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Meeting 2020 TMS Annual Meeting & Exhibition
Symposium Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
Organizer(s) Assel Aitkaliyeva, University of Florida
Peter Hosemann, University of California - Berkeley
Samuel Briggs, Oregon State University
David Frazer, Los Alamos National Laboratory
Scope The response of fuels and materials to radiation is critical to the performance of advanced nuclear systems. Key to understanding material performance in a nuclear environment is the analysis of materials irradiated using test reactors and ion beam facilities. This symposium will focus on recent results produced from irradiation programs around the world and will cover fundamental and applied science aspects of accelerated nuclear materials testing for fission and fusion reactors. Presentations combining experiment with theory, modeling and simulation to enhance our understanding of radiation-induced degradation in materials are especially encouraged.

Abstracts are solicited for (but not limited to) the following irradiation program topics:
- Fundamental science of radiation damage and defect processes
- Current and advanced nuclear fuels
- Current and advanced structural materials
- Fluence effects in materials
- In-situ testing of materials
Abstracts Due 07/15/2019
Proceedings Plan Undecided
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A comparative study of two nanoindentation approaches for assessing mechanical properties of ion-irradiated Stainless Steel 316
A comparison of ring pull and axial tensile tests of HT-9 and 14YWT thin-walled tubes
A Virtual Experiment Approach to Positron Annihilation Spectroscopy
ALPHA SELF-IRRADIATION OF ARCHIVE AND IRRADIATED FAST REACTOR FUELS
Analysis of the oxide nanoparticles trapping behavior in an ODS Eurofer steel by means of Positron Annihilation Spectroscopy
Behaviors of implanted Xe and Kr gas bubbles in CeO2 after annealing and rapid heating test
Changes in the Starting Microstructures of U-Mo Fuels due to the Effects of Neutron Irradiation
Characterization of Helium Implanted Single Crystal Titanium
Comparison of Radial Microstructural Changes in Fast Reactor MOX Fuels Across Varying Burnup Profiles
Controlling Helium Morphology in Pure Metals: Toward Uniform Samples for the Accelerated Measurement of Bulk Irradiated Properties
Damage Mitigation Strategies for Femtosecond Laser Machining of Micro-Tensile Bars
Defect evolution and radiation resistance of advanced fusion materials under heavy ion and low energy helium irradiation
Development of Advanced Low N-12Cr (wt.%) Ferritic/Martensitic Steel for Reactor Applications
Diffusion analysis of metallic fission products in tristructural-isotropic coated fuel using representative diffusion couples
Diffusion of Fission Products in Virgin Nuclear Graphite
Direct compaction of dispersion fuels using a matrix deposition on the fuel particles
Direct measurement of radiation damage through the energy stored in defects: simulations and experiments
Effective defect sinks in metallic composite with nanodispersoids: in situ ion radiation transmission electron microscopy and positron annihilation lifetime spectroscopy
Effects of Helium Ion Irradiation on Single Crystal Vanadium
Fabrication and characterization of massive crack-free delta phase-zirconium hydride for high-performance moderator application
Fabrication of Low-enriched Uranium Dispersion Targets with a High Uranium Density for Mo-99 Production
High-Throughput Synthesis and Ion Irradiation of High-Entropy Alloys using Additive Manufacturing
Impact of ionization effects and defect trapping on microstructure evolution in light ion irradiated uranium dioxide
In-situ neutron characterization of advanced nuclear fuels - the road to a new neutron irradiation testing capability
In-Situ Observation of Radiation-Induced Phase Transformation in U-Mo
In-situ Studies on the Mechanical Properties of He Ion Irradiated Nanotwinned Ag
In situ heavy ion irradiation of FCC and BCC high-entropy alloys at cryogenic and high temperatures
Interphase distribution behavior of oxide nanoparticles triggered by isothermal ferrite transformation in 9Cr ODS steels
Irradiation behavior of mechanically processed Zr-Nb multilayers at very high doses
Kinetic study on the evolution of nano-ceramic coatings under heavy ions irradiation
Linking Defect Structure and Property Evolution in Ion-Irradiated Tungsten: A Multi-Facetted View
Mechanical Properties of Ion Irradiated and Helium Implanted HT9 Micropillars
Microstructural and micro-chemical characterization of safety tested TRISO UCO Fuel Kernels Irradiated in the Advanced Test Reactor
Microstructural changes and corrosion of proton-pre-irradiated Hastelloy N in FLiNaK molten salt
Microstructure and Mechanical Behavior of Directed Energy Deposition Laser Additively Manufactured T-91
Microstructure of HT-9 cladding after fuel-cladding chemical interaction with an annular U-10Zr fuel irradiated to 3.3% FIMA
Multiple scale mechanical testing of neutron irradiated FeCrAl alloys
Neutron Irradiation Damage and Fission Product Transport in the SiC Layer of TRISO Fuel Particles
On a Theory Based Accelerated Testing Methodology for Swelling.
On the role of heterogeneity in concentrated solid‒solution alloys in enhancing their irradiation resistance
Promotion and suppression of the G-phase in steels
Radiation response of HT9 ferritic/martensitic alloys as a function of interstitial content
Rapid Investigation of Irradiation Temperature Sensitivity Using Charged Particles
Recent applications of ex situ transient grating spectroscopy to the study of radiation-induced degradation of nuclear materials
Small scale mechanical testing of ceramic interfaces in nuclear materials: Characterizing the impact of elastic mismatch on stress intensity and property extraction.
Synthesis of intermetallic UZr2+x and its phase transformation
Temperature shift evaluation for G-phase clustering in ferritic-martensitic alloys
Testing of Nuclear Fuels and Materials in the Advanced Fuels Campaign
Three-dimensional analysis of the IPyC/SiC interface in irradiated TRISO fuel particles


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