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.
Zircaloy Performance in Light-Water Reactors
Gelhaus F., Roberts J.
Ductility of Zircaloy Canning Tubes in Relation to Stress Ratio in Biaxial Testing
Andersson T., Wilson A.
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.
Effect of Irradiation on the Strength, Ductility, and Defect Sensitivity of Fully Recrystallized Zircaloy Tube
Andersson T., Pettersson K., Vesterlund G.
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.
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.
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.
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.
Tube-Burst Response of Irradiated Zircaloy Spent-Fuel Cladding
Bauer A., Gallagher W., Lowry L.
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.
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.
Paper ID: STP681-EB