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Meeting 2014 TMS Annual Meeting & Exhibition
Symposium Radiation Effects in Oxide Ceramics and Novel LWR Fuels
Sponsorship TMS Structural Materials Division
TMS/ASM: Nuclear Materials Committee
Organizer(s) Xian-Ming Bai, Virginia Tech
Todd R Allen, Idaho National Laboratory
Blas P Uberuaga, Los Alamos National Laboratory
Jianliang Lin, Colorado School of Mines
Michele Manuel, University of Florida
Dragos Staicu, European Commission, Joint Research Centre, Institute for Transuranium Elements
Yong Yang, University of Florida
Scope Oxide ceramics play a critical role in nuclear energy applications. Nuclear fuels, such as uranium dioxide (UO2) and Mixed Oxide (MOX) fuels, have been used in current light water reactors (LWRs) to produce about 15% of the electricity in the world. Other oxide ceramics have been proposed for novel inert matrix fuels or have been extensively studied as waste forms for the immobilization of nuclear waste. This symposium focuses on understanding the microstructural evolution of oxide ceramics in irradiation environments, with a particular focus on the effects of that evolution on thermal transport. The interested topics include but are not limited to: defect production and evolution, fission products dissolution in the lattice or precipitation, gas bubble formation, nucleation and accumulation and the associated volumetric swelling, changes in thermal transport induced by microstructural evolution, and radiation-induced phase changes. Experimental studies using various irradiation techniques such as laboratory accelerators, test reactors, and commercial nuclear power reactors are all of interest. The important fuel pellet-cladding interactions will also be discussed. Contributions focused on novel fuels such as doped UO2 with large grain sizes, high density uranium fuels like uranium nitrides and silicides, and accident-tolerant fuels are also encouraged. This symposium is intended to bring together national laboratory, university, and nuclear industry researchers from around the world to discuss the current understanding of the radiation response of oxide ceramics for nuclear applications through experiment, theory and computational multi-scale modeling.

Four sessions/topics are proposed:

Session 1: Experimental characterization of non-irradiated and irradiated oxide ceramics
Session 2: Multi-scale modeling and theoretical work on microstructure evolution under irradiation in oxide ceramics
Session 3: Effects of radiation on thermal and mechanical properties of ceramic oxide fuels
Session 4: Fuel Pellet-Cladding Interactions and novel LWR fuels
Abstracts Due 07/15/2013
Proceedings Plan Planned: Other (describe below)

Calculations of Threshold Displacement Energies in Y2Ti2O7 and Y2TiO5
Characterization of MOX Fuel Pellets by Photothermal Microscopy
Correlation between Thermal Conductivity and Microstructural Evolutions in CeO2 upon Radiation and Fission Gas Implantation
Damage Structure Evolution in Ion Irradiated UO2
Doping d-UO2 Fuel Pellets for Improved Hardness and Fracture Toughness at High Temperatures
Electrochemical Effect of Void Ensembles in UO2
Impact of Nano-pores on the Fuel Thermal Conductivity
Ion-implantation Induced Nano-channels for Optical Waveguide in LiNbO3
Ion Irradiation-induced Structural Transitions in Orthorhombic Ln2TiO5
Irradiation-induced Grain Growth in Nanocrystalline Ceria
Irradiation Behavior of High-burnup LWR-MOX (Mixed-Oxide) Fuels
Irradiation Response of Fluorite-structured Oxides to Extreme Irradiation Conditions
Kr and Xe Bubble Characterization in CeO2
Mechanical Behavior of UO2 under Irradiation: A Molecular Dynamics Study
Misorientation-dependence of Grain Boundary Thermal Resistance in CeO2
Modeling of Grain Growth in UO2 under a Temperature Gradient
Molecular Dynamics Simulations of the Effect of Point Defects and Embedded Xe Atoms on Thermal Transport in Uranium Oxide
Molecular Dynamics Study of Grain Boundary Properties in UO2
Multiscale Computer Simulation of Fission Gas Release in Oxide Fuels
Nano-scale Irradiation Induced Chemistry Changes in Oxide Fuel Materials
Radiation Effects in UO2
Radiation Resistance of Nickel, Iron, Chromium Spinels by MD Simulations
Reflections on Fuel Pellet-cladding Interaction (PCI)
Segregation of Fission Products to Dislocations in Uranium Dioxide
Structural Defects in Uranium Dioxide: From Oxidation to Irradiation
The Role of Non-stoichiometry in the Radiation Damage Evolution of SrTiO3
Thermal Conductivity, Microstructure and Gas Release from a 44 GWd/t MOX Fuel
Understanding Irradiation Induced Changes in Structure and Thermal Properties of UO2 Grain Boundaries
Understanding Nuclear Fuel Thermal Conductivity from Phonons in UO2
UO2 Fission Gas Release Rates from Atomistic Calculations of Intrinsic and Radiation-enhanced Diffusion Coefficients

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