STP1245

    Zirconium in the Nuclear Industry: Tenth International Symposium

    Garde AM, Bradley ER

    Pages: 796

    Published: 1994

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    4 sections are devoted to various aspects of zirconium alloy corrosion: Hydrogen Effects; • Lithium Effects and Second-Phase Particles; • Oxide Characterization; • and In-Reactor Corrosion.

    The remaining sections include Pressure Tubes; Fabrication; and Mechanical Properties.


    Table of Contents

    Development of Zirconium-Barrier Fuel Cladding
    Armijo J., Coffin L., Rosenbaum H.

    Zirconium Alloy Performance in Light Water Reactors: A Review of UK and Scandinavian Experience
    Pickman D.

    The Effect of Fluence and Irradiation Temperature on Delayed Hydride Cracking in Zr-2.5Nb
    Coleman C., Griffiths M., Sagat S., Wilkins B.

    A Study of the Hydrogen Uptake Mechanism in Zirconium Alloys
    Elmoselhi M., McIntyre S., Warr B.

    Hydrogen Absorption Kinetics During Zircaloy Oxidation in Steam
    Barberis P., Charquet D., Mikes-Lindback M., Rudling P.

    Scanning Electron Microscope Techniques for Studying Zircaloy Corrosion and Hydriding
    Pearce J., Schrire D.

    Growth and Characterization of Oxide Films on Zirconium-Niobium Alloys
    Chan P., Khatamian D., Urbanic V., Woo O.

    Correlation Between Irradiated and Unirradiated Fracture Toughness of Zr-2.5Nb Pressure Tubes
    Chow C., Davies P., Griffiths M., Hosbons R.

    Variability of Irradiation Growth in Zr-2.5Nb Pressure Tubes
    Causey A., Elder J., Fleck R., Holt R.

    Change of Mechanical Properties by Irradiation and Evaluation of the Heat-Treated Zr-2.5Nb Pressure Tube
    Akiyama T., Koike M., Nagamatsu K., Shibahara I.

    On the Anisotropy of In-Reactor Creep of Zr-2.5Nb Tubes
    Causey A., Elder J., Fleck R., Holt R.

    Fabrication of Zr-2.5Nb Pressure Tubes to Minimize the Harmful Effects of Trace Elements
    Aldridge S., Choubey R., Coleman C., Davis L., Graham R., Moan G., Theaker J.

    Modeling of Damage in Cold Pilgering
    Aubin J., Girard E., Montmitonnet P.

    Mitigation of Harmful Effects of Welds in Zirconium Alloy Components
    Bowden J., Coleman C., Doubt G., Fong R., Root J., Sagat S., Webster R.

    Fabrication Process of High Nodular Corrosion-Resistant Zircaloy-2 Tubing
    Abe H., Furugen M., Hama T., Konishi T., Matsuda K.

    Corrosion Behavior of Zircaloy-4 Sheets Produced Under Various Hot-Rolling and Annealing Conditions
    Anada H., Nomoto K., Shida Y.

    Optimization of PWR Behavior of Stress-Relieved Zircaloy-4 Cladding Tubes by Improving the Manufacturing and Inspection Process
    Charquet D., Mardon J., Senevat J.

    Experimental and Theoretical Studies of Parameters that Influence Corrosion of Zircaloy-4
    Amanrich H., Billot P., Giordano A., Peybernès J., Robin J., Thomazet J.

    Aqueous Chemistry of Lithium Hydroxide and Boric Acid and Corrosion of Zircaloy-4 and Zr-2.5Nb Alloys
    Balakrishnan P., Ramasubramanian N.

    Corrosion Behavior of Irradiated Zircaloy
    Adamson R., Cheng B., Kruger R.

    Correlation of Transmission Electron Microscopy (TEM) Microstructure Analysis and Texture with Nodular Corrosion Behavior for Zircaloy-2
    Herb B., McCarthy J., Ruhmann H., Wang C.

    Micro-Characterization of Corrosion Resistant Zirconium-Based Alloys
    Isobe T., Mae Y., Matsuo Y.

    An Experimental Investigation into the Oxidation of Zircaloy-4 at Elevated Pressures in the 750 to 1000°C Temperature Range
    Bramwell I., Haste T., Parsons P., Worswick D.

    Prediction of Creep Anisotropy in Zircaloy Cladding
    Perkins R., Shann S.

    Microstructure and Properties of Corrosion-Resistant Zirconium-Tin-lron-Chromium-Nickel Alloys
    Bradley E., Nyström A.

    Fatigue Behavior of Irradiated and Unirradiated Zircaloy and Zirconium
    Adamson R., Reynolds M., Wisner S.

    Effect of Irradiation on the Microstructure of Zircaloy-4
    Gilbon D., Simonot C.

    Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes
    Lansiart S., Mardon J., Royer J., Soniak A., Waeckel N.

    Grain-by-Grain Study of the Mechanisms of Crack Propagation During Iodine Stress Corrosion Cracking of Zircaloy-4
    Dorado A., Haddad R.

    Microstructure of Oxide Layers Formed During Autoclave Testing of Zirconium Alloys
    Andrén H., Lai Z., Nyström A., Pettersson H., Rudling P., Wadman B.

    Corrosion Performance of New Zircaloy-2-Based Alloys
    Andrén H., Lethinen B., Mikes-Lindbäck M., Rudling P., Stiller K.

    Examinations of the Corrosion Mechanism of Zirconium Alloys
    Beie H., Garzarolli F., Mitwalsky A., Ruhmann H., Sell H.

    On the Initial Corrosion Mechanism of Zirconium Alloy: Interaction of Oxygen and Water with Zircaloy at Room Temperature and 450°C Evaluated by X-Ray Absorption Spectroscopy and Photoelectron Spectroscopy
    Beie H., Döbler U., Knop A., Ruhmann H.

    How the Tetragonal Zirconia is Stabilized in the Oxide Scale that is Formed on a Zirconium Alloy Corroded at 400°C in Steam
    Godlewski J.

    Oxidation of Intermetallic Precipitates in Zircaloy-4: Impact of Irradiation
    Charquet D., Lefebvre F., Lemaignan C., Motta A., Pêcheur D.

    Corrosion Optimized Zircaloy for Boiling Water Reactor (BWR) Fuel Elements
    Garzarolli F., Schumann R., Steinberg E.

    In-Reactor Corrosion Performance of ZIRLO™ and Zircaloy-4
    Comstock R., Larouere P., Sabol G., Stanutz R., Weiner R.

    Phenomenological Study of In-Reactor Corrosion of Zircaloy-4 in Pressurized Water Reactors
    Kim Y., Min D., Rheem K.

    Corrosion Behavior of Zircaloy-4 Cladding with Varying Tin Content in High-Temperature Pressurized Water Reactors
    Endter R., Garde A., Krammen M., Pati S., Smith G.

    Effects of Pressurized Water Reactor (PWR) Coolant Chemistry on Zircaloy Corrosion Behavior
    Karlsen T., Vitanza C.

    Author Index

    Subject Index


    Committee: B10

    Paper ID: STP1245-EB

    DOI: 10.1520/STP1245-EB

    ISBN-EB: 978-0-8031-5286-1

    ISBN-PRINT: 978-0-8031-2011-2

    ASTM International is a member of CrossRef.

    0-8031-2011-7
    978-0-8031-2011-2
    STP1245-EB