-cascade Damage in Neutron Irradiated Nickel
Effect of Microstructure and Rolling Treatment on Static Recrystallization Behavior in Monolithic U-10Mo Fuel Foils
Effect of Nb Alloying and Thermo-mechanical Processing on the Anisotropic Biaxial Creep Behavior of Zircaloy Cladding
Effect of Residual Strain on Short Time Oxidation Behavior of Machined 304 Stainless Steel in High Temperature, High Pressure Deaerated Water
Effective Bias of Cavities in BCC Fe: As Revealed by Atomistic Calculations
Effects of Low-temperature Neutron Irradiation, Hydrogen Charging, and Post-weld Heat Treatment on Tensile Properties of Welded Zircaloy-4
Effects of Steel Composition and Grain Size on Diffusion with Neodymium
Examination of Hardening in Ion Irradiated FeCr Alloys by Nanoindentation
Fabrication and Thermophysical Properties of (U,Zr)C; A High Uranium Density Fuel Candidate for Nuclear Thermal Propulsion Reactors
Fiber/Matrix Debonding of SiC/SiC Composites Evaluated Using the Micropillar Compression
Formation of UN in U-Mo Systems by Mechanical Alloying
Helium Implantation Responses of Co-deposited Copper-tungsten Nanocomposites
High Temperature Mechanical Testing of Uranium Fuel Pellets
Hydrogen Accommodation in the TiZrNbHfTa High Entropy Alloy
Impact Fretting Wear Behavior of Cr-alloy Coating Layer for Accident-tolerant Fuel cladding
Impact of Lithium Accommodation on Defect Chemistry in ZrO2
Implementation of a Damage Degradation Function for Creep Predictions within a FFT-based Framework
In-cell Mechanical Properties of Cladding Geometries to Evaluate Accident-tolerant Fuel Concepts
In Situ Microstructural Evolution in Face-centered Cubic Compositionally Complex Alloys under Dual-beam Heavy-ion Irradiation
In Situ Study of High Temperature Mechanical Behavior of Irradiated FeCrAl Alloys
Influence of the Bulk Chemical Composition on the Microstructure Evolution of Irradiated Chemically-tailored Nuclear RPV Steels
Interface Characterization of an Explosion Welded Stainless Steel-clad Plate for Neutron Irradiation Studies
Investigating the Thermophysical Properties and Key Contributions to the Thermal Conductivity of Different Nitride Systems
Investigation of Degradation Mechanism of Accident Tolerant Fuel (ATF) Coated Cladding Concepts during Interim Storage and Transportation of Used Nuclear Fuels
Investigation of Elemental Segregation and Precipitation in Ion-irradiated Advance Austenitic Alloy A709 Using Advance Techniques
Investigation of Ion Irradiation Effects on Mineral Analogues of Concrete Aggregates
Ion Irradiation Effects on Microstructure Evolution and Mechanical Properties of Silicon Oxycarbide
Irradiation Response of a Metastable High Entropy Alloy
Manufacturing of Oxide Dispersion Strengthened (ODS) Steel Fuel Cladding Tubes Using Cold Spray Technology
Material Degradation Analysis through Machine Learning-based Information Extraction from Legacy SFR Metallic Fuel Performance Data
Mechanical Characterization of Neutron Irradiated HT-9 Heats at LWR and Fast Reactor Relevant Temperatures
Mechanical Testing of Surrogate Molybdenum Alloy CERMET and Zirconium-carbide CERCER Composites Up to Intermediate Temperatures
Mesoscale Modeling of the Relationships between Microstructure and Mechanical Properties in the Porous Pyrocarbon Buffer Layer for TRISO Particle Fuel
Micromechanical Testing of Femtosecond Laser Machined Tensile Samples of Varied Geometries
Microscale Thermal Conductivity and Residual Stress Measurements in TRISO Particle Coatings
Microstructural and Mechanical Properties of Hot Deformation Behavior of Zr-4 Alloy
Microstructural Evolution and Hardness Changes in Ion Irradiated Ni-based Superalloys
Microstructure and Surface Chemistry of FeCrAl Alloys Accident Tolerant Fuel Cladding Subjected to Fast Heating Rate in Aqueous Environment
Microstructure Characterization and Micro-mechanical Properties of 14YWT Tubing after Proton Irradiation
MiniFuel Experimental Capability at Oak Ridge National Laboratory
Modeling High-temperature Corrosion of Zirconium Alloys Using the Extended Finite Element Method
Neutron Irradiation Effects on a Powder Metallurgical Processed SA508 Pressure Vessel Steel
Neutron Irradiation Effects on Mechanical Anisotropy in Alloy 625
On the Diffusion of Hydrogen Atoms towards Notch Tips in Zirconium Polycrystals: A CPFE Analysis
Phase-field Simulations of Fission Gas Bubbles in High Burnup UO2 to Inform Engineering-scale Fuel Performance Modeling
Phase Field Fracture Study of the Effect of Gas Bubble on Fracture at U-Mo/Zr Interface
Phase Stability in FeCrAl Alloys: Mapping the Miscibility Gap and Understanding the Impact of Alpha Prime Precipitation on Material Properties
Progress in Accident Tolerant Fuels Research in the UK
Progress on Experimental Investigation of Degradation Mechanisms of ATF Coated Cladding under Transient Conditions
Radiation-Induced Segregation in Binary Alloy Systems Examined Via Phase Field Simulations
Role of Powder Microstructure and Mechanical Properties on Deposition and Properties of Cold Spray Cr Coatings
Simulation of Shearing-induced Edge and Interfacial Fractures in U-10Mo Monolithic Fuel Plates
Size-dependent Radiation Damage Mechanisms in Nanowires and Nanoporous Structures
Small Scale Mechanical Testing of Irradiated Cladding and Fuel for Nuclear Reactors
Strain Rate Sensitivity Studies of Commercial FeCrAl Alloy
Stress Distribution of Disk Geometry Under Three-point Bending Tests to Evaluate Mechanical Properties of Neutron-irradiated Tungsten for Future Fusion Devices
Study of Microstructure, Hydrogen Solubility and Corrosion of Ta-modified Zr-1Nb Alloys for Nuclear Applications
Study of the Irradiation Induced Microstructure and Mechanical Properties in Low Alloyed Ferritic Steels
Synergistic Effect of Negative Stacking Fault Energy and Irradiation-induced Defects on the Deformation Behavior of Fe-Ni-Cr Alloys
Temperature Sensitive Dislocation Dynamics Modeling of Hardening and Embrittlement
The Interaction between an Extended Edge Dislocation and a Helium Bubble in Copper
Thermal Creep Models Derived from a Comprehensive Multiple Heat 9Cr Tempered Martensitic Steels Database
Through-thickness Microstructure Characterization in a Centrifugally Cast Austenitic Stainless Steel Nuclear Reactor Primary Loop Pipe Using Time-of-flight Neutron Diffraction
Ultrasonic Additive Manufacturing of Zircaloy-4 Plates: Exploring Viability Through Microstructure and Mechanical Property Evaluation
Understanding of the Nature of MNSP Clusters and Their Impact on Mechanical Properties through the Influence of Irradiation Temperature and Ni Content
Unraveling the Early Stage Ordering of Krypton Solid Bubbles in Molybdenum: A Multi-modal Study
Uranium Nitride/Uranium Boride Composite Materials

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