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About this Symposium
Meeting 2010 TMS Annual Meeting & Exhibition
Symposium Mechanical Performance for Current and Next-Generation Nuclear Reactors
Sponsorship The Minerals, Metals and Materials Society
TMS Materials Processing and Manufacturing Division
TMS Structural Materials Division
TMS/ASM: Mechanical Behavior of Materials Committee
TMS: Nanomechanical Materials Behavior Committee
TMS/ASM: Nuclear Materials Committee
Organizer(s) Dylan Morris, NIST
Greg Oberson, Nuclear Regulatory Commission
Nicholas Barbosa, National Institute of Standards & Tech
Wolfgang Hoffelner, Paul Scherrer Institute
Scope This symposium will provide a forum for researchers engaged in experimental or theoretical investigations of the mechanical behavior of materials in nuclear reactors. In particular, this symposium is intended to accelerate the development and acceptance of materials and materials measurement techniques that may allow extended operating lives for current nuclear power plants, and facilitate increases in safety, reliability, and efficiency in the next generation of nuclear power plants.

Experimental, theoretical and computational studies are sought in the areas of (but not limited to):
Advances in non-destructive flaw detection, residual stress measurement, and corrosion monitoring
Hybrid and novel mechanical testing techniques, such as in-situtesting in reactors or simulated aggressive environments
High-throughput and small-scale mechanical testing techniques
Accelerated testing, e.g. to investigate late-blooming phases or hydrogen embrittlement
Micromechanical investigations of materials subjected to reactor-specific harsh environments such as very-high-temperature fluids or fast neutrons

Our confirmed Invited Speakers are:

Gene Carpenter, Nuclear Regulatory Commission
Jeffrey Henry, Energy Solutions Group
Stuart Maloy, Los Alamos National Laboratory
Randy Nanstad, Oak Ridge National Laboratory
Robert Odette, University of California – Santa Barbara
Mikhail Sokolov, Oak Ridge National Laboratory
Ralph Spolenak, ETH-Zurich
Makuteswara Srinivasan, Nuclear Regulatory Commission
Brian Wirth, University of California - Berkeley
Abstracts Due 11/15/2009
Proceedings Plan Definite: Publication Outside of TMS

Atomic-Scale Modeling of the Dislocation - Radiation Obstacle Interactions Responsible for Mechanical Property Changes in Irradiated Metals
Comparative Plant Performance of Stabilized and Non-Stabilized Austenitic Stainless Steels
Damage Related Information Contained in Small Material Volumes of Advanced Nuclear Plants
Diffusion of Silver and Gold in Ion Irradiated Glassy Polymeric Carbon
Effect of Neutron Radiation Exposure on Low Cycle Fatigue of 304SS
Elevated-Temperature Compression Testing and Characterization of Deformation and Fracture in Sintered Zrn Pellets as Surrogates for Pun Fuels
Ensuring the Performance of Nuclear Reactor Pressure Vessels for Long Time Service
Evolution of the Thermo-mechanical Response of Nitride and Oxide Nuclear Fuels through Microstructurally Explicit Models.
Experience of the Fossil Industry with the Creep-Strength Enhanced Ferritic Steels
Experimental Analysis and Computational Modeling of Temperature Dependent Cyclic Plastic Hardening and Strain Controlled Ratcheting
First Principles Study of Defects in Uranium.
Gen IV Materials (ASME-DOE Project)
High-Temperature Corrosion of YSZ Plasma-Sprayed on Nickel-Alloys in Molten Chloride Salts
High Temperature Oxidation Behavior of Grain-refined Inconel 617 for VHTRs
Hot Steam Corrosion Behavior of Ni-based Superalloys at High Temperature Steam Environments.
Influence of Hydrostatic Stress on Primary Defect Generation during Displacement Cascade in α-Fe
Intergranular Thermal Residual Strain in Rolled and Texture-free α-Uranium
Irradiation Effects in Thin Metal Films – Texture Control and Mechanical Properties
Late-Blooming Phase Investigation in an Ion Irradiated Fe-1wt.%Mn Alloy
Materials Issues Potentially Impacting Long-Term Safe Operations
Mechanical Properties of Fresh and Irradiated Monolithic U-Mo Fuels
Mechanical Testing of Core Fast Reactor Materials for the Advanced Fuel Cycle Initiative
Micro and Macro Scale Mechanical Testing and Characterization on Irradiated Structural Materials for Nuclear Application
Microscale Methods for Evaluating Mechanical Behavior of Ion Irradiated Metals at High Damage Levels
Microstructural and Mechanical Characteristics of Friction Stir Welded ODS Alloys
Modeling the Effect of Stress on Defect Migration and Void Formation using the Phase Field Method
Modelling Steels used in Nuclear Energy Applications
Multiscale Modeling of amorphous-Fe and Fe-Ni systems used in extreme environments such as nuclear reactors
Phase-Field Simulation of Void and Fission-Gas Bubble Evolution in Irradiated Polycrystalline Materials
Safety Evaluation Challenges for NGNP VHTR Materials of Construction and Components
Small Specimen and in situ Mechanical Test Methods in the US Fusion Reactor Materials Program
Small Specimen Test Techniques for Evaluating Properties of Irradiated Materials
Structural Modifications and Mechanical Degradation of Ion Irradiated Glassy Polymer Carbon
Studying the Effect of Carbon on DU-Mo Foil Fabrication using Small-Scale Specimen Testing
The Cause of Dynamic Strain Aging in Zirconium Alloys
The Effects of Stress and Temperature on the Fatigue Crack Growth Behavior and Microstructural Evolution of Alloy 230
The Thermal Stability and Weldability of a Lean Grade of Duplex Stainless Steel
Universal Scaling of Work Hardening Parameters in Type 316L(N)
When the Turtle Can’t Get There and The Rabbit Gets Lost: Predicting Low Flux High Fluence RPV Embrittlement

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