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    Zirconium in Nuclear Applications

    Schemel JH, Rosenbaum HS
    Published: 1974

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    The Symposium on Zirconium in Nuclear Applications was held 21–24 August 1973 in Portland, Ore., and was co-sponsored by the American Society for Testing and Materials (ASTM) Committee B-10 on Refractory Metals and Alloys and by the Institute of Metals Division, The Metallurgical Society of the American Institute of Mining, Metallurgical, and Petroleum Engineers (AIME). Program planning was done jointly by representatives of the ASTM Subcommittee B10.02 on Zirconium and Hafnium and by the TMS-IMD Nuclear Metallurgy Committee of AIME. J.H. Schemel, AMAX Specialty Metals Corp., presided as the symposium chairman, and H.S. Rosenbaum, General Electric Co., served as the symposium co-chairman. The session chairman and co-chairman included: A.L. Bement, Jr., Massachusetts Institute of Technology; A. Lowe, Jr., Babcock & Wilcox Company; M.L. Picklesimer, National Bureau of Standards; P.L. Rittenhouse, Union Carbide Corporation; D.E. Thomas, Westinghouse Electric Company; J. Wahler, Combustion Engineering, Inc.; and C.D. Williams, General Electric Company.

    Table of Contents






    Zirconium for Nuclear Primary Steam Systems

    Development of a Closed-End Burst Test Procedure for Zircaloy Tubing

    Uniform Ultrasonic Inspection of Fuel Sheath Tubing

    Defect Sensitivity in “Lamb Wave” Testing of Thin-Walled Tubing

    Improved Metallography of Zirconium Alloys

    Determination of Solid Solubility Limit of Hydrogen in Alpha Zirconium by Internal Friction Measurements

    Dual Analysis of Longitudinal and Transverse Zirconium Tensile Stress-Strain Data

    Determination of Complete Plane-Stress Yield Loci of Zircaloy Tubing

    Effect of Thermomechanical Processing and Heat Treatment on the Properties of Zr-3Nb-1Sn Strip and Tubing

    Potential for Improvement of Mechanical Properties in Zircaloy Cold-Rolled Strip and Sheet

    Effect of the Annealing Temperature on the Creep Strength of Cold-Worked Zircaloy -4 Cladding

    Thermomechanical Control of Texture and Tensile Properties of Zircaloy-4 Plate

    Creep Strength of Zircaloy Tubing at 400°C as Dependent on Metallurgical Structure and Texture

    Pilger Tooling Design for Texture Control

    Operable Deformation Systems and Mechanical Behavior of Textured Zircaloy Tubing

    Directionality of the Grain Boundary Hydride in Zircaloy-2

    Use of Ion Bombardment to Study Irradiation Damage in Zirconium Alloys

    Mechanisms of Irradiation Creep in Zirconium-Base Alloys

    In—Reactor Creep of Zr—2.5Nb Tubes at 570 K

    In—Reactor Stress Relaxation of Zirconium Alloys

    High Deformation Creep Behavior of 0.6-in.-Diameter Zirconium Alloy Tubes Under Irradiation

    Suppression of Void Formation in Zirconium

    Deformation of Irradiated Zirconium-Niobium Alloys

    Variation of Zircaloy Fracture Toughness in Irradiation

    Strength and Ductility of Neutron Irradiated and Textured Zircaloy-2

    Plastic Instability in Irradiated Zr-Sn and Zr-Nb Alloys

    Assessment of Fracture Studies on Zircaloy-2 Pressure Tubes

    Irradiation Damage Recovery in Some Zirconium Alloys

    Stress Corrosion Cracking of Zircaloys in Iodine Containing Environments

    Microstructure of the Oxide Films Formed on Zirconium-Based Alloys

    Corrosion and Hydriding Performance of Zircaloy Tubing After Extended Exposure in the Shippingport Pressurized Water Reactor

    Characterization of Zircaloy Oxidation Films

    Fracture of Zircaloy-2 in an Environment Containing Iodine

    Study of Zirconium Alloy Corrosion Parameters in the Advanced Test Reactor

    Effect of Surface Treatment on the Irradiation Enhancement of Corrosion of Zircaloy-2 in HBWR

    Committee: B10

    DOI: 10.1520/STP551-EB

    ISBN-EB: 978-0-8031-4640-2

    ISBN-13: 978-0-8031-0757-1