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SEDL / STP / STP633-EB
45 papers present vitally important data on the behavior of zirconium alloys in nuclear reactor environments.
Table of Contents
Introduction
Lowe A., Parry G.
Summary
Lowe A., Parry G.
Zircaloy—Three Years after the Hearings
Johnston W.
Diameter Increases in Steam Generating Heavy Water Reactor Zircaloy Cans under Loss-of-Coolant Accident Conditions
Hindle E., Rose K.
Rupture Characteristics of Zircaloy-4 Cladding with Internal and External Simulation of Reactor Heating
Barber A., Fiveland W., Lowe A.
High Temperature Strain Behavior of Zircaloy-4 and Zr-2.5Nb Fuel Sheaths
Foote D., Hunt C.
Superplasticity of Zircaloy-4
Boček M., Hofmann P., Petersen C.
Deformation and Rupture Behavior of Zircaloy Cladding under Simulated Loss-of-Coolant Accident Conditions
Chung H., Garde A., Kassner T.
Influence of Metallurgical Variables on the Performance of Zircaloy Fuel Sheathing
Bain A., Hardy D.
Diffusion of Oxygen in Beta-Zircaloy and the High Temperature Zircaloy-Steam Reaction
Cathcart J., Druschel R., McKee R., Pawel R., Perkins R., Yurek G.
Oxidation of Zirconium During a High-Temperature Transient
Jones S., Ledoux G., Sawatzky A.
Zircaloy-4 Oxidation in Steam Under Transient Oxidizing Conditions
Ballinger R., Biederman R., Dobson W.
Oxidation of Zirconium Alloys in Steam at 1000 to 1850°C
Urbanic V.
Zircaloy Cladding Behavior During Irradiation Tests Under Power-Cooling-Mismatch Conditions
Gibson G., Hobbins R., Smolik G.
Corrosion of Zirconium-Base Alloys—An Overview
Hillner E.
Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments
Garlick A., Sumerling R., Trowse F.
Corrosion Monitoring of Steam Generating Heavy Water Reactor Pressure Tubes
Sheppard M., Tyzack C.
Embrittlement of Zircaloy-4 by Liquid Cesium at 300°C
Cubicciotti D., Jones R., Syrett B.
Nodular Corrosion of the Zircaloys
Horton R., Johnson A.
Use of State Variables in the Description of Irradiation Creep and Deformation of Metals
Hart E., Li C.
Irradiation Growth of Zircaloy
Adamson R.
Zircaloy-2 Pressure Tube Elongation at the Hanford N Reactor
Alexander W., Fidleris V., Holt R.
Numerical Model for the Anisotropic Creep of Zircaloy
Franklin D., Franz W.
Modeling of Localized Deformation in Neutron Irradiated Zircaloy-2
Adamson R., Lee D.
In-Reactor Creep of Zr-2.5Nb Fuel Cladding
Kohn E.
Effect of Stress on Radiation Damage in Neutron Irradiated Zirconium Alloys
Coleman C., Gilbert R., Northwood D.
Irradiation Response of the Ordered Phase Zr3Al
Causey A., Fidleris V., Rosinger H., Schulson E., Urbanic V.
Fabrication of Zirconium Alloys into Components for Nuclear Reactors
Cheadle B.
Mechanical Properties, Anisotropy, and Microstructure of Zircaloy Canning Tubes
Stehle H., Steinberg E., Tenckhoff E.
Yield and Fracture of Biaxially Stressed Zircaloy-4 Cladding Tubes at Room Temperature and at 400°C
Dressler G., Matucha K.
Progress in Modeling of Zircaloy Nonelastic Deformation Using a Unified Phenomenological Model
Miller A.
Fracture of Zircaloy Cladding by Interactions with Uranium-Dioxide Pellets in Water Reactor Fuel Rods
Smith E.
Localized Ductility of Irradiated Zircaloy-2 Cladding in Air and Iodine Environments
Tomalin D.
Acoustic Emission during Room Temperature Fatigue Crack Growth and Fracture of Zr-2.5Cb Alloy Specimens
Hutton P., Igarashi M., Schwenk E.
Susceptibility of Zirconium Alloys to Delayed Hydrogen Cracking
Ambler J., Coleman C.
Factors Controlling Hydrogen Assisted Subcritical Crack Growth in Zr-2.5Nb Alloys
Nuttall K., Simpson L.
Hydride Reorientation and Fracture in Zirconium Alloys
Kupcis O., Leemans D., Simpson C.
Protective Behavior of Anodic Zirconium Oxide Film
Bedair S., El-Massry N., Hammad F.
Standardization Works on Testing Procedures for Zircaloy Cladding Tubes in Japan
Mishima Y.
Index
Committee: B10
Paper ID: STP633-EB
DOI: 10.1520/STP633-EB
ISBN-EB: 978-0-8031-4705-8
 ASTM International is a member of CrossRef.
0-8031-0602-5
978-0-8031-0602-4
STP633-EB
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