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    Zirconium in the Nuclear Industry: 17th Volume

    Comstock Bob, Barbéris Pierre
    Published: 2015

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    This STP contains the 40 peer-reviewed papers – two papers were named Kroll winners – that were presented at the 17th International Symposium on Zirconium in the Nuclear Industry held in Hyderabad, Andhra Pradesh, India in 2013.

    A dominant theme in these papers is the role of hydrogen on the performance of zirconium alloy components. Issues discussed in this volume where performance was dominated by hydrogen included the following:

    • Failure of BWR fuel rods was attributed to the localization of hydrides following accelerated corrosion and subsequent cracking of the hydride lenses. Despite an extensive investigation, the cause of the accelerated corrosion was not definitively identified.

    • The growth of beta-quenched Zircaloy-2 BWR channels was driven late in life by accelerated hydrogen pickup that coincided with the dissolution of second phase particles.

    • As reorientation of hydrides plays an important role during dry storage, in-situ measurements were performed to gain new insights into the reorientation of hydrides in Zircaloy-4.

    • Delayed hydride cracking (DHC) growth rate of in-service Zr-2.5Nb CANDU pressure tubes was controlled by thermal and irradiation effects on the microstructure (e.g, decomposition and reconstitution of the beta phase controlling hydrogen diffusion to the crack tip).

    In addition to papers that highlight the impact of hydrogen, several papers focus on understanding the mechanisms of hydrogen ingress into the metal or understanding the interaction of hydrogen with point defects and dislocation loops in the matrix.

    Table of Contents

    Reflections on the Development of the “f” Texture Factors for Zirconium Components and the Establishment of Properties of the Zirconium–Hydrogen System

    Displacive and Diffusional Transformations of the Beta Phase in Zirconium Alloys

    Effect of Hydrogen on Dimensional Changes of Zirconium and the Influence of Alloying Elements: First-Principles and Classical Simulations of Point Defects, Dislocation Loops, and Hydrides

    Phase Field Modeling of Microstructure Evolution in Zirconium Base Alloys

    Thermodynamics of Zr Alloys: Application to Heterogeneous Materials

    Influence of Sn on Deformation Mechanisms During Room Temperature Compression of Binary Zr–Sn Alloys

    Impact of Iron in M5TM6

    Microstructure and Properties of a Three-Layer Nuclear Fuel Cladding Prototype Containing Erbium as a Neutronic Burnable Poison

    Characterizing Quenched Microstructures in Relation to Processing

    Identification of Safe Hot-Working Conditions in Cast Zr-2.5Nb

    A Numerical Study of the Effect of Extrusion Parameters on the Temperature Distribution in Zr–2.5Nb

    Study on Effect of Processing on Texture Development in Zirconium-2.5 % Niobium Alloy Tubes

    Numerical Modeling of Fuel Rod Resistance Butt Welding

    Application of Coating Technology on Zirconium-Based Alloy to Decrease High-Temperature Oxidation

    Oxidation Mechanism in Zircaloy-2—The Effect of SPP Size Distribution

    Effect of Sn on Corrosion Mechanisms in Advanced Zr-Cladding for Pressurised Water Reactors

    Understanding of Corrosion Mechanisms of Zirconium Alloys after Irradiation: Effect of Ion Irradiation of the Oxide Layers on the Corrosion Rate

    Effect of Alloying Elements on Hydrogen Pickup in Zirconium Alloys

    Toward a Comprehensive Mechanistic Understanding of Hydrogen Uptake in Zirconium Alloys by Combining Atom Probe Analysis With Electronic Structure Calculations

    Corrosion and Hydrogen Uptake in Zirconium Claddings Irradiated in Light Water Reactors

    Oxidation and Hydrogen Uptake of ZIRLO Structural Components Irradiated to High Burn-Up

    Performance and Property Evaluation of High-Burnup Optimized ZIRLO™ Cladding

    Corrosion, Dimensional Stability and Microstructure of VVER-1000 E635 Alloy FA Components at Burnups up to 72 MWday/kgU

    Corrosion and Hydriding Model for Zircaloy-2 Pressure Tubes of Indian Pressurised Heavy Water Reactors

    Oxide Surface Peeling of Advanced Zirconium Alloy Cladding after High Burnup Irradiation in Pressurized Water Reactors

    The Effects of Microstructure and Operating Conditions on Irradiation Creep of Zr-2.5Nb Pressure Tubing

    Breakthrough in Understanding Radiation Growth of Zirconium

    Microstructural Evolution of M5TM7 Alloy Irradiated in PWRs up to High Fluences—Comparison With Other Zr-Based Alloys

    Modeling Irradiation Damage in Zr-2.5Nb and Its Effects on Delayed Hydride Cracking Growth Rate

    Understanding the Drivers of In-Reactor Growth of β-Quenched Zircaloy-2 BWR Channels

    Impact of Hydrogen Pick-Up and Applied Stress on C-Component Loops: Toward a Better Understanding of the Radiation Induced Growth of Recrystallized Zirconium Alloys

    Contribution to the Study of the Pseudobinary Zr1Nb–O Phase Diagram and Its Application to Numerical Modeling of the High-Temperature Steam Oxidation of Zr1Nb Fuel Cladding

    Experimental Comparison of the Behavior of E110 and E110G Claddings at High Temperature

    Effect of Pre-Oxide on Zircaloy-4 High-Temperature Steam Oxidation and Post-Quench Mechanical Properties

    Deviations From Parabolic Kinetics During Oxidation of Zirconium Alloys

    Influence of Steam Pressure on the High Temperature Oxidation and Post-Cooling Mechanical Properties of Zircaloy-4 and M5 Cladding (LOCA Conditions)

    Analysis of the Secondary Cladding Hydrogenation During the Quench–LOCA Bundle Tests With Zircaloy-4 Claddings and its Influence on the Cladding Embrittlement

    Effect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes

    Mechanisms of Hydride Reorientation in Zircaloy-4 Studied in Situ

    Hydriding Induced Corrosion Failures in BWR Fuel

    Author and Subject Index

    Committee: B10

    DOI: 10.1520/STP1543-EB

    ISBN-EB: 978-0-8031-7580-8

    ISBN-13: 978-0-8031-7529-7

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