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    STP754

    Zirconium in the Nuclear Industry

    Franklin DG
    Published: 1982


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    This publication provides information on the performance of zirconium and zirconium alloy components in nuclear application. Nuclear physicists have discovered that some materials -- zirconium, for example, provide much better neutron economy than conventional materials such as stainless steels.


    Table of Contents

    Introduction

    Discussion

    Summary

    A Review of Texture and Texture Formation in Zircaloy Tubing

    Texture Control of Zircaloy Tubing During Tube Reduction

    Texture Measurement Techniques for Zircaloy Cladding: A Round-Robin Study

    Application of Hydrostatic Extrusion to Fabrication of Zircaloy Tubing

    Beta-Quenching of Zircaloy Cladding Tubes in Intermediate or Final Size—Methods to Improve Corrosion and Mechanical Properties

    New Intermetallic Compounds in Zircaloy-4

    Long-Term Test Results of Promising New Zirconium Alloys

    Clad Failure Modeling

    Chemical Aspects of Iodine-Induced Stress Corrosion Cracking of Zircaloys

    Effect of Zirconium Oxide on the Stress-Corrosion Susceptibility of Irradiated Zircaloy Cladding

    Effects of Temperature and Pressure on the In-Reactor Creepdown of Zircaloy Fuel Cladding

    High-Strength, Creep-Resistant Excel Pressure Tubes

    High-Temperature Irradiation Growth in Zircaloy

    Zircaloy-4 Cladding Deformation During Power Reactor Irradiation

    Burst Criterion of Zircaloy Fuel Claddings in a Loss-of-Coolant Accident

    An Experimental Study of the Deformation of Zircaloy PWR Fuel Rod Cladding Under Mainly Convective Cooling

    Effect of Hydrogen on the Oxygen Embrittlement of Beta-Quenched Zircaloy-4 Fuel Cladding

    Lifetime and Failure Strain Prediction for Material Subjected to Non-stationary Tensile Loading Conditions; Applications to Zircaloy-4

    Flow Stress of Oxygen-Enriched Zircaloy-2 Between 1023 and 1873 K

    A Comparison of the High-Temperature Oxidation Behavior of Zircaloy-4 and Pure Zirconium

    Effect of Texture on Hydride Reorientation and Delayed Hydrogen Cracking in Cold-Worked Zr-2.5Nb

    Mechanism of Accelerated Corrosion in Zircaloy-4 Laser and Electron-Beam Welds

    Waterside Corrosion of Zircaloy-Clad Fuel Rods in a PWR Environment

    Long-Term In-Reactor Corrosion and Hydriding of Zircaloy-2 Tubing


    Committee: B10

    DOI: 10.1520/STP754-EB

    ISBN-EB: 978-0-8031-4823-9

    ISBN-13: 978-0-8031-0754-0

    0-8031-0754-4
    978-0-8031-0754-0
    STP754-EB