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    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

    Summary

    Zircaloy—Three Years after the Hearings

    Diameter Increases in Steam Generating Heavy Water Reactor Zircaloy Cans under Loss-of-Coolant Accident Conditions

    Rupture Characteristics of Zircaloy-4 Cladding with Internal and External Simulation of Reactor Heating

    High Temperature Strain Behavior of Zircaloy-4 and Zr-2.5Nb Fuel Sheaths

    Superplasticity of Zircaloy-4

    Deformation and Rupture Behavior of Zircaloy Cladding under Simulated Loss-of-Coolant Accident Conditions

    Influence of Metallurgical Variables on the Performance of Zircaloy Fuel Sheathing

    Diffusion of Oxygen in Beta-Zircaloy and the High Temperature Zircaloy-Steam Reaction

    Oxidation of Zirconium During a High-Temperature Transient

    Zircaloy-4 Oxidation in Steam Under Transient Oxidizing Conditions

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

    Zircaloy Cladding Behavior During Irradiation Tests Under Power-Cooling-Mismatch Conditions

    Corrosion of Zirconium-Base Alloys—An Overview

    Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments

    Corrosion Monitoring of Steam Generating Heavy Water Reactor Pressure Tubes

    Embrittlement of Zircaloy-4 by Liquid Cesium at 300°C

    Nodular Corrosion of the Zircaloys

    Use of State Variables in the Description of Irradiation Creep and Deformation of Metals

    Irradiation Growth of Zircaloy

    Zircaloy-2 Pressure Tube Elongation at the Hanford N Reactor

    Numerical Model for the Anisotropic Creep of Zircaloy

    Modeling of Localized Deformation in Neutron Irradiated Zircaloy-2

    In-Reactor Creep of Zr-2.5Nb Fuel Cladding

    Effect of Stress on Radiation Damage in Neutron Irradiated Zirconium Alloys

    Irradiation Response of the Ordered Phase Zr3Al

    Fabrication of Zirconium Alloys into Components for Nuclear Reactors

    Mechanical Properties, Anisotropy, and Microstructure of Zircaloy Canning Tubes

    Yield and Fracture of Biaxially Stressed Zircaloy-4 Cladding Tubes at Room Temperature and at 400°C

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

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

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

    Acoustic Emission during Room Temperature Fatigue Crack Growth and Fracture of Zr-2.5Cb Alloy Specimens

    Susceptibility of Zirconium Alloys to Delayed Hydrogen Cracking

    Factors Controlling Hydrogen Assisted Subcritical Crack Growth in Zr-2.5Nb Alloys

    Hydride Reorientation and Fracture in Zirconium Alloys

    Protective Behavior of Anodic Zirconium Oxide Film

    Standardization Works on Testing Procedures for Zircaloy Cladding Tubes in Japan

    Index


    Committee: B10

    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