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Meeting MS&T22: Materials Science & Technology
Symposium Tackling Structural Materials Challenges for Advanced Nuclear Reactors
Sponsorship TMS Corrosion and Environmental Effects Committee
TMS Nuclear Materials Committee
TMS: Advanced Characterization, Testing, and Simulation Committee
Organizer(s) Miaomiao Jin, Pennsylvania State University
Xing Wang, Pennsylvania State University
Karim Ahmed, Texas A&M University
Jeremy Bischoff, Framatome
Adrien Couet, University of Wisconsin-Madison
Kevin Field, University of Michigan
Lingfeng He, North Carolina State University
Raul B. Rebak, GE Global Research
Scope Advanced nuclear reactors will play a critical role in meeting the ever-increasing demand for carbon-free energy worldwide. Compared to light water reactors (LWRs), the proposed advanced nuclear energy systems present an exceptionally harsh environment for the structural materials due to a combination of elevated temperature, increased radiation damage, extended service time, and more corrosive coolants. Furthermore, the growing interest in demonstrating advanced reactor designs requires the qualification process of structural materials to be accelerated. All of these challenges must be tackled in order to realize the desired safety, efficiency, and economics of future nuclear reactors. Meanwhile, rapid progress in other emerging fields, such as additive manufacturing, high-throughput testing and simulation, multiscale modeling, and data analytics provide new avenues to addressing these challenges in structural materials for advanced reactors. This symposium emphasizes not only the evaluation of existing material systems under new conditions, but also the design of advanced structural materials spanning across alloys, ceramics, composites, etc. Both experimental and computational work are welcome.

Abstracts are solicited in, but not limited to, the following areas:
• Novel structural material concepts for enhanced radiation tolerance
• New manufacturing processes (e.g., additive manufacturing)
• High-throughput testing and characterization of materials for nuclear applications
• Multiscale modeling and simulation
• High-throughput simulation and machine learning
• Corrosion in non-LWR and accidental conditions
• Microstructural evolution under extreme environments

Abstracts Due 05/15/2022
Proceedings Plan Undecided

Assessing Materials Susceptibility to Environmentally-assisted Cracking in Advanced Reactor Coolant Environments
Atomistic Calculations and Theoretical Formulations of Thermal Vacancies in Complex Concentrated Alloys
BCC CrAl Thin Film, A Solution for Next-generation High-performance Inert Gas Cooled Nuclear Microreactors
Computer Modeling of Oxidation-induced Grain Boundary Embrittlement in Nickel
Convolutional Neural Networks Screening Radiation-resistant High Entropy Alloys
Correlated Characterization of Ni-based Superalloys Corroded in Uranium-containing Molten Salt Systems
Defect Dynamics and Far-from-Equilibrium Microstructure Evolution in Concentrated Alloys
Deformation Behaviour of Ion-irradiated FeCr – A Nanoindentation Study
Electrochemical Determination Kinetic Properties of Ni2+ and Cr3+/Cr2+ in FLiNaK Molten Salt
Electrodeposition of Functionally Graded Interlayers for Enhanced Divertor/Heatsink Bonding for Fusion Reactors
Elucidating Interfacial Phenomena in Molten Salt Corroded Nickel-Chromium Alloys using Analytical Transmission Electron Microscopy
Heavy Ion Irradiation Response of an Additively Manufactured 316L Stainless Steel
Hierarchical Microstructures: A Potential Route to Enhanced Stability in Structural Materials for Advanced Nuclear Reactors
High-throughput Testing and Characterization of Materials For Nuclear Applications
High Temperature Zirconium Alloys by Titanium Analogy
ICME and ML Modeling Framework of U-10%wt Mo Fabrication Processes
Imaging Local Vacancy Supersaturation in Metals After Corrosion in Molten Salt
Impact of Chemical Short-range Order on Radiation Damage in Fe-Ni-Cr Alloys
In Situ Dual Ion Irradiation of Additively Manufactured Reduced Activation Ferritic-martensitic Steels
Interfacial Interactions between Molten Salt and Structural Materials
Investigation of Fracture Behavior of Nuclear Graphite NBG-18 Using In-situ Mechanical Testing Coupled with Micro-CT
Mechanical Behavior of Additively Manufactured Steels with Monotonic and Graded Microstructures
Mechanistic Calculation of the Effective Silver Diffusion Coefficient in Polycrystalline Silicon Carbide: Application to Silver Release in AGR-1 TRISO Particles
Microplasticity of Irradiated Inhomogeneous Alloys
Microstructural Evolution of High-throughput Additively Manufactured 316L Stainless Steel with Increasing Hafnium Dopants
Microstructural Response of HT-UPS Steel to Thermal Annealing and Neutron Irradiation
Microstructural Self-organization of Phase-separating Alloys during Irradiation into Global Compositional Patterns at Grain-Boundaries and Inside Grains
Multi-scale Modeling of the Mechanical Response of Structural Metals Subjected to Thermo-mechanical Loads and Irradiation: the Role of Microstructure.
Neutron Irradiation Effects in PM-HIP Nuclear Structural Alloys
Novel Refractory High Entropy Alloys for Applications in Extreme Environments
Phase-field Modeling of Radiation Induced Segregation for Multicomponent Alloys: Kinetic Monte Carlo and CALPHAD-Informed Simulations
Phase Field Modeling of Hot Isostatic Pressing for Joining of Dissimilar Metals
Progress Toward Additive Manufacturing of Ferritic-martensitic, In situ Tempered Steels for Nuclear Applications
Quantifying Cr and Fe Dissolution to Understand Stainless Steel Molten Salt Compatibility
Radiation Resistance of MAX and MAB Phase Materials
Structural Material Design for Power Plants Using Additive Manufacturing
Studying Microstructural Evolution in an Oxide Dispersion Strengthened 14YWT Ferritic Steel Tube Manufactured using SolidStirTM Technology
Synchrotron High-energy X-ray Studies of Nuclear Structural Materials
Understanding of Alloying Additions for Design of Gas Atomization Reaction Synthesis Produced Oxide Dispersoid Strengthened Alloys

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