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Digital Library / STP / STP1354-EB
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Zirconium in the Nuclear Industry: Twelfth International Symposium
Sabol GP, Moan GD
Pages: 943
Published: 2000
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44 papers discuss the history, processing, use, and properties of Zirconium and its alloys. Nearly half are concerned with the corrosion and hydriding of different Zirconium alloys. The techniques used to characterize the oxides and their microstructures and properties have become very advanced. The industry, including producers and users, expects that the work will lead to important conclusions about alloy development, optimization of the processing parameters, and station operating procedures so that the current and new Zirconium materials will continue to operate reliably to very high neutron fluences. Seven major topics include: • Properties at High Neutron Fluences • Hydrogen and Temperature Effects • Deformation and Fracture Studies • Processing and Alloy Development • Effect of Composition and Microstructure on Corrosion • Corrosion Simulation and the Effect of the Environment • Effects of Oxides on Oxidation and H Pickup Rates Audience:Scientists and engineers involved in the production and properties of components for use in nuclear reactors, their behavior during service and the property changes that occur with increasing neutron fluences.
Table of Contents
Overview PDF
Microstructure and Properties of Zirconium Alloys in the Absence of Irradiation
Charquet D.
Effects of Neutron Irradiation on Microstructure and Properties of Zircaloy
Adamson R.
Understanding Hydrogen in Zirconium
Sawatzky A., Ells C.
Irradiation Creep and Growth Behavior, and Microstructural Evolution of Advanced Zr-Base Alloys
Gilbon D., Soniak A., Doriot S., Mardon J.
Irradiation-Enhanced Deformation of Zr-2.5Nb Tubes at High Neutron Fluences
Causey A., Holt R., Christodoulou N., Ho E.
High-Fluence Irradiation Growth of Cold-Worked Zr-2.5Nb
Holt R., Causey A., Griffiths M., Ho E.
Evolution of Dislocation and Precipitate Structure in Zr Alloys Under Long-Term Irradiation
Averin S., Panchenko V., Kozlov A., Sinelnikov L., Shishov V., Nikulina A.
Effect of Long-Term Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes
Hosbons R., Davies P., Griffiths M., Sagat S., Coleman C.
Post-Irradiation Characterization of Ultra-High-Fluence Zircaloy-2 Plate
Mahmood S., Farkas D., Adamson R., Etoh Y.
The Terminal Solid Solubility of Hydrogen in Zirconium Alloys
McMinn A., Darby E., Schofield J.
The Long-Range Migration of Hydrogen Through Zircaloy in Response to Tensile and Compressive Stress Gradients
Kammenzind B., Berquist B., Bajaj R., Kreyns P., Franklin D.
Influence of Cladding Microstructure on the Low Enthalpy Failures in RIA Simulation Tests
Garde A.
Experiment and Modeling of Advanced Fuel Rod Cladding Behavior Under LOCA Conditions: Alpha-Beta Phase Transformation Kinetics and EDGAR Methodology
Forgeron T., Brachet J., Barcelo F., Castaing A., Hivroz J., Mardon J., Bernaudat C.
Behavior of Irradiated Zircaloy-4 Fuel Cladding Under Simulated LOCA Conditions
Ozawa M., Takahashi T., Homma T., Goto K.
Influence of Texture on the Fracture Toughness of Unirradiated Zircaloy Cladding
Grigoriev V., Pettersson K., Andersson S.
Degraded Fuel Cladding Fractography and Fracture Behavior
Edsinger K., Davies J., Adamson R.
Delayed Hydride Cracking in Irradiated Zircaloy Cladding
Efsing P., Pettersson K.
Size, Geometry, and Material Effects in Fracture Toughness Testing of Irradiated Zr-2.5Nb Pressure Tube Material
Davies P., Shewfelt R.
Failure Mechanisms of Irradiated Zr Alloys Related to PCI: Activated Slip Systems, Localized Strains, and Iodine-Induced Stress Corrosion Cracking
Fregonese M., Régnard C., Rouillon L., Magnin T., Lefebvre F., Lemaignan C.
Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes
Bouffioux P., Rupa N.
A New Fabrication Process for Zr-Lined Zircaloy-2 Tubing
Abe H., Takeda K., Uehira A., Anada H., Furugen M.
Zirconium Alloy Cold Pilgering Process Control by Modeling
Aubin J., Montmitonnet P., Mulot S.
In-Process Investigation of Precipitate Growth in Zirconium Alloys
Herb B., Ruhmann H., König A.
Influence of Composition and Fabrication Process on Out-of-Pile and In-Pile Properties of M5 Alloy
Mardon J., Charquet D., Senevat J.
Effect of Dilute Alloy Additions of Molybdenum, Niobium, and Vanadium on Zirconium Corrosion
Sabol G., Comstock R., Nayak U.
E110 Alloy Cladding Tube Properties and Their Interrelation with Alloy Structure-Phase Condition and Impurity Content
Shebaldov P., Peregud M., Nikulina A., Bibilashvili Y., Lositski A., Kuz'menko N., Belov V., Novoselov A.
Contribution to a Better Understanding of the Detrimental Role of Hydrogen on the Corrosion Rate of Zircaloy-4 Cladding Materials
Blat M., Legras L., Noel D., Amanrich H.
Mechanism of Corrosion Rate Degradation Due to Tin
Takeda K., Anada H.
Influence of Transition Elements Fe, Cr, and V on Long-Time Corrosion in PWRs
Broy Y., Garzarolli F., Seibold A., Van Swam L.
Fundamental Study on the Corrosion Mechanism of Zr-0.2Fe, Zr-0.2Cr, and Zr-0.1Fe-0.1Cr Alloys
Murai T., Isobe T., Takizawa Y., Mae Y.
Microstructural Aspects of Corrosion and Hydrogen Ingress in Zr-2.5Nb
Urbanic V., Griffiths M.
The Effect of Microstructure on the Corrosion Behavior of Zircaloy-2 in BWRs
Etoh Y., Nonaka Y., Kubo T., Nomata T., Hayashi H., Kitamura M.
Impact of Second Phase Particles on BWR Zr-2 Corrosion and Hydriding Performance
Rudling P., Wikmark G., Lehtinen B., Pettersson H.
Corrosion of Electron-Irradiated Zr-2.5Nb and Zircaloy-2
Woo O., McDougall G., Hutcheon R., Urbanic V., Griffiths M., Coleman C.
Evaluation of Zircaloy-2 Cladding Corrosion Characteristics by Simulated BWR Corrosion Loop Test
Shimada S., Asahi K., Nonaka Y., Sasaki M., Kogai T., Hayashi H., Kitamura M., Sakamoto M., Yamawaki M.
Studies of Zirconium Alloy Corrosion and Hydrogen Uptake During Irradiation
McDougall G., Urbanic V., Aarrestad O.
Behavior of Lithium and Boron in Irradiated and Unirradiated Oxides Formed on Zircaloy-4 Claddings
Kido T., Wada S., Takahashi T., Uchida H., Komine I., Inoue Y.
Contribution to the Understanding of the Effect of the Water Chemistry on the Oxidation Kinetics of Zircaloy-4 Cladding
Pêcheur D., Godlewski J., Peybernès J., Fayette L., Noé M., Frichet A., Kerrec O.
Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance
Nanikawa S., Etoh Y., Shimada S., Kubo T., Ito K., Harada H.
Electrochemical Examinations in 350°C Water with Respect to the Mechanism of Corrosion-Hydrogen Pickup
Baur K., Garzarolli F., Ruhmann H., Sell H.
Hydrogen Transport in the Oxide and Hydrogen Pickup by the Metal During Out- and In-Reactor Corrosion of Zr-2.5Nb Pressure Tube Material
Ramasubramanian N., Perovic V., Leger M.
Stress Distribution Measured by Raman Spectroscopy in Zirconia Films Formed by Oxidation of Zr-Based Alloys
Godlewski J., Bouvier P., Lucazeau G., Fayette L.
Role of Intermetallic Precipitates in Hydrogen Transport through Oxide Films on Zircaloy
Hatano Y., Sugisaki M., Kitano K., Hayashi M.
Multi-Scale Characterization of the Metal-Oxide Interface of Zirconium Alloys
Bossis P., Lelièvre G., Barberis P., Iltis X., Lefebvre F.
Author Index
Subject Index
Committee: B10
Paper ID: STP1354-EB
DOI: 10.1520/STP1354-EB
ISBN-EB: 978-0-8031-5416-2
 ASTM International is a member of CrossRef.
0-8031-2499-6
978-0-8031-2499-8
STP1354-EB
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