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Meeting Materials in Nuclear Energy Systems (MiNES) 2021
Symposium Materials in Nuclear Energy Systems (MiNES) 2021
Organizer(s) Todd R. Allen, University of Michigan
Clarissa A. Yablinsky, Los Alamos National Laboratory
Anne A. Campbell, Oak Ridge National Laboratory
Scope MiNES first met in 2019 to serve the fission reactor materials community that grew out of biannual symposia held at the TMS annual meeting (Microstructure Processes in Irradiated Materials – MPIM) and the American Nuclear Society (ANS) annual meeting (Nuclear Fuel and Structural Materials – NFSM). Discussions previously explored at MPIM and NFSM are now part of the MiNES forum and include the following topics:

- Fundamental Irradiation Damage
- Evolution of Material Propertie
- Integrated Phenomena in Reactor Materials
- Advanced/Novel Materials
- Fuels and Actinide Materials
- Nuclear Fuel Cycles

The second installment of MiNES will be held November 8-12, 2021 in Pittsburgh, PA, USA. ANS and TMS alternate sponsorship of the conference and assist an organizing committee headed by three conference chairs in planning the content and venue for each meeting. While held in the U.S., the organizers of MiNES strongly encourage international participation.
Abstracts Due 04/18/2021
Proceedings Plan Undecided

3D-reconstruction via Genetic Algorithms: Application to Metallic Fuel
3D Reconstruction and Quantification of Oxide Nano-porosity in Zirconium Alloys
A First-principles Database Approach to Predicting Trans-uranic Waste Forms
A New Statistical Approach for Atomistic Calculations of Point Defect Formation Energies in Multicomponent Solid-solution Alloys
A Review of Current Understanding of Fluff Formation in Metallic Fuel via EBR-II Data and Modelling and Simulations.
A Study of the Corrosion Behavior of Cold-sprayed 304L Stainless Steel for Dry Storage Canisters
Accessing High Damage Level Microstructures Using Combined Ion and Neutron Irradiation of a 304L Stainless Steel
ACTINIS: Shielded SIMS for Analysis of Highly Radioactive Samples
Additive Manufacturing (AM) of Oxide Dispersion Strengthened (ODS) FeCrAl Using In Situ Oxidation
Advanced Manufacturing for Novel Material Design and Development
Advanced Technology Fuel Accelerated Development at Bangor University
An Investigation of FCCI Using Diffusion Couple Test between UMTZ Alloys and Cladding
Atom Probe Tomography for Nuclear Materials
Atom Probe Tomography Study of Elemental Segregation and Precipitation in Ion-irradiated Advance Austenitic Alloy A709
Atomic Scale Investigation of Thermodynamic and Defect Properties of (U,Pu)O2 Mixed Oxide
Atomistic Calculations on the Effective Bias of Cavities in BCC Fe
Beta Transmutations in Apatite with Ferric Iron as an Electron Acceptor – Implication for Nuclear Waste form Development
Calculation of Irradiation Enhanced Diffusivities Using Centipede
Cavity Formation in Ion Irradiated Fe and Fe-Cr Ferritic Alloys
Chemical Interaction and Incorporation of Lead with Uranium Nitride Fuels
Cluster Dynamics Simulations of Fission Gas and Product (Xe, Ag) Diffusivities in TRISO UCO Fuel Kernels
Cold Spray for Repair of Nuclear Power Plant Components
Comparison of Temperature-dependent Swelling Behavior in FCC Compositionally Complex Alloys and 316H Stainless Steel under Heavy-ion Irradiation
Constructing Multi-component Diffusion under Irradiation in U-Mo Alloys
Contextualizing Dispersoid Evolution within Friction Stir Welded and Ion Irradiated MA956
Correlating Properties of Irradiation Produced Nanoscale Superlattices with Irradiation Condition Parameters
Defect Cluster Configurations and Mobilities in α-zirconium: Implications for Breakaway Irradiation Growth
Defect Clustering in UO2 Doped Systems Studied Using XAS and Neutron Scattering
Deliquescence of Eutectic LiCl-KCl Diluted with NaCl for Interim Waste Salt Storage
Dependence of Sink Strength Effects on Defect Evolution in Dual-ion Irradiated Additive-Manufactured HT9
Design and Operation of an Out-of-pile Liquid Sodium Experimental Facility for Mechanical Testing
Design of a Test System for Hot Hydrogen-facing Components in Nuclear Thermal Propulsion Systems
Determination of Chromium Corrosion Potential in the Na-K-Mg-U(III) Chloride Molten Salt
Determination of Tritium Trapping Mechanisms in the TPBAR Aluminide Coating
Developing Neural Network Model for Automated Analysis of Radiation-induced Grain Growth in UO2
Development and Application of a UN Potential to Defect Properties and High Temperature Elastic Constants
Development and Application of an Interatomic Potential for the Investigation of Mixed Oxide Compounds Containing Americium
Development of a Multicomponent Ideal-solution (MCIS) Free Energy Phase-field Model for Simulation of Nuclear Materials Microstructural Evolution
Discerning the Effects of Solute Additions in FeCrAl on Dislocation Dynamics under Irradiation Using a Machine Learning Object Detection Algorithm
Dislocation Loop Evolution in Fluorite Oxides
Does the Fuel Fabrication Method Have an Impact on the Fuel Performance Microstructure in Uranium-molybdenum?
Dose and Temperature Effect on Dispersoids in Neutron Irradiated Oxide Dispersion Strengthened (ODS) Alloys
Effect of Cr and Temperature on Dislocation Loops in Heavy Ion Irradiated Ultra-high Purity FeCr Alloys
Effect of Damage Rate and Cascade Size on α' Precipitate Stability in Fe-15Cr
Effect of Helium Injection Rate on Cavity Microstructure in Dual Ion Irradiated T91 Steel
Effect of the Inner Liner on Radial Delayed Hydride Cracking
Effects of Heat Treatment, Build Angle and Radiation Type on the Hardness and Microstructure of Inconel 625 and 718 Fabricated via Laser-powder Bed Fusion Additive Manufacturing
Electron Probe Microanalysis of Fuel from EBR-II Experiment X441A: Effects of Varying U:Pu:Zr Elemental Ratios
Explorations in Automated Cavity Detection Using an Expanded Machine Learning Training Data Domain
Fabrication and Properties of Uranium Dioxide-uranium Boride Composites
Fabrication of Potentially High Burnup Annular U-10Zr Fuel by SPS
Finding a Balance in FeCrAl Alloys: Optimization of Alloy Chemistry for Balanced Properties
First-principles Study of the Interfaces between Gamma-U and Uranium Carbide
Fracture Mechanics-based Testing and DCPD in FLiNaK
Free Surface Impact on Radiation Damage in Pure Nickel by In-situ Self-ion Irradiation: Can It be Avoided?
Grain Growth and Mechanical Properties of Nano ZrO2 Oxide Dispersion Strengthened Mo30W
High Throughput Study of Hardening and Void Swelling in Ion Irradiated Compositionally Complex Alloys
How Does PUREX Actually Work and What Do Chemists Do?
IASCC Initiation Testing of ex-PWR Baffle-former Bolts
Identifying Crystalline Phases in Irradiated U-Pu-Zr Fuels Using TEM
Impact of Grain Boundary and Surface Diffusion on Fission Gas Release in UO2 Nuclear Fuel Using a Phase Field Model
Impact of Zirconium Concentration Variation on Metal Fuel Constituent Redistribution
In Situ Mechanical Testing Method for Materials in Gaseous Environments
Influence of Different Heat Treatments and Ion Irradiation on the Microstructural Evolution and Microhardness of Inconel 625 Fabricated via Laser-powder Bed Fusion
Innovative Elaboration Method of ODS Ferritic Steels Reinforced by Y2Ti2O7 Pyrochlore Phase Oxide
Insights into Prediction of Thermodynamic Properties for Chloride Salts for Generation IV MSRs
Instrumentation in Molten Salt Systems: Commercial Availability, Custom Solutions, and Gaps
Irradiation Creep and Fatigue Observed via In-situ Electron Microscopy
Kinetics of SiC Reaction with Water and Oxygen Under Light Water Reaction Conditions
MAX Phases for Nuclear Applications
Mechanical Behavior of Additively Manufactured 316L Stainless Steel and SiC before and after Neutron Irradiation
Mechanical Response of HT9 and T91 under Dual-ion and Neutron Irradiations
Mechanical Testing of Fuel Cladding Tubes
Mesoscale Simulations of Interactions between Dislocation Loop and Point Defects in bcc Iron
Mesoscale YellowJacket: A Phase-field Model for Microstructure Dependent Corrosion of Ni-Cr Alloys by Molten Fluoride Salts
Metal Hydride Moderator Development at Los Alamos National Laboratory
Microstructural Analysis of Oxidized Tristructural Isotropic Particles (TRISO) in Mixed Gas Atmospheres
Mitigating Irradiated Assisted Stress Corrosion Cracking with Minor Refractory Element Modification – A High-throughput Approach Using Compositionally-graded Specimen
Modeling the Mechanisms of Fuel Pulverization Using Cluster and Molecular Dynamics
Molten Salt Thermal Properties Database-Thermochemical (MSTDB-TC) Status and New Assessment of MF-UF4 (M = Li, Na, K, Cs) Systems
Neutron Irradiation Effects on PM-HIP Inconel 625
New Microscopic Insights into the Fuel Cladding Interaction Layer of High Burnup Fuel
Next Steps for Improved Defect Production and Mixing Parameters: Beyond NRT DPA, ARC-DPA and RPA
Novel Nickel-based Alloys for Molten Salt Fast Reactor Structural Applications
Opportunities for Advanced Concepts in Nuclear Fuel Development
Overview of Fuel System Options for Nuclear Thermal Propulsion
Overview of in Reactor Mechanical Testing in the Versatile Test Reactor
Oxidation Performance of High Uranium Density Fuels for Light Water Reactors
Perovskite-derived Cs2SnCl6-Silica Composites as Advanced Waste Forms for Chloride Salt Wastes
Phase-field Simulations of Fission Gas Bubbles in High Burnup UO2 during Steady-state and LOCA Transient Conditions
Phase and Thermodynamic Analysis of Uranium Mononitride in High-temperature Steam Light Water Reactor Atmospheres
Physical Understanding of Radiation Hardening of Neutron Irradiated FeCr Alloys
Plutonium Defect Characterization through Mechanical Deformation
Point Defect Evolution under Irradiation: Finite Size Effects and Spatio-temporal Correlations
Predicting Phase Stability of Potential Actinide-bearing Hollandite Waste Forms Using First Principles Calculations
Pushing towards the Limits in Characterization of Radiation Damage
Quantifying the Impact of an Electronic Drag Force on Defect Production from High-Energy Displacement Cascades in α-zirconium
Radiation-decelerated Corrosion of Nuclear Structural Materials in Gen IV Reactor Environments
Radiation-induced Segregation in Nanocrystalline FeCrNi under Concurrent Grain Boundary Movement
Radiation Damages Bohr’s Metrics: The Elemental Landscape
Radiation Effects and Thermal Stability in Ferritic Steels and High Entropy Alloys
Radiation Enhanced Diffusion (RED) and the Coupled Effects of Irradiation and Corrosion in Fe2O3
Radiation Tolerance of Capacitive Discharge Resistance Welded 14YWT
Rapid Simulation of the Irradiated Microstructure in Flux Thimble Tubes to High Dose Using Ion Irradiation
Recent Advances in Pyroprocessing of Light Water Reactor Fuel
SKAPHIA: Presentation of the Latest Shielded Electron Probe Micro Analysis (EPMA)
Solute Segregation and Precipitation Across Damage Rates in Dual Ion Irradiated T91 Steel
Strengthening Effects across Ultrasonic Additive Manufacturing (UAM) Interfaces
Structural Materials Testing for the Westinghouse Lead Fast Reactor
Study on Role of Irradiation Induced Vacancies and Voids on Strain-induced Martensitic Transformations by Molecular Dynamics
Suppressing Irradiation Instabilities in Nanocrystalline Tungsten through Grain Boundary Doping
Synthesis of UN-U3Si2 Composite Fuels by Spark Plasma Sintering and Properties Characterization
Temperature-controlled Friction Stir Welding: A Potential Crack Repair Technology for 304L Stainless Steel Spent Nuclear Fuel-dry Storage Canisters (SNF-DSC)
The Challenges of α-uranium: Fundamental Understanding of a Past and Future Nuclear Fuel Material
The Effect of Phase Structure on the Aqueous Corrosion of Yttrium Disilicate
The Kinetics and Stability of Alpha Prime (α’) Precipitates in FeCr Binary Alloy under Ion Irradiations
The Role of Alloying Species on Radiation Tolerance of BCC Fe Binary Alloys
The Subtle Effects of Nitrogen on Radiation Effects in Tempered Martensitic Steels
The Xe-100 Advanced Reactor Concept
Thermal Analysis of Advanced Nuclear Fuels during Simulated Off-normal Events
Thermal Annealing and Irradiation Behavior of Ultrafine-grained and Nanocrystalline FeCrAl Alloys
Thermal Diffusivity of Nuclear Materials at the Miniature Scale
Thermal Gradient Effect on the Helium and Intrinsic Defects Transport Properties in Tungsten
Three-dimensional Characterization of Microstructural Features in Oxide Fuels
Three-dimensional Characterization of Pore Evolution in High-burnup U-Mo
Transmission Electron Microscopy of the Uranium-22.5 Atom% Zirconium System Following Casting, Cold-working, and Annealing
Ultra-fine Lattice Wicking Structures Additively Manufactured from Tungsten
Understanding Tritium Permeation in FeCrAl Alloys
Utilization Potential for the Molten Salts Thermal Properties Database – Thermochemical (MSTDB-TC) in Operational and Safety Analysis for MSRs
Wear and Friction Behavior of Fuel Pebbles in Molten Fluoride Salt
What’s Driving the Acceleration of Nuclear Materials Technology?

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