STP754

    Zirconium in the Nuclear Industry

    Franklin DG

    Pages: 477

    Published: 1982

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    This publication provides information on the performance of zirconium and zirconium alloy components in nuclear application. Nuclear physicists have discovered that some materials -- zirconium, for example, provide much better neutron economy than conventional materials such as stainless steels.


    Table of Contents

    Introduction
    Franklin D.

    Discussion

    Summary
    Franklin D.

    A Review of Texture and Texture Formation in Zircaloy Tubing
    Tenckhoff E.

    Texture Control of Zircaloy Tubing During Tube Reduction
    Fujita K., Kakuma T., Nagai N.

    Texture Measurement Techniques for Zircaloy Cladding: A Round-Robin Study
    Adamson R., Lewis J., Schoenberger G.

    Application of Hydrostatic Extrusion to Fabrication of Zircaloy Tubing
    Kakuma T., Kimpara M., Matsushita T., Nagai N.

    Beta-Quenching of Zircaloy Cladding Tubes in Intermediate or Final Size—Methods to Improve Corrosion and Mechanical Properties
    Andersson T., Vesterlund G.

    New Intermetallic Compounds in Zircaloy-4
    Charquet D., Grange J., Moulin L.

    Long-Term Test Results of Promising New Zirconium Alloys
    Castaldelli L., Fizzotti C., Lunde L.

    Clad Failure Modeling
    Ballinger R., Christensen R., Eilbert R., Oldberg S., Rumble E., Was G.

    Chemical Aspects of Iodine-Induced Stress Corrosion Cracking of Zircaloys
    Cubicciotti D., Jones R., Syrett B.

    Effect of Zirconium Oxide on the Stress-Corrosion Susceptibility of Irradiated Zircaloy Cladding
    Mattas R., Neimark L., Yaggee F.

    Effects of Temperature and Pressure on the In-Reactor Creepdown of Zircaloy Fuel Cladding
    Dodd C., Hobson D., Thoms K., van der Kaa T.

    High-Strength, Creep-Resistant Excel Pressure Tubes
    Causey A., Cheadle B., Fidleris V., Holt R., Urbanic V.

    High-Temperature Irradiation Growth in Zircaloy
    Adamson R., Fidleris V., Tucker R.

    Zircaloy-4 Cladding Deformation During Power Reactor Irradiation
    Franklin D.

    Burst Criterion of Zircaloy Fuel Claddings in a Loss-of-Coolant Accident
    Erbacher F., Neitzel H., Rosinger H., Schmidt H., Wiehr K.

    An Experimental Study of the Deformation of Zircaloy PWR Fuel Rod Cladding Under Mainly Convective Cooling
    Hindle E., Mann C.

    Effect of Hydrogen on the Oxygen Embrittlement of Beta-Quenched Zircaloy-4 Fuel Cladding
    Seiffert S.

    Lifetime and Failure Strain Prediction for Material Subjected to Non-stationary Tensile Loading Conditions; Applications to Zircaloy-4
    Boček M.

    Flow Stress of Oxygen-Enriched Zircaloy-2 Between 1023 and 1873 K
    Choubey R., Ells C., Holt R., Jonas J.

    A Comparison of the High-Temperature Oxidation Behavior of Zircaloy-4 and Pure Zirconium
    Campbell J., Pawel R.

    Effect of Texture on Hydride Reorientation and Delayed Hydrogen Cracking in Cold-Worked Zr-2.5Nb
    Coleman C.

    Mechanism of Accelerated Corrosion in Zircaloy-4 Laser and Electron-Beam Welds
    McDonald S., Sabol G.

    Waterside Corrosion of Zircaloy-Clad Fuel Rods in a PWR Environment
    Garzarolli F., Jorde D., Manzel R., Politano J., Smerd P.

    Long-Term In-Reactor Corrosion and Hydriding of Zircaloy-2 Tubing
    Hillner E.

    Index


    Committee: B10

    Paper ID: STP754-EB

    DOI: 10.1520/STP754-EB

    ISBN-EB: 978-0-8031-4823-9

    ISBN-PRINT: 978-0-8031-0754-0

    ASTM International is a member of CrossRef.

    0-8031-0754-4
    978-0-8031-0754-0
    STP754-EB