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    Zirconium in the Nuclear Industry: Fourteenth International Symposium

    Rudling P, Kammenzind B
    Published: 2005

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    Forty-one peer-reviewed papers cover all aspects of the fuel cycle. The Basic Metallurgy and Processing sections represent the beginning of the fuel cycle from component design and manufacture. The BWR Corrosion, PWR Corrosion, and Deformation Mechanism sections represent component performance under nominal service conditions. The Primary Failure Mechanism and the Transient and High Temperature Performance sections cover aspects that might end the useful service life of the component. The remaining papers deal with post service life of the component in a dry storage and in a repository environment.

    Table of Contents

    ZIRLO™ — An Alloy Development Success

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

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

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

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

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

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

    Microstructural Stability of M5™ Alloy Irradiated up to High Neutron Fluences

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

    Simulation of Cold Pilgering Process by a Generalized Plane Strain FEM

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

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

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

    Destruction of Crystallographic Texture in Zirconium Alloy Tubes

    Improved ZIRLO™ Cladding Performance through Chemistry and Process Modifications

    Role of Iron for Hydrogen Absorption Mechanism in Zirconium Alloys

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

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

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

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

    Shadow Corrosion Mechanism of Zircaloy

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

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

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

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

    Inhibitors for Reducing Hydrogen Ingress During Corrosion of Zirconium Alloys

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

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

    Thermal Creep of Irradiated Zircaloy Cladding

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

    Influence of Structure—Phase State of Nb Containing Zr Alloys on Irradiation-Induced Growth

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

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

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

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

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

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

    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

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

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

    Author Index

    Subject Index

    Author Index

    Subject Index

    Committee: B10

    DOI: 10.1520/STP1467-EB

    ISBN-EB: 978-0-8031-5519-0

    ISBN-13: 978-0-8031-3493-5