STP633

    Zirconium in the Nuclear Industry

    Lowe AL, Parry GW
    Published: 1977


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    45 papers present vitally important data on the behavior of zirconium alloys in nuclear reactor environments.


    Table of Contents

    Introduction
    Lowe A., Parry G.

    Summary
    Lowe A., Parry G.

    Zircaloy—Three Years after the Hearings
    Johnston W.

    Diameter Increases in Steam Generating Heavy Water Reactor Zircaloy Cans under Loss-of-Coolant Accident Conditions
    Hindle E., Rose K.

    Rupture Characteristics of Zircaloy-4 Cladding with Internal and External Simulation of Reactor Heating
    Barber A., Fiveland W., Lowe A.

    High Temperature Strain Behavior of Zircaloy-4 and Zr-2.5Nb Fuel Sheaths
    Foote D., Hunt C.

    Superplasticity of Zircaloy-4
    Boček M., Hofmann P., Petersen C.

    Deformation and Rupture Behavior of Zircaloy Cladding under Simulated Loss-of-Coolant Accident Conditions
    Chung H., Garde A., Kassner T.

    Influence of Metallurgical Variables on the Performance of Zircaloy Fuel Sheathing
    Bain A., Hardy D.

    Diffusion of Oxygen in Beta-Zircaloy and the High Temperature Zircaloy-Steam Reaction
    Cathcart J., Druschel R., McKee R., Pawel R., Perkins R., Yurek G.

    Oxidation of Zirconium During a High-Temperature Transient
    Jones S., Ledoux G., Sawatzky A.

    Zircaloy-4 Oxidation in Steam Under Transient Oxidizing Conditions
    Ballinger R., Biederman R., Dobson W.

    Oxidation of Zirconium Alloys in Steam at 1000 to 1850°C
    Urbanic V.

    Zircaloy Cladding Behavior During Irradiation Tests Under Power-Cooling-Mismatch Conditions
    Gibson G., Hobbins R., Smolik G.

    Corrosion of Zirconium-Base Alloys—An Overview
    Hillner E.

    Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments
    Garlick A., Sumerling R., Trowse F.

    Corrosion Monitoring of Steam Generating Heavy Water Reactor Pressure Tubes
    Sheppard M., Tyzack C.

    Embrittlement of Zircaloy-4 by Liquid Cesium at 300°C
    Cubicciotti D., Jones R., Syrett B.

    Nodular Corrosion of the Zircaloys
    Horton R., Johnson A.

    Use of State Variables in the Description of Irradiation Creep and Deformation of Metals
    Hart E., Li C.

    Irradiation Growth of Zircaloy
    Adamson R.

    Zircaloy-2 Pressure Tube Elongation at the Hanford N Reactor
    Alexander W., Fidleris V., Holt R.

    Numerical Model for the Anisotropic Creep of Zircaloy
    Franklin D., Franz W.

    Modeling of Localized Deformation in Neutron Irradiated Zircaloy-2
    Adamson R., Lee D.

    In-Reactor Creep of Zr-2.5Nb Fuel Cladding
    Kohn E.

    Effect of Stress on Radiation Damage in Neutron Irradiated Zirconium Alloys
    Coleman C., Gilbert R., Northwood D.

    Irradiation Response of the Ordered Phase Zr3Al
    Causey A., Fidleris V., Rosinger H., Schulson E., Urbanic V.

    Fabrication of Zirconium Alloys into Components for Nuclear Reactors
    Cheadle B.

    Mechanical Properties, Anisotropy, and Microstructure of Zircaloy Canning Tubes
    Stehle H., Steinberg E., Tenckhoff E.

    Yield and Fracture of Biaxially Stressed Zircaloy-4 Cladding Tubes at Room Temperature and at 400°C
    Dressler G., Matucha K.

    Progress in Modeling of Zircaloy Nonelastic Deformation Using a Unified Phenomenological Model
    Miller A.

    Fracture of Zircaloy Cladding by Interactions with Uranium-Dioxide Pellets in Water Reactor Fuel Rods
    Smith E.

    Localized Ductility of Irradiated Zircaloy-2 Cladding in Air and Iodine Environments
    Tomalin D.

    Acoustic Emission during Room Temperature Fatigue Crack Growth and Fracture of Zr-2.5Cb Alloy Specimens
    Hutton P., Igarashi M., Schwenk E.

    Susceptibility of Zirconium Alloys to Delayed Hydrogen Cracking
    Ambler J., Coleman C.

    Factors Controlling Hydrogen Assisted Subcritical Crack Growth in Zr-2.5Nb Alloys
    Nuttall K., Simpson L.

    Hydride Reorientation and Fracture in Zirconium Alloys
    Kupcis O., Leemans D., Simpson C.

    Protective Behavior of Anodic Zirconium Oxide Film
    Bedair S., El-Massry N., Hammad F.

    Standardization Works on Testing Procedures for Zircaloy Cladding Tubes in Japan
    Mishima Y.

    Index


    Committee: B10

    Paper ID: STP633-EB

    DOI: 10.1520/STP633-EB

    ISBN-EB: 978-0-8031-4705-8

    ISBN-13: 978-0-8031-0602-4

    ASTM International is a member of CrossRef.

    0-8031-0602-5
    978-0-8031-0602-4
    STP633-EB