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    STP1423

    Zirconium in the Nuclear Industry: Thirteenth International Symposium

    Moan GD, Rudling P
    Published: 2002


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    This unique publication provides an international overview on the production and use of Zr alloys, their properties and behavior during nuclear service, the design of Zr components and their testing after service. Papers cover the historical aspects of research on Zr alloys; basic metallurgy, including studies of second phase particles; irradiation creep and growth; material performance during LOCA (loss of coolant accidents); and RIA (reactivity initiated accident).

    Half of the papers in STP 1423 relate to corrosion and hydriding behavior, the most important current issues in the industry. A quarter of the papers deal with in-reactor studies. The remaining papers discussed the behavior and properties of Zr alloys for the intermediate storage of spent fuel.

    STP 1423 is a valuable resource for engineers and scientists involved in the processing and properties of Zr alloys, the production and use of Zr alloy reactor components, their behavior during service and property changes that occur with increasing neutron fluences.

    42 peer-reviewed papers provide an international overview of the production and use of Zr alloys.

    Learn about:

    • Production and use of Zr alloys in the nuclear industry

    • Properties and behavior during nuclear service

    • Corrosion and hydriding during service

    • Dimensional changes during service

    • Design of Zr components

    • Testing after service

    • Effect of alloy microstructure and composition on behavior

    • Studies of the composition, properties and behavior of the oxides formed during service


    Table of Contents

    Simulating the Behavior of Zirconium-Alloy Components in Nuclear Reactors

    Physical Phenomena Concerning Corrosion Under Irradiation of Zr Alloys

    Role of the Second-Phase Particles in Zirconium Binary Alloys

    Synchrotron Radiation Study of Second-Phase Particles and Alloying Elements in Zirconium Alloys

    The Behavior of Intermetallic Precipitates in Highly Irradiated BWR LTP Cladding

    Effects of Hydrogen Pickup and Second-Phase Particle Dissolution on the In-Reactor Corrosion Performance of BWR Claddings

    Alternative Zr Alloys with Irradiation Resistant Precipitates for High Burnup BWR Application

    Mössbauer Investigations of the Chemical States of Tin and Iron Atoms in Zirconium Alloy Oxide Film

    Chemical State Analysis of Sn and Fe in ZrO2 by Mössbauer Spectroscopy

    The Role of Lithium and Boron on the Corrosion of Zircaloy-4 Under Demanding PWR-Type Conditions

    Study of the Mechanisms Controlling the Oxide Growth Under Irradiation: Characterization of Irradiated Zircaloy-4 and Zr-1Nb-O Oxide Scales

    Hydrogen Evolution and Pickup During the Corrosion of Zirconium Alloys: A Critical Evaluation of the Solid State and Porous Oxide Electrochemistry

    The Influence of Material Variables on Corrosion and Deuterium Uptake of Zr-2.5Nb Alloy During Irradiation

    The Cause for Enhanced Corrosion of Zirconium Alloys by Hydrides

    The Effect of Minor Alloying Elements on Oxidation and Hydrogen Pickup in Zr-2.5Nb

    Role of Microchemistry and Microstructure on Variability in Corrosion and Deuterium Uptake of Zr-2.5Nb Pressure Tube Material

    Temperature and Hydrogen Concentration Limits for Delayed Hydride Cracking in Irradiated Zircaloy

    Experimental Study and Preliminary Thermodynamic Calculations of the Pseudo-Ternary Zr-Nb-Fe-(O,Sn) System

    Activated Slip Systems and Localized Straining of Irradiated Zr Alloys in Circumferential Loadings

    Effects of Neutron Irradiation on the Microstructure of Alpha-Annealed Zircaloy-4

    Plastic Deformation and Fracture During the Zr1%Nb Tube Production

    Development of Crystallographic Texture in CANDU Calandria Tubes

    Impact of Hydrogen on Dimensional Stability of ZIRLO Fuel Assemblies

    In-PWR Irradiation Performance of Dilute Tin-Zirconium Advanced Alloys

    Predicting the In-Reactor Mechanical Behavior of Zr-2.5Nb Pressure Tubes from Postirradiation Microstructural Examination Data

    The Influence of In-Situ Clad Straining on the Corrosion of Zircaloy in a PWR Water Environment

    Characteristics of Hydride Precipitation and Reorientation in Spent-Fuel Cladding

    Test Reactor Studies of the Shadow Corrosion Phenomenon

    An In-Reactor Simulation Test to Evaluate Root Cause of Secondary Degradation of Defective BWR Fuel Rod

    The Effect of Short-Term Dry-Out Transients on the Cladding Properties of Fresh and Pre-Irradiated Fuel Rods

    Effects of Alpha-Beta Transformation on High Temperature (LOCA) Creep behavior of Zr-Alloys

    Influence of Hydrogen Content on the α/β Phase Transformation Temperatures and on the Thermal-Mechanical Behavior of Zy-4, M4 (ZrSnFeV), and M5™ (ZrNbO) Alloys During the First Phase of LOCA Transient

    On the Embrittlement of Zircaloy-4 Under RIA-Relevant Conditions

    Impact of Irradiation Defects Annealing on Long-Term Thermal Creep of Irradiated ircaloy-4 Cladding Tube

    Influence of Composition and Condition on In-PWR Behavior of Zr-Sn-Nb-FeCrV Alloys

    Influence of Zirconium Alloy Chemical Composition on Microstructure Formation and Irradiation Induced Growth

    Quantitative Assessment of Irradiation Effect on Creep and Corrosion Properties of Zr-Base Alloys

    Variability of In-Reactor Diametral Deformation for Zr-2.5Nb Pressure Tubing

    About the Mechanisms Governing the Hydrogen Effect on Viscoplasticity of Unirradiated Fully Annealed Zircaloy-4 Sheet

    Irradiation Creep Behavior of Zr-Base Alloys

    The Effect of Small Concentrations of Sulfur on the Plasticity of Zirconium Alloys at Intermediate Temperatures

    Author Index

    Subject Index


    Committee: B10

    DOI: 10.1520/STP1423-EB

    ISBN-EB: 978-0-8031-5468-1

    ISBN-13: 978-0-8031-2895-8

    ASTM International is a member of CrossRef.

    0-8031-2895-9
    978-0-8031-2895-8
    STP1423-EB