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    Zirconium in the Nuclear Industry: Eleventh International Symposium

    Bradley ER, Sabol GP
    Published: 1996

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    From its inaugural meeting in Philadelphia in 1968, the ASTM Symposium on Zirconium in the Nuclear Industry has been the premiere vehicle for discussion and documentation of the scientific and technological bases for the utilization of zirconium-based alloys in water-cooled reactors. The eleventh conference in this symposium series, held in September 1995 in Garmisch-Partenkirchen, Germany, continued this tradition of excellence. Attendees to this conference numbered 209, representing 16 countries. After careful peer review and editing, forty-one technical papers presented at this conference are published in this book. The highlights of the oral discussions have also been captured and appear at the end of each paper. This publication also includes two papers that are significant contributions to zirconium technology which served as the basis for the authors receiving the W. J. Kroll awards for 1993 and 1994. In Garmisch-Partenkirchen the awards for these years were presented to J. A. L. Robertson and to the team comprised of Friedrich Garzarolli, Heinz Stehle, and Eckard Steinberg, respectively. These Kroll Award papers represent historical as well as technical significance for the use of zirconium alloys in the nuclear industry.

    Table of Contents

    Learning from History: A Case Study in Nuclear Fuel

    Behavior and Properties of Zircaloys in Power Reactors: A Short Review of Pertinent Aspects in LWR Fuel

    Microstructure of Oxides on Zircaloy-4, 1.0Nb Zircaloy-4, and Zircaloy-2 Formed in 10.3-MPa Steam at 673 K

    The Importance of Oxide Morphology for the Oxidation Rate of Zirconium Alloys

    Effect of Annealing Temperature on Corrosion Behavior and ZrO2 Microstructure of Zircaloy-4 Cladding Tube

    Microstructure of Oxide Films Formed during the Waterside Corrosion of the Zircaloy-4 Cladding in Lithiated Environment

    Mechanisms of LiOH Degradation and H3BO3Repair of ZrO2 Films

    PWR Zircaloy Fuel Cladding Corrosion Performance, Mechanisms, and Modeling

    Correlation Between Electrochemical Properties and Corrosion Resistance of Zirconium Alloys

    Long-Term In Situ Corrosion Investigation of Zr Alloys in Simulated PWR Environment by Electrochemical Measurements

    Anodic Protection Provided by Precipitates in Aqueous Corrosion of Zircaloy

    Investigation of In-Pile Grown Corrosion Films on Zirconium-Based Alloys

    Microstructure Evolutions and Iron Redistribution in Zircaloy Oxide Layers: Comparative Effects of Neutron Irradiation Flux and Irradiation Damages

    Oxide Characteristics and Corrosion and Hydrogen Uptake in Zr-2.5 Nb CANDU Pressure Tubes

    The Effect of the Trace Impurity Uranium on PWR Aqueous Corrosion of Zircaloy-4

    Detrimental Role of Hydrogen on the Corrosion Rate of Zirconium Alloys

    Hydrogen Pickup and Redistribution in Alpha-Annealed Zircaloy-4

    A Unified Model to Describe the Anisotropic Viscoplastic Behavior of Zircaloy-4 Cladding Tubes

    The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2

    Effects of Hydride Precipitate Localization and Neutron Fluence on the Ductility of Irradiated Zircaloy-4

    Fracture Toughness of Zircaloy Cladding Tubes

    A Model for Analysis of the Effect of Final Annealing on the In- and Out-of-Reactor Creep Behavior of Zircaloy Cladding

    Properties of an Irradiated Heat-Treated Zr-2.5Nb Pressure Tube Removed From the NPD Reactor

    Link Between Results of Small- and Large-Scale Toughness Tests on Irradiated Zr-2.5Nb Pressure Tube Material

    Modeling In-Reactor Deformation of Zr-2.5Nb Pressure Tubes in CANDU Power Reactors

    Effect of In-PWR Irradiation on Size, Structure, and Composition of Intermetallic Precipitates of Zr Alloys

    In Situ Studies of Phase Transformations in Zirconium Alloys and Compounds Under Irradiation

    Evolution of Microstructure in Zirconium Alloys During Irradiation

    Influence of Neutron Irradiation on Dislocation Structure and Phase Composition of Zr-Base Alloys

    Non-Linear Irradiation Growth of Cold-Worked Zircaloy-2

    Influence of Iron in the Nucleation of ⟨c⟩ Component Dislocation Loops in Irradiated Zircaloy-4

    Effects of Extrusion-Billet Preheating on the Microstructure and Properties of Zr-2.5Nb Pressure Tube Materials

    Zircaloy-2 Lined Zirconium Barrier Fuel Cladding

    Effects of Microstructure on Ductility and Fracture Resistance of Zr-1.2Sn-1 Nb-0.4Fe Alloy

    Influence of Processing Variables and Alloy Chemistry on the Corrosion Behavior of ZIRLO Nuclear Fuel Cladding

    Effects of Thermomechanical Processing on In-Reactor Corrosion and Post-Irradiation Mechanical Properties of Zircaloy-2

    Embrittlement of Reactor Core Materials

    Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores

    Corrosion Behavior of Duplex and Reference Cladding in NPP Grohnde

    Development of New Zirconium Alloys for a BWR

    Comparison of the Long-Time Corrosion Behavior of Certain Zr Alloys in PWR, BWR, and Laboratory Tests

    In-BWR and Out-of-Pile Nodular Corrosion behavior of Zry-2/4 Type Melts with Varying Fe, Cr, and Ni Content and Varying Process History

    Development of Pressure Tubes with Service Life Greater Than 30 Years

    Committee: B10

    DOI: 10.1520/STP1295-EB

    ISBN-EB: 978-0-8031-5343-1

    ISBN-13: 978-0-8031-2406-6