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    Zirconium in the Nuclear Industry

    Franklin DG, Adamson RB
    Published: 1984

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    Discusses zirconium extraction and reduction to tubes; recent work on basic properties; fracture properties, including stress-corrosion cracking and hydrogen cracking; high-temperature properties; texture-related work with special interest in creep and growth; and waterside corrosion.

    Table of Contents













    Extractive Metallurgy of Zirconium—1945 to the Present

    New Process for Zirconium and Hafnium Separation

    New Solvent Extraction Process for Zirconium and Hafnium

    Electron-Beam Remelting of Zircaloys

    Characteristics of Hydrostatically Extruded Zr-2.5Nb Alloy

    Pressure Tube Development for CANDU Reactors

    Analytical Approaches and Experimental Verification to Describe the Influence of Cold Work and Heat Treatment on the Mechanical Properties of Zircaloy Cladding Tubes

    On-Line Ultrasonic Inside-Diameter Control System for Zircaloy Tube Etching

    Variation in the Strain Anisotropy of Zircaloy with Temperature and Strain

    A Study on the Nonelastic Behavior of Zircaloy Fuel Cladding

    Flow Softening and Anneal Hardening in Two-Phase Zr-2.5Nb

    On the Short-Time Recrystallization of Zircaloy-4 Above 600°C

    Orientation of Hydrides in Zirconium Alloy Tubes

    Correlation Between Fabrication Parameters, Microstructure, and Texture in Zircaloy Tubing

    On the Texture Measurements for Zircaloy Cladding Tube

    Texture Effect of Zircaloy on Ultrasonic Velocity

    In-Reactor Creep of Zr-2.5Nb

    Influence of Metallurgical Condition on the In-Reactor Dimensional Changes of Zircaloy Fuel Rods

    Deformation Characteristics of Cold-Worked and Recrystallized Zircaloy-4 Cladding

    Irradiation Creep and Growth During Proton and Neutron Bombardment of Zircaloy-2 Plate

    Irradiation Growth of Annealed Zircaloy-2

    Irradiation-Induced Dimensional Changes in Polycrystalline Iodide Zirconium and Zircaloy-2

    Irradiation Growth in Zr-2.5Nb

    High-Fluence Irradiation Growth of Zirconium Alloys at 644 to 725 K

    Irradiation Growth of Zircaloy (LWBR Development Program)

    Characterization of ZrO2 Films Formed In-Reactor and Ex-Reactor to Study the Factors Contributing to the In-Reactor Waterside Corrosion of Zircaloy

    Improved Characterization of Aqueous Corrosion Kinetics of Zircaloy-4

    Effect of Lithium Hydroxide on the Corrosion Behavior of Zircaloy-4

    Application of Alternating-Current Impedance Measurements to Characterize Zirconium Alloy Oxidation Films

    Observations of Accelerated Hydriding in Zirconium Alloys

    Out-of-Pile Nickel Alloy-Induced Accelerated Hydriding of Zircaloy Fasteners

    Modeling of Zircaloy Stress-Corrosion Cracking: Texture Effects and Dry Storage Spent Fuel Behavior

    Influence of Crystallographic Texture and Test Temperature on Initiation and Propagation of Iodine Stress-Corrosion Cracks in Zircaloy

    Effect of Direction of Approach to Temperature on the Delayed Hydrogen Cracking Behavior of Cold-Worked Zr-2.5Nb

    Cracking Zirconium Alloys in Hydrogen

    Effect of Bundle Size on Cladding Deformation in LOCA Simulation Tests

    Deformation and Failure of Zircaloy Fuel Sheaths Under LOCA Conditions

    Estimation of Conservatism of Present Embrittlement Criteria for Zircaloy Fuel Cladding Under LOCA

    Out-of-Pile Experiments on the High-Temperature Behavior of Zircaloy-4 Clad Fuel Rods

    Comparison of High-Temperature Steam Oxidation Kinetics Under LWR Accident Conditions: Zircaloy-4 Versus Austenitic Stainless Steel No. 1.4970

    Fuel Rod Deformation in LOCA and Severe Core Damage Accidents

    High-Temperature Oxidation of Zircaloy in Hydrogen-Steam Mixtures

    Physical and Chemical Phenomena Associated with the Dissolution of Solid UO2 by Molten Zircaloy-4

    Committee: B10

    DOI: 10.1520/STP824-EB

    ISBN-EB: 978-0-8031-4895-6

    ISBN-13: 978-0-8031-0270-5