STP1529

    Zirconium in the Nuclear Industry: 16th International Symposium

    Limbäck Magnus, Barbéris Pierre
    Published: 2012


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    Forty-two peer-reviewed papers cover the entire spectrum of zirconium metallurgy, from basic metallurgy to accidental conditions and transport, through fabrication, creep and growth, and corrosion and hydriding.

    Sections cover:
    • Kroll Award Papers on explosion cladding; zirconium alloys in light water reactors; and much more.
    • Schemel Award Paper on photoelectrochemical investigation of radiation enhanced shadow corrosion phenomenon
    • Basic Metallurgy
    • Fabrication and Mechanical Properties
    • Hydriding—Hydrogen Effect
    • Corrosion—Oxide Layer Characterization
    • In Pile Behavior
    • Creep and Deformation
    • Failure Mechanisms and Transients


    Table of Contents

    Explosion Cladding: An Enabling Technology for Zirconium in the Chemical Process Industry

    Performance of Zirconium Alloys in Light Water Reactors with a Review of Nodular Corrosion

    The Evolution of Microstructure and Deformation Stability in Zr-Nb-(Sn,Fe) Alloys Under Neutron Irradiation

    The Development of Zr-2.5Nb Pressure Tubes for CANDU Reactors

    Photoelectrochemical Investigation of Radiation-Enhanced Shadow Corrosion Phenomenon

    Dynamic Recrystallization in Zirconium Alloys

    Measurement and Modeling of Second Phase Precipitation Kinetics in Zirconium Niobium Alloys

    Texture Evolution of Zircaloy-2 During Beta-Quenching: Effect of Process Variables

    In Situ Studies of Variant Selection During the α—β—α Phase Transformation in Zr-2.5Nb

    Segregation in Vacuum Arc Remelted Zirconium Alloy Ingots

    Damage Build-Up in Zirconium Alloys During Mechanical Processing: Application to Cold Pilgering

    Multiscale Analysis of Viscoplastic Behavior of Recrystallized Zircaloy-4 at 400°C

    Polycrystalline Modeling of the Effect of Texture and Dislocation Microstructure on Anisotropic Thermal Creep of Pressurized Zr-2.5Nb Tubes

    Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors

    Microstructural Studies of Heat Treated Zr-2.5Nb Alloy for Pressure Tube Applications

    High Temperature Aqueous Corrosion and Deuterium Uptake of Coupons Prepared from the Front and Back Ends of Zr-2.5Nb Pressure Tubes

    Hydrogen Absorption Mechanism of Zirconium Alloys Based on Characterization of Oxide Layer

    In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes

    Study on the Role of Second Phase Particles in Hydrogen Uptake Behavior of Zirconium Alloys

    Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation Diffraction

    Statistical Analysis of Hydride Reorientation Properties in Irradiated Zircaloy-2

    The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program

    Neutron Radiography: A Powerful Tool for Fast, Quantitative and Non-Destructive Determination of Hydrogen Concentration and Distribution in Zirconium Alloys

    Detailed Analysis of the Microstructure of the Metal/Oxide Interface Region in Zircaloy-2 after Autoclave Corrosion Testing

    Study of the Initial Stage and Anisotropic Growth of Oxide Layers Formed on Zircaloy-4

    Studies Regarding Corrosion Mechanisms in Zirconium Alloys

    Understanding Crack Formation at the Metal/Oxide Interface During Corrosion of Zircaloy-4 Using a Simple Mechanical Model

    16th International Symposium on Zirconium in the Nuclear Industry, 9th May 2010 – 13th May 2010, Chengdu, Sichuan Province, China


    Optimization of Zry-2 for High Burnups

    Effects of Secondary Phase Particle Dissolution on the In-Reactor Performance of BWR Cladding

    Hydrogen Solubility and Microstructural Changes in Zircaloy-4 Due to Neutron Irradiation

    Advanced Zirconium Alloy for PWR Application

    Ultra Low Tin Quaternary Alloys PWR Performance—Impact of Tin Content on Corrosion Resistance, Irradiation Growth, and Mechanical Properties

    Radiation Damage of E635 Alloy Under High Dose Irradiation in the VVER-1000 and BOR-60 Reactors

    ZIRLO® Irradiation Creep Stress Dependence in Compression and Tension

    Experimental Investigation of Irradiation Creep and Growth of Recrystallized Zircaloy-4 Guide Tubes Pre-Irradiated in PWR

    REFLET Experiment in OSIRIS: Relaxation under Flux as a Method for Determining Creep Behavior of Zircaloy Assembly Components

    Impact of Irradiation Damage Recovery During Transportation on the Subsequent Room Temperature Tensile Behavior of Irradiated Zirconium Alloys

    Shadow Corrosion-Induced Bow of Zircaloy-2 Channels

    Characterization of Oxygen Distribution in LOCA Situations

    Effect of Hydrides on Mechanical Properties and Failure Morphology of BWR Fuel Cladding at Very High Strain Rate

    Simulation of Outside-in Cracking in Boiling Water Reactor Fuel Cladding Tubes under Power Ramp

    RIA Failure of High Burnup Fuel Rod Irradiated in the Leibstadt Reactor: Out-of-Pile Mechanical Simulation and Comparison with Pulse Reactor Tests

    Subject Index

    Author and Subject Index


    Committee: B10

    DOI: 10.1520/STP1529-EB

    ISBN-EB: 978-0-8031-8893-8

    ISBN-13: 978-0-8031-7515-0

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    978-0-8031-7515-0
    STP1529-EB