ProgramMaster Logo
Conference Tools for 2024 TMS Annual Meeting & Exhibition
Login
Register as a New User
Help
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Symposium
Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Accelerated Qualification of Nuclear Materials Integrating Experiments, Modeling, and Theories
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
Organizer(s) Tianyi Chen, Oregon State University
Assel Aitkaliyeva, University of Florida
Antoine Claisse, Westinghouse Electric Sweden
Caleb Clement, Westinghouse Electric Company
Michael W. Cooper, Los Alamos National Laboratory
Eric M. Focht, US Nuclear Regulatory Commission
David Frazer, Idaho National Laboratory
Lingfeng He, North Carolina State University
Walter J. Williams, Idaho National Laboratory/Nuclear Regulatory Commission
Scope Rapid development, evaluation, qualification, and licensing of nuclear materials are critical to deploying advanced nuclear power systems to timely meet zero-emission goals. The response of reactor fuel, core, and structural materials to irradiation, among other environments, is vital to the system's performance. These four steps can be accelerated from the traditional way by focusing on understanding the governing mechanisms of the thermomechanical properties, which often rely on the microstructure and microchemistry. Advanced lower-length modeling methods allow to systematize the prediction of promising materials while advanced manufacturing techniques broaden the envelope of microstructure engineering and enable rapid prototyping. Continuous development of accelerated testing technologies is accompanied by advances in material modeling, in-situ and ex-situ characterization techniques for microstructure and properties. The resultant abundant and high-quality experimental and lower-length modeling data are designed to provide input to develop and verify engineering-scale modeling and simulation tools, essential to modern material system design, integral performance analyses, and support materials qualification and licensing. Key to demonstrating the suitability of such experimental techniques to understanding material performance in an applied setting is the integration with mechanistic models and quantification of uncertainty.

This symposium will focus on recent results from nuclear-material development and qualification programs worldwide and cover fundamental and applied science aspects of accelerated nuclear materials testing for fission and fusion reactors. Presentations integrating experiments with theory, modeling, and simulation to develop the methodologies for accelerated qualification and enhance our understanding of environment induced degradation in materials are especially encouraged.

Abstracts are solicited for (but not limited to) the following irradiation program topics:
- Accelerated and in-situ material testing
- Development and application of modeling in supporting qualification and licensing
- Microstructure and property changes in response to service conditions
- Performance prediction based on processing-microstructure-property correlations
- Fundamental science of radiation damage and defect processes

Abstracts Due 07/15/2023
Proceedings Plan Planned:
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A Diffusion-controlled Creep Model in Monolithic Umo Fuels Under Irradiation
A Novel In-situ Miniature Creep Tester for Evaluation of New Cladding Alloys.
A Spatially Resolved Scale-bridging Model for Point-defect Clustering Under Irradiation
A Study in the Thermal Transport Properties Related to Microstructure of Irradiated Annular U-Zr Metallic Fuels
Accelerated Fuel Qualification using In Situ Neutron Diffraction
Accelerated Nuclear Materials Development through Additive Manufacturing and Accelerated Materials Testing
Accelerated Nuclear Materials Qualification and the Swedish SUNRISE Programme
Accelerating the Qualification of High Temperature Structural Materials for Nuclear Reactor Applications
Advancing Silicon Carbide Composite Modeling Within the Accelerated Fuel Qualification Framework
An Approach for In-situ Loading and Corrosion Testing of Accident Tolerant Fuel Cladding
An Integrated Statistical-thermodynamic Model for Fission Gas Swelling and Release in Nuclear Fuels
Application of the Internal State Variable (ISV) Constitutive Model for Creep-fatigue-induced Damage of Advanced High-temperature Nuclear Reactor Steels
Assessment of Effective Elastic Constants of Irradiated U-10Mo Fuel Microstructures
Atomistic-scale Simulations of Creep in Uranium Oxide Nuclear Fuel
Atomistic Modeling of Fission Gas (Xe) Diffusivity at UO2 Grain Boundaries
Challenges Related to Traditional Qualification of Fuels and Materials for Nuclear Reactors
Comparison of Cavity Microstructures from BOR-60, FFTF and Dual-ion Irradiations up to 208 dpa in T91 steel
Comparison of High-dose Microstructure Evolution in Ferritic-martensitic Steels Across Reactor Environments
Current Status on Dislocation Modeling at the Atomic Scale in UO2
Determination of the Radiation Induced Athermal Diffusivity in Uranium Mononitride from an Integrated Approach
Development of a Fission Gas Swelling Model for U-Mo Fuel Incorporating Fission Density, Grain Size, Fission Rate, and Coolant Inlet Temperature
Development of Microscale In-situ Corrosion and Irradiation Experiment
Development of Radiometry-based Instruments for Rapid Thermal Property and Microstructure Characterization, and the Application on Advanced and Additive Manufacturing Components
Dose Dependence of Grain Boundary Radiation-induced Segregation in Fe-Ni-Cr Alloys
Dust Particle Impact on Plasma-facing Materials in Tokamaks: Insights from Molecular Dynamics Simulations
Dynamical System Analysis of Time-accelerated UO2 Fission-gas-release During Power Ramp Transients
E-10: Microstructural Characterization of Harvested High Dose Zorita Light Water Reactor Internals by Atom Probe Tomography and High-resolution TEM
E-11: Monte Carlo Modelling of Neutron Irradiation Displacement Damage in Uranium Mononitride (UN) Fuel When Used in A Small lead-cooled Fast Reactor
E-12: Post-irradiation Examination of High-dose Ion and Neutron Irradiated MA956 ODS Alloy
E-13: Proton Irradiation-induced Cracking and Microstructural Defects in UN and (U,Zr)N Composite Fuels
E-14: Quantifying the Spatial Distribution of Primary Radiation Damage in Real Materials
E-15: Radiation Induced Segregation around Helium Bubbles in Reduced-Activation Ferritic/Martensitic (RAFM) Steels
E-16: Stability and Diffusion of Lanthanide Fission Products in HCP Zirconium and BCC Iron Revealed by Density Functional Theory Calculations
E-17: Superionic-like Diffusion in Yttrium Hydride
E-18: Surface Chemistry and Microstructure of FeCrAl Alloys Under High Heating Rates Post-quenching
E-19: The Effects of Irradiation, Orientation, and Temperature on the Compressive Strength of Single-Crystal Zirconium via In-Situ TEM Micropillar Testing
E-1: Characterization of Alkylammonium Functionalized Smectite Organoclays from Molecular Dynamics Simulations
E-20: Tritium Population Near Dislocations in Zirconium from Molecular Dynamics
E-2: Characterization of Proton Irradiation-induced Nanoscale Precipitates in Model Low Alloy Steels Using Transmission Electron Microscopy and Nanoindentation
E-3: Coating Adherence Measurements Enabling Accelerated Screening of Accident Tolerant Claddings
E-4: Cr-Coated Zircaloy-4 Surface Chemistry and Microstructure Following High Temperature Excursions and Quenching
E-5: Development of Oxide Dispersion Strengthened Nickel-based Alloys for Enhanced Radiation Resistance
E-6: Fundamental Surface Reconstruction and Formation of Phyllosilicate Waste Barrier Materials
E-7: High-temperature Nano-indentation Response of Al0.3Ti0.2Co0.7CrFeNi1.7 High Entropy Alloy Processed Via Advanced Solid Phase Manufacturing Technique
E-8: High Temperature Compressive Creep Tests of Uranium Mononitride Using the Spark Plasma Sintering Apparatus
E-9: Mechanical Investigations on Diffusion Bonding for Compact Heat Exchangers Utilizing Digital Image Correlation (DIC) and Electron Back-Scattering Diffraction (EBSD)
Effective Parameterization of Phase-field Models of Fission Gas Bubble Growth
Effects of Heterogeneous Porosity on Buffer Layer Fracture Mode
Effects of Neutron Irradiation on the Fracture Behavior of PM-HIP and Cast Grade 91 Steel
Enhanced Properties of CrAl Coated ATF Cladding
Evaluation of Low-length Kr Diffusion in UO2 and ADOPTŪ using Time-of-flight Elastic Recoil Detection (ToF-ERDA)
Evolution of Microstructures in FeCr Binary Alloys Under Low-PKA Proton Irradiation
Experimental Methods for Accelerating Nuclear Structural Material Qualification
Fission Accelerated Steady-state Testing Experimental-BISON Comparison
Gaps and Question Remained in Bridging Neutron Irradiation and Accelerator Ion Irradiation
High-burnup Structure Formation and Associated Fission Product Diffusion in UO2
High Temperature Ring-pull Mechanical Tests of Thin-walled Tube
High Throughput Assessment of Creep Behavior of Advanced Nuclear Reactor Structural Alloys by Nanoindentation
Implications of Defect Induced Thermal Conductivity Degradation on Accelerated Irradiation of Nuclear Fuels
In-situ Characterization of FeCrAl Claddings Under Simulated LOCA Conditions
In Situ Ion Irradiation of a Spent UO2 Fuel
Integrated Experimental and Computational Qualification of Nuclear Structural Materials
Ion Beams: Unique Tools Contributing to Accelerated Qualification of Nuclear Materials
Ion Irradiation and Examination of Additive Friction Stir Manufactured 316 Stainless Steel Component
Irradiation-induced Helium Evolution and Damage Effects in REBCO Coated Conductors Used for Compact Fusion
Irradiation Effects on Mechanical Properties of PM-HIP Electron Beam Welded RPV Steels
Microstructure, Mechanical Properties, and Performance of Cold Spray Cr Coatings on Zr-alloy Fuel Cladding
Modeling of Fission Gas Behavior in Uranium Nitride Fuel
Multi-scale Simulations on Preferential Absorption Behavior of Cavities in BCC Fe
Nanoprecipitates to Enhance Radiation Tolerance in High-Entropy Alloys
Perspectives on the Pesky Problem of Post-Irradiation Hardening in Wrought FeCrAl Alloys
Phonon Dispersion, Lifetimes, and Thermal Transport in Nuclear Fuel Materials
Physics-based Model Prediction of Microstructure and Creep Properties for As-built Additively Manufactured Stainless Steel 316-H
Physics-informed Smart Scaling for Accelerated Fuel Testing
Plastic Deformation of Uranium Dioxide at High Temperature: Modeling of the Single Crystal Plastic Anisotropy
Primary Radiation Damage Evaluation on Thick Films Using a High-throughput Approach
Probing Anharmonicity Effects at Elevated Temperatures in Ceramic Nuclear Fuels and Surrogates using Raman Spectroscopy
Radiation-driven Diffusion of U, Si, and Xe in Amorphous U3Si2
Results of MARGARET Fission Gas and Microstructure Model, Following Latest Developments
Temperature Dependence of Helium Cavity Behavior in Ion-irradiated Ductile-phase-toughened Tungsten
The Role of Grain Boundaries in Irradiation Enhanced Creep: A Cluster Dynamics Study of UO2
Thermal Dependence of Mechanical Anisotropy In Zircaloy-4 Cladding
Three-dimensional Quantitative Defect Analysis in Tungsten Heavy Alloys Under the Simulated Nuclear Fusion Environment
Toward Qualification of PM-HIP RPV Steels and their Electron Beam Weldments
Two-step Upscaling for the Response of Ceramic Based Composites in Nuclear Reactors
Uncertainty Quantification and Bayesian Calibration Applied to Mechanistic Models of Nuclear Fuel Performance
Uncovering Grain Boundary Metastability as a Response to Radiation in FCC and BCC Single Phase Compositionally Complex Alloys
Understanding Nuclear Graphite’s Changes Upon Irradiation Across Microscopy to Mesoscopic Length Scales
Unfaulting of Dislocation Loops in Metals: Atomistic Simulations and Continuum Modeling
U(X)N-based SIMFUEL with X= Zr, Nb, Mo and Ru: Fabrication, Characterization and Phase Equilibria Evaluation


Questions about ProgramMaster? Contact programming@programmaster.org