Conference Logo ProgramMaster Logo
Conference Tools for Materials in Nuclear Energy Systems (MiNES) 2025
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools

About this Abstract
Meeting Materials in Nuclear Energy Systems (MiNES) 2025
Symposium Materials in Nuclear Energy Systems (MiNES) 2025
Presentation Title Influence of Mn and Co addition on Radiation-Induced Segregation at Grain Boundaries in Austenitic Multi-Principal Element Alloys
Author(s) Daniele Fatto Offidani, Emmanuelle Marquis
On-Site Speaker (Planned) Daniele Fatto Offidani
Abstract Scope Austenitic multi-principal element alloys (MPEAs) are promising for nuclear applications due to their enhanced mechanical strength and corrosion resistance. However, their performance can be influenced by radiation-induced non-equilibrium phenomena such as radiation-induced segregation (RIS). In this study, we explore the effects of Mn and Co additions on RIS behavior at grain boundaries in Fe-Ni-Cr based austenitic alloys. To this end, we systematically characterized the chemical composition changes at grain boundaries, dislocation loops, and grain interiors in three model alloys using transmission electron microscopy and energy dispersive spectroscopy over a range of irradiation doses. The incorporation of Mn and Co alters the evolution of the bulk of the material during irradiation, which in turn impacts the overall segregation behavior, affecting both the magnitude and morphology of segregation profiles compared to the base Fe-Ni-Cr alloy.
Proceedings Inclusion? Undecided

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Methods for In-situ Ion Irradiation Tracking in Coupled Environments
1D Experiment for Material Property Applied to Phase Field Model Simulating Pitting Corrosion of Copper Canister
2D Simulations of Localized Palladium Silicide Inclusions in TRISO Fuel With BISON
A Computational Approach to Investigate the Role of Impurities on Corrosion Behavior in Molten Salt Reactors
A New Route for Fabricating Oxide Dispersion Strengthened (ODS) Alloy Cladding With Friction Extrusion and Pilgering
A Statistically Robust Area Fraction Measurement Tool for Digital Micrographs
A Survey of the Microstructures Observed From MiniFuel Irradiations With Comparisons to Prototypic Irradiated Fuel Microstructures
A Time Dependent Multiaxial Inelastic Constitutive Model for Nuclear Graphite
Additive Manufacturing: Replacing Traditional Metals With Weld Metal
Advanced Characterization of TRISO Fuels
Advanced Diagnostics of Actinides and Fission Products in Molten Salts for a Sustainable Nuclear Fuel Cycle
Advanced Experimental Setup for Simulating Creep-Fatigue in Nuclear Reactor Environments Using Ion Irradiations
Advanced Materials to Enable Timely Deployment of Fusion Energy
Advancing Liquid Metal Compatibility for Fusion Devices
Advancing the Understanding of Mechanical Properties of Diffusion-Bonded 316H and Alloy 617 for High-Temperature Applications
Analysis of Historical HALEU TRISO Fuel in Support of the Deimos Experiment
Application of Miniaturized Specimen Testing Methods for Mechanical Characterization of Irradiated and Non-Irradiated Structural Materials
Arc-Melt Synthesis and Characterization of Advanced Fuels and Surrogates for Next-Generation Reactor Applications
Assessing Pathways to Mitigate Irradiation Induced Hardening & Embrittlement Within the FeCrAl Alloy System
Assessment of a Mechanistically Informed Engineering Scale Fuel Performance Model to Predict Microstructure and Thermal Conductivity Evolution in Oxide Fuel
Assessment of Nuclear Structural Materials’ Dynamic Response for Microreactor Survivability
ATHENA – Off-Gas Abatement Modeling for Nuclear Fuel Reprocessing Using the ARTEMIS Framework
Atomistic Investigation of a<100> Loop Formation in BCC Iron
Atomistic Modeling of Fusion Reactor Materials Using Machine Learned Interatomic Potentials
Beyond Empirical Models: Mechanistic Irradiation-Enhanced Densification in UO2 Pellets
Biaxial Irradiation Creep Deformation of HT9 From 360°C-600°C: Experimental Results and Deformation Analysis
Bridging Uncertainties From Atomistics to Fission Gas Swelling in UN Nuclear Fuel Using Neural Networks and Bayesian Calibration
Can We Predict Failure Probability Distribution in Irradiated Nuclear Graphite?
Carbide Precipitate Dissolution in Advanced RAFM Steels Studied Using In-Situ TEM Ion Irradiation
Cavity Formation Induced by Swift Heavy Ion (SHI) Irradiation in AlxGaN
Cellular Structure Evolution Comparison of LPBF and LDED 316H SS Following Ion Irradiation at Elevated Temperatures
Challenges With In Situ X-Ray Tomography Mechanical Testing on Small Scale Nuclear Materials
Characterization of High-Fluence A-60 Surveillance Capsule From Palisades NPP
Characterization of Hydrogen and Helium at Cavities in Triple Ion Irradiated F82H-IEA With High-Resolution Electron Energy Loss Spectroscopy
Characterization of Mn-Si Rich Precipitates in Neutron-Irradiated Reactor Pressure Vessel Weld Metal
Characterization of Thermal Physical Properties and Defects in Gd-Doped UO2
Characterizing Thermal Pd Transport in SiC-Coated Tristructural Isotropic Particles
Chemical Anisotropy Governs Radiation-Induced Defects in Refractory Multiple Principal Element Alloys
Chemistry-Informed Machine Learning of Molten Salt Properties Across the Periodic Table
Cluster Dynamics Simulations of the Full Xenon-Vacancy-Interstitial Phase-Space in Irradiated UO2
Cluster Dynamics Simulations Using Log-Time Stochastic Algorithms to Study Solute Precipitation and Defect Formation in Alloys Under Irradiation
Combined Dual-Ion Irradiation and Rate Theory Model: A Tool to Predict Reactor-Irradiated Cavity Microstructure
Comparative Study of Oxidation of AISI 304 and AISI 310 Stainless Steels for Use as Heating Element in a Vertical Oxidation Furnace for LOCA Testing
Comparison of the Effects of Carbon Content on Kernel Characteristics Across Small- and Large-Scale Conversions
Complexion Engineered Nanocrystalline Alloys for Improved Radiation Damage Tolerance
Comprehensive Analysis of Constituent Redistribution in U-Zr Fuels
Corrosion Resistance of Single and Bilayer Boronized Coatings on Stainless Steel for Molten Salt Reactor Environments
Coupled Evolutions of Grain Boundaries and Grain Interiors in Alloys Under Irradiation
Creep Damage Quantification in LPBF Ni-Based Alloys Using Non-Destructive Evaluation and Advanced Microstructure Characterization
Defect Evolution Under Irradiation and Stress in bcc-Fe and fcc-Ni: Insights From Defect Rate-Based Long-Time Dynamics
Defect Kinetics and Radiation Induced Segregation in High-Entropy Carbides
Depth-Resolved Elastic, Photoelastic and Stain Properties of Swift Heavy Ion Irradiated MgAl2O4 and Y3Al5O12 Ceramics
Designing Amorphous-Crystalline Composites for Applications Under Extreme Conditions Through Tailoring Nanosized Heterogeneities
Developing a Predictive Tool for Emulating Neutron Irradiation With Ion Beams in Structural Alloys
Development and Validation of UO2 Fission Gas Behavior Modeling Capabilities in BISON
Development of a Sealing Strategy For Tri-Layer Nuclear Fuel Cladding
Development of an Extensive Thermodynamic Database for Modeling and Simulation of MSRs and Examples of Applications
Development of an IASCC Resistant 316L ODS Alloy for LWR Core Internals
Development of an ODS FeCrAl Alloy With Optimized High Temperature Properties and Irradiation Resistance
Development of CrY-YAl Coatings for Enhanced Corrosion and High Thermal Stresses Resistance of Zircaloy Claddings in Boiling Water Reactors (BWRs) Environments
Development of ZrC Coatings Using Halogen and Metal Organic Precursors
Direct Thermal Conductivity Measurement of In-Pile Irradiated U-Mo Particles/ILs and U3Si2-Al ILs by the Suspended-Bridge Method
Distinct Irradiation Resistance Mechanisms of High Entropy Alloys
Dose Rate Effects of Simultaneous Proton Irradiation and Molten FLiNaK Corrosion of Ni-20Cr
Dual-Beam Ion-Irradiation and Larger Heat Properties of Ni-Base Alloys for MSR/FHR Applications
Dual Ion Irradiation Response of Wrought and Additively Manufactured Alloy 709
Effect of Carbon Concentration and M23C6 Precipitates on Creep Properties of Castable Nanostructured Alloys (CNAs)
Effect of Dispersoids and Cavities on H Permeation in Materials
Effect of Helium Bubbles on Screw Dislocation Mobility in Iron: An Atomistic Simulation Study
Effect of Minor Alloying Elements on Helium Bubble Formation in T91 and F82H
Effect of Nickel on Brazed Joints Between TZM and Stainless Steel
Emerging Next Generation Fuels TRISO R&D Efforts at ORNL
Enabling Dissimilar Metal Integration in Molten Salt Reactors via Additive Manufacturing of Graded Transition Joints
Engineering Y2O3/Cr Multilayer Barriers for FCCI Mitigation at High-Temperature Cladding Conditions in Fast Reactor Environment
Enhanced Performance of Chromia-Doped Uranium Dioxide Fuel for Light Water Reactors
Enhancing Tritium Capture in High-Temperature Gas-Cooled Reactors: A Computational Perspective With Density Functional Theory
Establishing a Rigorous Methodology for the Calculation of He/dpa Rates Using FISPACT-II
Evaluating the Mechanisms Underlying Surface Degradation and Hydrogen Isotope Transport in Vanadium for Fusion Fuel Cycling
Evaluation of Hydrogen Isotope Transport in Advanced W Alloysfor Plasma Facing Materials
Evaluation of the SiC Microstructure in TRISO Particles From the AGR Archive
Evolution of Chloride-Induced Stress Corrosion Cracks in 304L Stainless Steels: A Mechanical Perspective
Evolution of Dislocation Structure and Defects in Additively Manufactured 316 Stainless Steels Under Ion Irradiation – A Cluster Dynamics Study
Explicit Modeling of Pd Penetration Using BISON
Exploring Microstructural Evolution and Radiation Effects in Nanotube-Integrated Metal Composite Alloys
Exploring Overcoating Methodology for Integrating TRISO Particles Into Oxide Fuel Matrices
Exploring the Impact of Relative Humidity on Pit Morphology and SCC Crack Growth in SS304H Used for Nuclear Waste Storage Canisters
Fabrication and Mechanical Testing of Li2TiO3 Pebbles With Controlled Microstructures
Fission Product Release Data Summary From the AGR TRISO Fuel Program
Fission Product Transport and Impacts to TRISO Carbide Layers
From the Inside Out: Correlating Radially Resolved Thermal Properties With Local Composition in High Burnup U-Pu-Zr
Fuel Inventory in and on Fusion Materials Studied With Ion and Molecular Beams
Full-Field Modelling of Local Heterogeneities in Thin Films and Coatings With Digital Material Representation Concept
Fundamental Mechanisms and Phase Transformations in Irradiation-Induced Superlattice Formation
Grain Evolution in Tungsten Under Isothermal Annealing
Graphite Application to Charge Selector Device Received With High Power Ion Beams in Facility for Rare Isotope Beams
High-Energy Synchrotron Based X-Ray Computed Tomography {X/XRD/PDF/XRF}-CT: Enabling Multimodal Nuclear Materials Characterization
High-Entropy Alloys for Accelerator Beam Window Applications
High-Throughput Heavy Ion Irradiation and Characterization Methods for Efficient Study of Void Swelling at End-of-Life Doses
High Energy Radiation Induced Damage and Defect Evolution in Aluminum Using Molecular Dynamics
High Temperature Ion Irradiation of Advanced Manufactured Refractory High Entropy Alloys
High Temperature Neutron Study of Uranium Carbide Systems
High Throughput Assessment of Creep Behavior of Additively Manufactured Nuclear Reactor Alloys by Nanoindentation
How Faulted Loops Form in Fluoride Oxide Under Irradiation: Insights From Atomistic Simulations
How Fission Products Affect the Chemistry of Fuel Chloride Salts
Impact of Light and Heavy Elements on Crystal, Thermal, and Micromechanical Properties of Oxide Fuels
Impact of Post-Irradiation Annealing on Mechanical Properties and Microstructure of High Fluence RPV Steel and Weldments
Impact of Ti and Mo Alloying on Fuel Cladding Chemical Interaction (FCCI) in ATR-Irradiated HT9 Clad U-10M (10M=5Mo-4.3Ti-0.7Zr wt.%) Metallic Fuel
Implementing and Validating a Mechanistic UO2 Fission Gas Release Model Using the BISON Fuel Performance Code
Improved Radiography Method for TRISO Buffer Density Measurements
Improving High-Temperature Stability of Fe–Cr–Al–Ti–Y₂O₃ Alloy Through In-Situ Synthesis Nitride and Oxides in Laser Powder Bed Fusion
Impurity-Induced Changes in Polymerization and Fluoroacidity of Molten FLiBe: A Computational Study
In-Situ Investigation of Radiation-induced Cluster Evolution in Surveillance RPV Steels Under Varied Annealing Conditions
In-Situ Synchrotron Study of Non-Equilibrium Processing of Nanoscale Oxide-Dispersion Strengthened Steel Powder
Influence of Carbon Stoichiometry and Microstructure on the Irradiation Performance of Uranium Carbide Fuels
Influence of Dislocations on High Temperature He Embrittlement in Iron-Based Alloys
Influence of Mn and Co addition on Radiation-Induced Segregation at Grain Boundaries in Austenitic Multi-Principal Element Alloys
Initial Sintering Behavior of Ultra-High Temperature Carbide Ceramics
Innovative Approaches to Materials Irradiation Experiments at HFIR
Integrated Experiments and Modelling of Radiation Effects in Fusion Materials
Interstitial Loop Growth, Unfaulting, Migration, and Coalescence in Fluorite Oxides
Inverse Design Applied Towards the Design of New Molten Salt Compositions
Investigating the Effect of Dose Rate on Cavity Evolution Using Ex-Situ and In-Situ Irradiation
Investigation of Ion-Irradiated Beryllium Carbide (Be2C) for High Temperature Nuclear Reactor Moderators
Investigation of ZrC Microstructural Evolution for High-Temperature Coated Particle Fuels
Ion-Induced Defects and Recovery in Semiconductors: An Atomistic Modeling Perspective
Ion Irradiation Effects in Wrought FeCrAl, Additively Manufactured ODS FeCrAl and Conventional Powder Metallurgy ODS FeCrAl
Irradiation-Induced Glissile Interstitial Cluster Interaction in the Region Free From Irradiation Damage
Irradiation Effects on Additively Manufactured 316H Stainless Steel
Irradiation Effects on the Pore Structure of Graphite
Irradiation Hardening Behavior of Brittele Secondary Phases Formed in Tempered Martensitic Steels
Is Metal Additive Manufacturing Ready for Nuclear Service in Light Water Reactors?
Kinetics of Dislocation Recovery in Irradiated vs. Deformed Materials
Laser Beam Welding Properties and Microstructural Characterization of Kanthal APMT FeCrAl Alloys
Long-Term Stability of Ceramic Wasteform Evaluated by Machine Learning Assisted Experimental Calorimetry
Low-Temperature Solution Chemistry for Immobilization of Fluoride Salt Waste in Metal Halide Perovskite Structures
Low Temperature Testing of HT-9 After Neutron Irradiation to Very High Doses
Lower-Length-Scale Informed Mechanistic Models to Predict Creep Rates in Doped UO2 Fuel
Machine Learning Approach to Predict Solubility of Hydrogen in Zircalloy Alloys
Manufacturing Li-Based Tritium Breeder Materials by Volume-Controlled Spark Plasma Sintering
Mapping NEUP Investments Across the Nuclear Fuel Cycle: Illuminating Growth in Advanced Reactor Materials and Design
Materials for Advanced Sodium Fast Reactors: Critical Assessment of Corrosion Resistance and Mechanical Integrity of Alumina-Forming Austenitic Alloys at 700 °C
Materials in Emulated Conditions: Their Limits and our Limitations
Materials Selection and Development Needs for the Infinity Two Fusion Pilot Plant
Mechanisms-Based Creep-Fatigue Analysis of Alloy 617 for Advanced High-Temperature Nuclear Applications
Mechanistic Fuel Performance Simulation of UN TRISO Kernels
Mechanistic Understanding of TRISO Fuel Performance Through Targeted Experiments
Mesoscale Modeling of Restructuring in High Burnup UO2 Fuel
Metal-Water Oxidation Behavior of Niobium-Stabilized Austenitic Stainless Steel Under LOCA Conditions
Microstructural Characterisation of CRUD Formed Under Two-Cycle PWR Conditions
Microstructural Characterization of Irradiated Metallic Fuel Using Complementary EPMA, SEM, and TEM Analyses
Microstructural Evolution of Neutron-Irradiated Additively Manufactured 316L and 316H Stainless Steel
Microstructural Study of Diffusion Bonded Joints for NTP Radiation Shielding
Microstructure and Mechanical Properties of Heat-Treated Grade-92 Stainless Steels Fabricated by Directed Energy Deposition for Nuclear Applications
Mitigating FCCI in Metallic Fuels: Evaluating Cladding Liners Using Multiscale Modeling
Mobility of a<100> Self-Interstitial Atom Dislocation Loops in bcc Fe
Modeling and Simulation of Constituent Redistribution in U-Zr Annular Fuel
Modeling Long-Term Uniform Corrosion of Dry Cask Internal Structures Under Residual Water Radiolysis Conditions
Modeling the Impact of Irradiation on the Corrosion of Molten Salt Facing Structural Components
Molecular Dynamics Simulations of the Phase Reversal of U2Mo Subjected to Irradiation
Morphology and Interaction of 〈100〉 Dislocation Loops Under Ion Irradiation in High Purity Fe-Cr Binary Alloys
Multiscale, Mechanistic Modeling of Ag and Cs Fission Product Transport in TRISO
Near-Term Strategies for Rapid Progress in Fusion Materials Research
Neutron Imaging at the MIT Reactor: Capabilities and Applications in Nuclear Materials
Neutrons in Missouri: An Overview of the University of Missouri Research Reactor (MURR)
Nuclear Fuel Cycle Performance of Microreactors
Nuclear Fuels and Materials Library Development: Structural and Cladding Materials Selection for Neutron Irradiation
Oxidation of the UN-UB2 System During Off Normal Reactor Conditions
Phase-Field Investigation of Grain Growth in U–7Mo Alloy Considering Bimodal Initial Grain Size and Zener Pinning
Phase-Field Modeling of Microstructural Evolution With Inclination-Dependent Grain Boundary Energies
Phase Interactions and Local Mechanical Response in EB-Welded RPV Steels
Physically Informed Cluster Dynamics Modeling of Microstructural Evolution in α-Zr
PM-HIP Processing Delays Irradiation-Induced Cu Clustering and Embrittlement in Grade 91 Ferritic Steel
Preliminary Insights From Post Irradiation Examination of AGR 5/6/7 TRISO Fuel
Preliminary Results of a Computationally Generated Dataset for use in a Constitutive Semi-Empirical Buffer Densification Model for BISON
Pressure-Enhanced Bonding Covalency Anisotropy and Mechanical Anisotropy in Coffinite (USiO₄): A DFT+U Study
Progress Towards Large-Scale, 3D Printed Metallic Hydride Structures
Proton Irradiated Microstructure Evolution of δ-Ferrite in an Austenitic Stainless Steel Weld: Effect of Pre-Existed Precipitate
Pyrolytic Decomposition of Optical Fiber Sensor Coatings and Implications for Radiation-Induced Sensor Drift
Qualification and Codification of Advanced Fuels and Materials: Status and Challenges Related to HTGR Deployment in the USA
Quantification and Uncertainty Determination of Cavities in Under-Focused TEM Images
Quantifying SiC Recession in TRISO Particles Under Steam Oxidation During Accident Scenarios in High-Temperature Gas-Cooled Reactors (HTGRs)
Radial Examination of Xenon Fission Gas Bubbles in As-Irradiated and Post-LOCA Tested High-Burnup UO2 Fuel using S/TEM-EDS
Radiation-Tolerant Cr-Al2O3-Y2O3 Based Multilayer Tritium Permeation Barriers (TPB) for Fusion Reactors: Microstructural and Functional Insights via Proton Irradiation
Radiation Effects on the Mechanical Properties of Austenitic and Ferritic Stainless Steels
Radiation Resistance of Nanostructured Ferritic Alloys With Tailored Grain Boundary Characteristics Produced via Various Methods
Recent Innovation in Scanning Electron Microscope (SEM) In-Situ Extreme Mechanical Testing in Nuclear Environments
Reproducibility in Studies of Metal Behavior in Gen-IV Irradiation-Corrosion Environments
Role of He Bubbles on Deformation Microstructure in Irradiated Alloy 625
Scientific Machine Learning Estimation of Material Thermal Diffusivity From Non-Destructive Evaluation
Short Range Order in High Entropy Carbides and its Impact on Radiation Resistance and Mechanical Properties
Simulated Fuel Cladding Chemical Interaction Diffusion Between Lanthanide Fission Products and HT-9 at Sub-Eutectic and Eutectic Temperatures
Simulation of Coupled Radiation-Induced Segregation and Cavity Growth in Alloys
Simulation of Xenon and Silver Diffusion in Irradiated TRISO (UO2 and UCO) Fuel Kernels
Simulations of Matrix Crack Deflection at OPyC-SiC Interface
Sintering Additives for Improved Densification and Microstructure Control of UN Fuel Kernels
Stabilizing Nanocrystalline Microstructures Dynamically With Grain Boundary Phase Transformations
Stress-Gradient Effects on Radiation-Induced Segregation in Ion-Irradiated 316L Stainless Steel
Surface Finishing-Induced Microstructural Evolution and Deformation Mechanisms in Austenitic Stainless Steels: Role of Stacking Fault Energy
Swelling and Fission Gas Release of U-10Mo and U-17Mo Following Neutron Irradiation at 250 – 500°C
Swift Heavy Ion Irradiation Capabilities of Nuclear Materials at AMIS
Synergistic Insights From Ion Irradiations for the Development of Advanced Fission and Fusion Steels
Synthesis and Stability of Nanoporous Ultra-High Temperature Ceramics at High Temperatures
Synthesis of Uranium Dinitride by Ammonolysis of Uranium Tetrafluoride
Tailoring PM-HIP FeCrAl Alloys for Accident-Tolerant Fuel Cladding Materials
Temperature and Dose Effects on Dislocation Loops in Self-Ion Irradiated High-Purity Iron and Fe-Cr Alloys
Testing Refractory Compositionally Complex Alloys for Phase Stability in High Radiation Environments
The Alloying Effects on Vacancy Diffusion in Austenitic Stainless Steel by Gaussian Process Regression
The Effect of Alloy Composition on Mechanisms of Helium and Vacancy Diffusion in MoNbTaTi High Entropy Alloys
The Effect of Fe3+/He+ Ion Beam Irradiations on Local Deformation Behavior in Tempered Martensitic Steels
The Evolving Morphology of Porosities in Ni-Cr-Based Alloys Under Molten Salt Corrosion
The Hydrogen Effect on Swelling in F82H Revealed by Shuttered Triple-Ion Irradiation
The Impact of Manufacturing Conditions on TRISO Fuel Performance
The International Conference on Nuclear Security: An Opportunity to Engage the Next Generation of Nuclear Professionals
The Ion Irradiation Response and Precipitate Stability of V-4Cr-4Ti Alloys for Fusion Structural Applications
The Mechanisms Behind Irradiation-Induced Grain Subdivision
The Role of Grain Boundary Defect Chemistry on Grain Growth of Cr Doped UO2
The Use of Powder Injection Molding for the Mitigation of Room Temperature Oxidation of Uranium Mononitride
Thermal and Irradiation Creep in Pure Metals and Martensitic Steels: A Statistical-Thermodynamic Model for the Coupling of Point-Defect Diffusion With Dislocation Plasticity
Thermal, Mechanical, and Transient Testing of Fission Gas-Bearing UO2 With High-Burnup Microstructures Under LOCA and RIA Conditions
Thermodynamics, Synthesis, and Characterization of Multi-Component Uranium Carbides for Nuclear Systems in Extreme Environments
Ti Alloy Application to Water Beam Dump Device Received With High Power Ion Beams in Facility for Rare Isotope Beams
Toward High‑Temporal‑Fidelity 3D Microstructure Reconstruction for In Situ Radiation Experiments: Validation of Methods
Toward Qualification of PM-HIP Structural Alloys for Irradiation-Facing Environments
Towards a Better Understanding of Irradiation Effects on the Corrosion of Model Ni-Base Alloys in a Molten Chloride Salt
Transport Properties of Chloride Molten Salts From Machine Learning Interatomic Potentials
U-Cr-O System Study in Support of Chromium doped UO2 Manufacturing
Ultrafast Time-Resolved Electron Diffraction Studies of Radiation-Damaged Materials
Understanding Molten Salt Corrosion via Atomistic Modeling
Understanding the Chemical Evolution of the ZrB2 Fuel Coating of the Integral Fuel Burnable Absorber During Irradiation
Use of CALPHAD Based Tools to Solve Material Challenges in the Nuclear Industry
Use of Data Mining, Machine Learning and Computational Fluid Dynamics for the Fabrication of Coated Particle Fuel
Weibull Fracture Statistics for Ceramic Fuel Pellets
Questions about ProgramMaster? Contact programming@programmaster.org | TMS Privacy Policy | Accessibility Statement