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Digital Library / STP / STP1245-EB
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Zirconium in the Nuclear Industry: Tenth International Symposium
Garde AM
Pages: 796
Published: 1994
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Table of Contents
Development of Zirconium-Barrier Fuel Cladding
Armijo JS, Coffin LF, Rosenbaum HS
Zirconium Alloy Performance in Light Water Reactors: A Review of UK and Scandinavian Experience
Pickman DO
The Effect of Fluence and Irradiation Temperature on Delayed Hydride Cracking in Zr-2.5Nb
Sagat S, Coleman CE, Griffiths M, Wilkins BJS
A Study of the Hydrogen Uptake Mechanism in Zirconium Alloys
Elmoselhi MB, Warr BD, McIntyre S
Hydrogen Absorption Kinetics During Zircaloy Oxidation in Steam
Charquet D, Rudling P, Mikes-Lindback M, Barberis P
Scanning Electron Microscope Techniques for Studying Zircaloy Corrosion and Hydriding
Schrire DI, Pearce JH
Growth and Characterization of Oxide Films on Zirconium-Niobium Alloys
Urbanic VF, Chan PK, Khatamian D, Woo O-TT
Correlation Between Irradiated and Unirradiated Fracture Toughness of Zr-2.5Nb Pressure Tubes
Davies PH, Hosbons RR, Griffiths M, Chow CK
Variability of Irradiation Growth in Zr-2.5Nb Pressure Tubes
Fleck RG, Elder JE, Causey AR, Holt RA
Change of Mechanical Properties by Irradiation and Evaluation of the Heat-Treated Zr-2.5Nb Pressure Tube
Koike MH, Akiyama T, Nagamatsu K, Shibahara I
On the Anisotropy of In-Reactor Creep of Zr-2.5Nb Tubes
Causey AR, Elder JE, Holt RA, Fleck RG
Fabrication of Zr-2.5Nb Pressure Tubes to Minimize the Harmful Effects of Trace Elements
Theaker JR, Choubey R, Moan GD, Aldridge SA, Davis L, Graham RA, Coleman CE
Modeling of Damage in Cold Pilgering
Aubin JL, Girard E, Montmitonnet P
Mitigation of Harmful Effects of Welds in Zirconium Alloy Components
Coleman CE, Doubt GL, Fong RWL, Root JH, Bowden JW, Sagat S, Webster RT
Fabrication Process of High Nodular Corrosion-Resistant Zircaloy-2 Tubing
Abe H, Matsuda K, Hama T, Konishi T, Furugen M
Corrosion Behavior of Zircaloy-4 Sheets Produced Under Various Hot-Rolling and Annealing Conditions
Anada H, Nomoto K, Shida Y
Optimization of PWR Behavior of Stress-Relieved Zircaloy-4 Cladding Tubes by Improving the Manufacturing and Inspection Process
Mardon J-P, Charquet D, Senevat J
Experimental and Theoretical Studies of Parameters that Influence Corrosion of Zircaloy-4
Billot P, Robin J-C, Giordano A, Peybernčs J, Thomazet J, Amanrich H
Aqueous Chemistry of Lithium Hydroxide and Boric Acid and Corrosion of Zircaloy-4 and Zr-2.5Nb Alloys
Ramasubramanian N, Balakrishnan PV
Corrosion Behavior of Irradiated Zircaloy
Cheng B-C, Kruger RM, Adamson RB
Correlation of Transmission Electron Microscopy (TEM) Microstructure Analysis and Texture with Nodular Corrosion Behavior for Zircaloy-2
Herb BJ, McCarthy JM, Wang CT, Ruhmann H
Micro-Characterization of Corrosion Resistant Zirconium-Based Alloys
Isobe T, Matsuo Y, Mae Y
An Experimental Investigation into the Oxidation of Zircaloy-4 at Elevated Pressures in the 750 to 1000°C Temperature Range
Bramwell IL, Haste TJ, Worswick D, Parsons PD
Prediction of Creep Anisotropy in Zircaloy Cladding
Perkins RA, Shann S-H
Microstructure and Properties of Corrosion-Resistant Zirconium-Tin-lron-Chromium-Nickel Alloys
Bradley ER, Nyström A-L
Fatigue Behavior of Irradiated and Unirradiated Zircaloy and Zirconium
Wisner SB, Reynolds MB, Adamson RB
Effect of Irradiation on the Microstructure of Zircaloy-4
Gilbon D, Simonot C
Irradiation Effect on Fatigue Behavior of Zircaloy-4 Cladding Tubes
Soniak A, Lansiart S, Royer J, Mardon J-P, Waeckel N
Grain-by-Grain Study of the Mechanisms of Crack Propagation During Iodine Stress Corrosion Cracking of Zircaloy-4
Haddad RE, Dorado AO
Microstructure of Oxide Layers Formed During Autoclave Testing of Zirconium Alloys
Wadman B, Lai Z, Andrén H-O, Nyström A-L, Rudling P, Pettersson H
Corrosion Performance of New Zircaloy-2-Based Alloys
Rudling P, Mikes-Lindbäck M, Lethinen B, Andrén H-O, Stiller K
Examinations of the Corrosion Mechanism of Zirconium Alloys
Beie HJ, Mitwalsky A, Garzarolli F, Ruhmann H, Sell HJ
On the Initial Corrosion Mechanism of Zirconium Alloy: Interaction of Oxygen and Water with Zircaloy at Room Temperature and 450°C Evaluated by X-Ray Absorption Spectroscopy and Photoelectron Spectroscopy
Döbler U, Knop A, Ruhmann H, Beie H-J
How the Tetragonal Zirconia is Stabilized in the Oxide Scale that is Formed on a Zirconium Alloy Corroded at 400°C in Steam
Godlewski J
Oxidation of Intermetallic Precipitates in Zircaloy-4: Impact of Irradiation
Pęcheur D, Lefebvre F, Motta AT, Lemaignan C, Charquet D
Corrosion Optimized Zircaloy for Boiling Water Reactor (BWR) Fuel Elements
Garzarolli F, Schumann R, Steinberg E
In-Reactor Corrosion Performance of ZIRLO and Zircaloy-4
Sabol GP, Comstock RJ, Weiner RA, Larouere P, Stanutz RN
Phenomenological Study of In-Reactor Corrosion of Zircaloy-4 in Pressurized Water Reactors
Kim YS, Rheem KS, Min DK
Corrosion Behavior of Zircaloy-4 Cladding with Varying Tin Content in High-Temperature Pressurized Water Reactors
Garde AM, Pati SR, Krammen MA, Smith GP, Endter RK
Effects of Pressurized Water Reactor (PWR) Coolant Chemistry on Zircaloy Corrosion Behavior
Karlsen T, Vitanza C
Author Index
Subject Index
Committee: B10
Paper ID: STP1245-EB
DOI: 10.1520/STP1245-EB
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STP1245-EB
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