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    Zirconium in the Nuclear Industry

    Schemel JH, Papazoglou TP
    Published: 1979

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    The Fourth International Conference on Zirconium in the Nuclear Industry was held 26-29 June 1978, in Stratford-upon-Avon, England. This conference was sponsored by the American Society for Testing and Materials (ASTM) Committee B10 on Refractory Metals and Alloys in cooperation with the American Nuclear Society, the British Nuclear Energy Society, and The Metals Society (U.K).. The program was planned by representatives of ASTM Subcommittee B10.02 on Zirconium and Hafnium and K. G. Sumner of British Nuclear Fuels, Ltd. J. H. Schemel, Sandvik Special Metals Corp., and K. G. Sumner, BNFL, served as the conference co-chairmen, T. P. Papazoglou, The Babcock & Wilcox Company, as editorial chairman, and H. M. Cobb, ASTM Staff, as conference coordinator.

    Table of Contents


    Zirconium Technology—Twenty Years of Evolution

    Zircaloy Performance in Light-Water Reactors

    Effect of Material and Environmental Variables on Localized Corrosion of Zirconium Alloys

    Ductility of Zircaloy Canning Tubes in Relation to Stress Ratio in Biaxial Testing

    Localized Ductility Method for Evaluating Zircaloy-2 Cladding

    Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions

    Further Evidence of Zircaloy Corrosion in Fuel Elements Irradiated in a Steam Generating Heavy Water Reactor

    Performance of Irradiated Copper and Zirconium Barrier-Modified Zircaloy Cladding Under Simulated Pellet-Cladding Interaction Conditions

    A Survey of the Chemical Environments for Activity in the Embrittlement of Zircaloy-2

    Effect of Irradiation on the Strength, Ductility, and Defect Sensitivity of Fully Recrystallized Zircaloy Tube

    Primary Creep of Zircaloy-2 Under Irradiation

    Analysis of Irradiation Growth and Multiaxial Deformation Behavior of Nuclear Fuel Cladding

    Analysis of Pressurized Water Reactor Fuel Pin Length Changes

    Irradiation Growth in Cold-Worked Zircaloy-2

    Stress Corrosion Crack Initiation and Growth and Formation of Pellet-Clad Interaction Defects

    Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon

    Iodine-Induced Stress Corrosion Cracking of Fixed Deflection Stressed Slotted Rings of Zircaloy Fuel Cladding

    A Stress Corrosion Cracking Model for Pellet-Cladding Interaction Failures in Light-Water Reactor Fuel Rods

    Hydride Cracks as Initiators for Stress Corrosion Cracking of Zircaloys

    Predicting High-Temperature Transient Deformation from Microstructural Models

    A Microstructure-Based Constitutive Relation for Dilute Alloys of α-Zirconium

    A Comparison of Experimental Cladding Microstructure Resulting from Uranium Oxide-Zircaloy Interaction with a Diffusional Assessment of Oxygen Transport for a Coupled Two-Media Problem

    Advances in Understanding and Predicting Inelastic Deformation in Zircaloy

    Zirconium Cladding Deformation in a Steam Environment with Transient Heating

    Influence of Iodine on the Strain and Rupture Behavior of Zircaloy-4 Cladding Tubes at High Temperatures

    Studies on Zircaloy Fuel Clad Ballooning in a Loss-of-Coolant Accident—Results of Burst Tests with Indirectly Heated Fuel Rod Simulators

    Effect of β-Phase Heat Treatment on the Subsequent α-Phase Ballooning Behavior of Zircaloy-4 Fuel Sheaths

    Tube-Burst Response of Irradiated Zircaloy Spent-Fuel Cladding

    A Proposed Criterion for the Oxygen Embrittlement of Zircaloy-4 Fuel Cladding

    Effect of Oxygen on the Deformation of Zircaloy-2 at Elevated Temperatures

    Evaluation Models of Zircaloy Oxidation in Light of Recent Experiments

    Chemical Interaction Between Uranium Oxide and Zircaloy-4 in the Temperature Range Between 900 and 1500°C

    Interaction of Oxidation and Creep in Zircaloy-2

    Embrittlement of Zircaloy-Clad Fuel Rods Irradiated Under Film Boiling Conditions

    Development of an Oxygen Embrittlement Criterion for Zircaloy Cladding Applicable to Loss-of-Coolant Accident Conditions in Light-Water Reactors

    Committee: B10

    DOI: 10.1520/STP681-EB

    ISBN-EB: 978-0-8031-4749-2

    ISBN-13: 978-0-8031-0601-7