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SEDL / STP / STP681-EB
The Fourth International Conference on Zirconium in the Nuclear Industry was held 26-29 June 1978, in Stratford-upon-Avon, England. This conference was sponsored by the American Society for Testing and Materials (ASTM) Committee B10 on Refractory Metals and Alloys in cooperation with the American Nuclear Society, the British Nuclear Energy Society, and The Metals Society (U.K).. The program was planned by representatives of ASTM Subcommittee B10.02 on Zirconium and Hafnium and K. G. Sumner of British Nuclear Fuels, Ltd. J. H. Schemel, Sandvik Special Metals Corp., and K. G. Sumner, BNFL, served as the conference co-chairmen, T. P. Papazoglou, The Babcock & Wilcox Company, as editorial chairman, and H. M. Cobb, ASTM Staff, as conference coordinator.
Table of Contents
Introduction
Papazoglou T.
Zirconium Technology—Twenty Years of Evolution
Lustman B.
Zircaloy Performance in Light-Water Reactors
Gelhaus F., Roberts J.
Effect of Material and Environmental Variables on Localized Corrosion of Zirconium Alloys
Lunde L., Videm K.
Ductility of Zircaloy Canning Tubes in Relation to Stress Ratio in Biaxial Testing
Andersson T., Wilson A.
Localized Ductility Method for Evaluating Zircaloy-2 Cladding
Coffin L.
Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions
Garzarolli F., Manzel R., Reschke S., Tenckhoff E.
Further Evidence of Zircaloy Corrosion in Fuel Elements Irradiated in a Steam Generating Heavy Water Reactor
Garlick A., Hartog J., Sims P., Stuttard A., Sumerling R., Trowse F.
Performance of Irradiated Copper and Zirconium Barrier-Modified Zircaloy Cladding Under Simulated Pellet-Cladding Interaction Conditions
Adamson R., Gangloff> R., Tomalin D.
A Survey of the Chemical Environments for Activity in the Embrittlement of Zircaloy-2
Grubb W., Morgan M.
Effect of Irradiation on the Strength, Ductility, and Defect Sensitivity of Fully Recrystallized Zircaloy Tube
Andersson T., Pettersson K., Vesterlund G.
Primary Creep of Zircaloy-2 Under Irradiation
Fidleris V.
Analysis of Irradiation Growth and Multiaxial Deformation Behavior of Nuclear Fuel Cladding
Adams B., Clevinger G., Murty K.
Analysis of Pressurized Water Reactor Fuel Pin Length Changes
Cornell R., Harbottle J.
Irradiation Growth in Cold-Worked Zircaloy-2
Murgatroyd R., Rogerson A.
Stress Corrosion Crack Initiation and Growth and Formation of Pellet-Clad Interaction Defects
Lunde L., Videm K.
Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon
Peehs M., Stehle H., Steinberg E.
Iodine-Induced Stress Corrosion Cracking of Fixed Deflection Stressed Slotted Rings of Zircaloy Fuel Cladding
Sejnoha R., Wood J.
A Stress Corrosion Cracking Model for Pellet-Cladding Interaction Failures in Light-Water Reactor Fuel Rods
Cubicciotti D., Jones R., Miller A., Roberts J., Smith E., Wachob H., Yaggee F.
Hydride Cracks as Initiators for Stress Corrosion Cracking of Zircaloys
Cox B.
Predicting High-Temperature Transient Deformation from Microstructural Models
Holt R., Sills H.
A Microstructure-Based Constitutive Relation for Dilute Alloys of α-Zirconium
Chung H., Garde A., Hartley C., Kassner T., Lee J.
A Comparison of Experimental Cladding Microstructure Resulting from Uranium Oxide-Zircaloy Interaction with a Diffusional Assessment of Oxygen Transport for a Coupled Two-Media Problem
Cronenberg A., El-Genk M.
Advances in Understanding and Predicting Inelastic Deformation in Zircaloy
Lucas G., Miller A., Oldberg S.
Zirconium Cladding Deformation in a Steam Environment with Transient Heating
Chapman R., Crowley J., Hofmann G., Longest A.
Influence of Iodine on the Strain and Rupture Behavior of Zircaloy-4 Cladding Tubes at High Temperatures
Hofmann P.
Studies on Zircaloy Fuel Clad Ballooning in a Loss-of-Coolant Accident—Results of Burst Tests with Indirectly Heated Fuel Rod Simulators
Erbacher F., Neitzel H., Wiehr K.
Effect of β-Phase Heat Treatment on the Subsequent α-Phase Ballooning Behavior of Zircaloy-4 Fuel Sheaths
Hunt C., Newell W.
Tube-Burst Response of Irradiated Zircaloy Spent-Fuel Cladding
Bauer A., Gallagher W., Lowry L.
A Proposed Criterion for the Oxygen Embrittlement of Zircaloy-4 Fuel Cladding
Sawatzky A.
Effect of Oxygen on the Deformation of Zircaloy-2 at Elevated Temperatures
Holt R., Jonas J., Rizkalla A.
Evaluation Models of Zircaloy Oxidation in Light of Recent Experiments
Biederman R., Hann C., Ocken H., Westerman R.
Chemical Interaction Between Uranium Oxide and Zircaloy-4 in the Temperature Range Between 900 and 1500°C
Hofmann P., Politis C.
Interaction of Oxidation and Creep in Zircaloy-2
Burton B., Donaldson A., Reynolds G.
Embrittlement of Zircaloy-Clad Fuel Rods Irradiated Under Film Boiling Conditions
Hobbins R., MacDonald P., Mehner A., Ploger S., Seiffert S.
Development of an Oxygen Embrittlement Criterion for Zircaloy Cladding Applicable to Loss-of-Coolant Accident Conditions in Light-Water Reactors
Chung H., Garde A., Kassner T.
Committee: B10
Paper ID: STP681-EB
DOI: 10.1520/STP681-EB
ISBN-EB: 978-0-8031-4749-2
 ASTM International is a member of CrossRef.
0-8031-0601-7
978-0-8031-0601-7
STP681-EB
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