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Digital Library / STP / STP1295-EB
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Zirconium in the Nuclear Industry: Eleventh International Symposium
Bradley ER, Sabol GP
Pages: 898
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
Overview PDF
Learning from History: A Case Study in Nuclear Fuel
Robertson J.
Behavior and Properties of Zircaloys in Power Reactors: A Short Review of Pertinent Aspects in LWR Fuel
Garzarolli F., Stehle H., Steinberg E.
Microstructure of Oxides on Zircaloy-4, 1.0Nb Zircaloy-4, and Zircaloy-2 Formed in 10.3-MPa Steam at 673 K
Anada H., Takeda K.
The Importance of Oxide Morphology for the Oxidation Rate of Zirconium Alloys
Wikmark G., Rudling P., Lehtinen B., Hutchinson B., Oscarsson A., Ahlberg E.
Effect of Annealing Temperature on Corrosion Behavior and ZrO2 Microstructure of Zircaloy-4 Cladding Tube
Anada H., Herb B., Nomoto K., Hagi S., Graham R., Kuroda T.
Microstructure of Oxide Films Formed during the Waterside Corrosion of the Zircaloy-4 Cladding in Lithiated Environment
Pêcheur D., Godlewski J., Billot P., Thomazet J.
Mechanisms of LiOH Degradation and H3BO3Repair of ZrO2 Films
Cox B., Ungurelu M., Wong Y., Wu C.
PWR Zircaloy Fuel Cladding Corrosion Performance, Mechanisms, and Modeling
Cheng B., Gilmore P., Klepfer H.
Correlation Between Electrochemical Properties and Corrosion Resistance of Zirconium Alloys
Ito Y., Furuya T.
Long-Term In Situ Corrosion Investigation of Zr Alloys in Simulated PWR Environment by Electrochemical Measurements
Göhr H., Schaller J., Ruhmann H., Garzarolli F.
Anodic Protection Provided by Precipitates in Aqueous Corrosion of Zircaloy
Isobe T., Murai T., Mae Y.
Investigation of In-Pile Grown Corrosion Films on Zirconium-Based Alloys
Gebhardt O., Hermann A., Bart G., Blank H., Garzarolli F., Ray I.
Microstructure Evolutions and Iron Redistribution in Zircaloy Oxide Layers: Comparative Effects of Neutron Irradiation Flux and Irradiation Damages
Iltis X., Lefebvre F., Lemaignan C.
Oxide Characteristics and Corrosion and Hydrogen Uptake in Zr-2.5 Nb CANDU Pressure Tubes
Warr B., Van Der Heide P., Maguire M.
The Effect of the Trace Impurity Uranium on PWR Aqueous Corrosion of Zircaloy-4
Peters H., Harlow J.
Detrimental Role of Hydrogen on the Corrosion Rate of Zirconium Alloys
Blat M., Noel D.
Hydrogen Pickup and Redistribution in Alpha-Annealed Zircaloy-4
Kammenzind B., Franklin D., Peters H., Duffin W.
A Unified Model to Describe the Anisotropic Viscoplastic Behavior of Zircaloy-4 Cladding Tubes
Delobelle P., Robinet P., Bouffioux P., Geyer P., Le Pichon I.
The Influence of Temperature and Yield Strength on Delayed Hydride Cracking in Hydrided Zircaloy-2
Efsing P., Pettersson K.
Effects of Hydride Precipitate Localization and Neutron Fluence on the Ductility of Irradiated Zircaloy-4
Garde A., Smith G., Pirek R.
Fracture Toughness of Zircaloy Cladding Tubes
Grigoriev V., Josefsson B., Rosborg B.
A Model for Analysis of the Effect of Final Annealing on the In- and Out-of-Reactor Creep Behavior of Zircaloy Cladding
Limbäck M., Andersson T.
Properties of an Irradiated Heat-Treated Zr-2.5Nb Pressure Tube Removed From the NPD Reactor
Chow C., Coleman C., Koike M., Causey A., Ells C., Hosbons R., Sagat S., Urbanic V., Rodgers D.
Link Between Results of Small- and Large-Scale Toughness Tests on Irradiated Zr-2.5Nb Pressure Tube Material
Davies P., Shewfelt R.
Modeling In-Reactor Deformation of Zr-2.5Nb Pressure Tubes in CANDU Power Reactors
Christodoulou N., Causey A., Holt R., Tomé C., Badie N., Klassen R., Sauvé R., Woo C.
Effect of In-PWR Irradiation on Size, Structure, and Composition of Intermetallic Precipitates of Zr Alloys
Garzarolli F., Goll W., Seibold A., Ray I.
In Situ Studies of Phase Transformations in Zirconium Alloys and Compounds Under Irradiation
Motta A., Faldowski J., Howe L., Okamoto P.
Evolution of Microstructure in Zirconium Alloys During Irradiation
Griffiths M., Mecke J., Winegar J.
Influence of Neutron Irradiation on Dislocation Structure and Phase Composition of Zr-Base Alloys
Shishov V., Nikulina A., Markelov V., Peregud M., Kozlov A., Averin S., Kolbenkov S., Novoselov A.
Non-Linear Irradiation Growth of Cold-Worked Zircaloy-2
Holt R., Causey A., Christodoulou N., Griffiths M., Ho E., Woo C.
Influence of Iron in the Nucleation of ⟨c⟩ Component Dislocation Loops in Irradiated Zircaloy-4
de Carlan Y., Regnard C., Griffiths M., Gilbon D., Lemaignan C.
Effects of Extrusion-Billet Preheating on the Microstructure and Properties of Zr-2.5Nb Pressure Tube Materials
Choubey R., Aldridge S., Theaker J., Cann C., Coleman C.
Zircaloy-2 Lined Zirconium Barrier Fuel Cladding
Williams C., Marlowe M., Adamson R., Wisner S., Rand R., Armijo J.
Effects of Microstructure on Ductility and Fracture Resistance of Zr-1.2Sn-1 Nb-0.4Fe Alloy
Nikulin S., Goncharov V., Markelov V., Shishov V.
Influence of Processing Variables and Alloy Chemistry on the Corrosion Behavior of ZIRLO Nuclear Fuel Cladding
Comstock R., Schoenberger G., Sabol G.
Effects of Thermomechanical Processing on In-Reactor Corrosion and Post-Irradiation Mechanical Properties of Zircaloy-2
Huang P., Mahmood S., Adamson R.
Embrittlement of Reactor Core Materials
Kreyns P., Bourgeois W., White C., Charpentier P., Kammenzind B., Franklin D.
Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores
Nikulina A., Markelov V., Peregud M., Bibilashvili Y., Kotrekhov V., Lositsky A., Kuzmenko N., Shevnin Y., Shamardin V., Kobylyansky G., Novoselov A.
Corrosion Behavior of Duplex and Reference Cladding in NPP Grohnde
Besch O., Yagnik S., Woods K., Eucken C., Bradley E.
Development of New Zirconium Alloys for a BWR
Etoh Y., Shimada S., Yasuda T., Ikeda T., Adamson R., Chen J., Ishii Y., Takei K.
Comparison of the Long-Time Corrosion Behavior of Certain Zr Alloys in PWR, BWR, and Laboratory Tests
Garzarolli F., Broy Y., Busch R.
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
Ruhmann H., Manzel R., Sell H., Charquet D.
Development of Pressure Tubes with Service Life Greater Than 30 Years
Coleman C., Cheadle B., Cann C., Theaker J.
Author Index
Subject Index
Committee: B10
Paper ID: STP1295-EB
DOI: 10.1520/STP1295-EB
ISBN-EB: 978-0-8031-5343-1
 ASTM International is a member of CrossRef.
0-8031-2406-6
978-0-8031-2406-6
STP1295-EB
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