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    STP1354

    Zirconium in the Nuclear Industry: Twelfth International Symposium

    Sabol GP, Moan GD
    Published: 2000


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    44 papers discuss the history, processing, use, and properties of Zirconium and its alloys. Nearly half are concerned with the corrosion and hydriding of different Zirconium alloys. The techniques used to characterize the oxides and their microstructures and properties have become very advanced. The industry, including producers and users, expects that the work will lead to important conclusions about alloy development, optimization of the processing parameters, and station operating procedures so that the current and new Zirconium materials will continue to operate reliably to very high neutron fluences.

    Seven major topics include:

    • Properties at High Neutron Fluences

    • Hydrogen and Temperature Effects

    • Deformation and Fracture Studies

    • Processing and Alloy Development

    • Effect of Composition and Microstructure on Corrosion

    • Corrosion Simulation and the Effect of the Environment

    • Effects of Oxides on Oxidation and H Pickup Rates

    Audience:Scientists and engineers involved in the production and properties of components for use in nuclear reactors, their behavior during service and the property changes that occur with increasing neutron fluences.


    Table of Contents

    Microstructure and Properties of Zirconium Alloys in the Absence of Irradiation

    Effects of Neutron Irradiation on Microstructure and Properties of Zircaloy

    Understanding Hydrogen in Zirconium

    Irradiation Creep and Growth Behavior, and Microstructural Evolution of Advanced Zr-Base Alloys

    Irradiation-Enhanced Deformation of Zr-2.5Nb Tubes at High Neutron Fluences

    High-Fluence Irradiation Growth of Cold-Worked Zr-2.5Nb

    Evolution of Dislocation and Precipitate Structure in Zr Alloys Under Long-Term Irradiation

    Effect of Long-Term Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes

    Post-Irradiation Characterization of Ultra-High-Fluence Zircaloy-2 Plate

    The Terminal Solid Solubility of Hydrogen in Zirconium Alloys

    The Long-Range Migration of Hydrogen Through Zircaloy in Response to Tensile and Compressive Stress Gradients

    Influence of Cladding Microstructure on the Low Enthalpy Failures in RIA Simulation Tests

    Experiment and Modeling of Advanced Fuel Rod Cladding Behavior Under LOCA Conditions: Alpha-Beta Phase Transformation Kinetics and EDGAR Methodology

    Behavior of Irradiated Zircaloy-4 Fuel Cladding Under Simulated LOCA Conditions

    Influence of Texture on the Fracture Toughness of Unirradiated Zircaloy Cladding

    Degraded Fuel Cladding Fractography and Fracture Behavior

    Delayed Hydride Cracking in Irradiated Zircaloy Cladding

    Size, Geometry, and Material Effects in Fracture Toughness Testing of Irradiated Zr-2.5Nb Pressure Tube Material

    Failure Mechanisms of Irradiated Zr Alloys Related to PCI: Activated Slip Systems, Localized Strains, and Iodine-Induced Stress Corrosion Cracking

    Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes

    A New Fabrication Process for Zr-Lined Zircaloy-2 Tubing

    Zirconium Alloy Cold Pilgering Process Control by Modeling

    In-Process Investigation of Precipitate Growth in Zirconium Alloys

    Influence of Composition and Fabrication Process on Out-of-Pile and In-Pile Properties of M5 Alloy

    Effect of Dilute Alloy Additions of Molybdenum, Niobium, and Vanadium on Zirconium Corrosion

    E110 Alloy Cladding Tube Properties and Their Interrelation with Alloy Structure-Phase Condition and Impurity Content

    Contribution to a Better Understanding of the Detrimental Role of Hydrogen on the Corrosion Rate of Zircaloy-4 Cladding Materials

    Mechanism of Corrosion Rate Degradation Due to Tin

    Influence of Transition Elements Fe, Cr, and V on Long-Time Corrosion in PWRs

    Fundamental Study on the Corrosion Mechanism of Zr-0.2Fe, Zr-0.2Cr, and Zr-0.1Fe-0.1Cr Alloys

    Microstructural Aspects of Corrosion and Hydrogen Ingress in Zr-2.5Nb

    The Effect of Microstructure on the Corrosion Behavior of Zircaloy-2 in BWRs

    Impact of Second Phase Particles on BWR Zr-2 Corrosion and Hydriding Performance

    Corrosion of Electron-Irradiated Zr-2.5Nb and Zircaloy-2

    Evaluation of Zircaloy-2 Cladding Corrosion Characteristics by Simulated BWR Corrosion Loop Test

    Studies of Zirconium Alloy Corrosion and Hydrogen Uptake During Irradiation

    Behavior of Lithium and Boron in Irradiated and Unirradiated Oxides Formed on Zircaloy-4 Claddings

    Contribution to the Understanding of the Effect of the Water Chemistry on the Oxidation Kinetics of Zircaloy-4 Cladding

    Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance

    Electrochemical Examinations in 350°C Water with Respect to the Mechanism of Corrosion-Hydrogen Pickup

    Hydrogen Transport in the Oxide and Hydrogen Pickup by the Metal During Out- and In-Reactor Corrosion of Zr-2.5Nb Pressure Tube Material

    Stress Distribution Measured by Raman Spectroscopy in Zirconia Films Formed by Oxidation of Zr-Based Alloys

    Role of Intermetallic Precipitates in Hydrogen Transport through Oxide Films on Zircaloy

    Multi-Scale Characterization of the Metal-Oxide Interface of Zirconium Alloys

    Author Index

    Subject Index


    Committee: B10

    DOI: 10.1520/STP1354-EB

    ISBN-EB: 978-0-8031-5416-2

    ISBN-13: 978-0-8031-2499-8

    ASTM International is a member of CrossRef.

    0-8031-2499-6
    978-0-8031-2499-8
    STP1354-EB