Zirconium in the Nuclear Industry: Fourteenth International Symposium

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Digital Library/Symposia/B10/Zirconium in the Nuclear Industry: Fourteenth International Symposium

Sponsored by ASTM committee B10 on Reactive and Refractory Metals and Alloys
January 13 2004 - January 15 2004


	Papers Published from this Symposium

Role of Iron for Hydrogen Absorption Mechanism in Zirconium Alloys

The Effect of Duplex Cladding Outer Component Tin Content on Corrosion, Hydrogen Pick-up, and Hydride Distribution at Very High Burnup

Identification of Crystalline Behavior on Macroscopic Response and Local Strain Field Analysis: Application to Alpha Zirconium Alloys

Effect of Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes at the End of Design Life

The Effect of Liner Component Iron Content on Cladding Corrosion, Hydriding, and PCI Resistance

Destruction of Crystallographic Texture in Zirconium Alloy Tubes

Plastic Deformation of Irradiated Zirconium Alloys: TEM Investigations and Micro-Mechanical Modeling

Influence of StructurePhase State of Nb Containing Zr Alloys on Irradiation-Induced Growth

Delayed Hydrogen Cracking Velocity and J-Integral Measurements on Irradiated BWR Cladding

Fretting-Wear Behavior of Zircaloy-4, OPTIN, and ZIRLO Fuel Rods and Grid Supports Under Various Autoclave and Hydraulic Loop Endurance Test Conditions

Effect of Fabrication Variables on Irradiation Response of Crack Growth Resistance of Zr-2.5Nb

Microstructural Stability of M5 Alloy Irradiated up to High Neutron Fluences

Temperature and Strain Rate Effects on Zr-1%Nb Alloy Deformation

Damage Dependence of Irradiation Deformation of Zr-2.5Nb Pressure Tubes

The Effect of Beta-Quenching in Final Dimension on the Irradiation Growth of Tubes and Channels

Improved ZIRLO Cladding Performance through Chemistry and Process Modifications

Shadow Corrosion Mechanism of Zircaloy

TEM Examinations of the Metal-Oxide Interface of Zirconium Based Alloys Irradiated in a Pressurized Water Reactor

Failure of Hydrided Zircaloy-4 Under Equal-Biaxial and Plane-Strain Tensile Deformation

Mechanical Properties of Zircaloy-4 PWR Fuel Cladding with Burnup 54-64MWd/kgU and Implications for RIA Behavior

On Secondary -Nb Phase Precipitation within Primary -Zr Phase in Zr-Nb Alloys During Tensile Deformation

Study of Nb and Fe Precipitation in -Phase Temperature Range (400 to 550°C) in Zr-Nb-(Fe-Sn) Alloys

Predicting Oxidation and Deuterium Ingress for Zr-2.5Nb CANDU Pressure Tubes

In-Core Tests of Effects of BWR Water Chemistry Impurities on Zircaloy Corrosion

Microstructure and Growth Mechanism of Oxide Layers Formed on Zr Alloys Studied with Micro-Beam Synchrotron Radiation

Ductility of Zircaloy-4 Fuel Cladding and Guide Tubes at High Fluences

Microstructure and Phase Control in Zr-Fe-Cr-Ni Alloys: Thermodynamic and Kinetic Aspects

Inhibitors for Reducing Hydrogen Ingress During Corrosion of Zirconium Alloys

Overload Fracture of Flaw Tip Hydrides in Zr-2.5Nb Pressure Tubes

In-Situ Studies of the Oxide Film Properties on BWR Fuel Cladding Materials

Review of Deformation Mechanisms, Texture, and Mechanical Anisotropy in Zirconium and Zirconium Base Alloys

Simulation of Cold Pilgering Process by a Generalized Plane Strain FEM

Use of the Irradiation-Thermal Creep Model of Zr-1% Nb Alloy Cladding Tubes to Describe Dimensional Changes of VVER Fuel Rods

Influence of Long Service Exposures on the Thermal-Mechanical Behavior of Zy-4 and M5 Alloys in LOCA Conditions

ZIRLO An Alloy Development Success

The Correlation Between Microstructures and in-BWR Corrosion Behavior of Highly Irradiated Zr-based Alloys

Modeling of the Simultaneous Evolution of Vacancy and Interstitial Dislocation Loops in hcp Metals Under Irradiation

Phase Composition, Structure, and Plastic Deformation Localization in Zr1%Nb alloys

Effect of Alloying Elements and Impurities on in-BWR Corrosion of Zirconium Alloys

Comparison of the High Burn-Up Corrosion on M5 and Low Tin Zircaloy-4

Thermal Creep of Irradiated Zircaloy Cladding

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