RPS2

    Behavior and Properties of Zircaloys in Power Reactors: A Short Review of Pertinent Aspects in LWR Fuel

    Published: Jan 2010

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    Abstract

    Zircaloy-2 and -4, developed mainly in the United States, have been used in Germany for fuel rod claddings and in-core structural components from the beginning of reactor technology. Extensive studies of the material properties of the Zircaloys have been performed in Siemens laboratories 1957. Irradiation testing was done in several test reactors. However, the combined effects of irradiation and real environmental conditions were determined through many experimental irradiations in existing power reactors. Elaborate examinations in the reactor pools of such experimental materials and of many standard fuel rods and assemblies after intermediate, full, and intentionally extended exposures were the main source of information and hard data. These programs were supported by several ulilities and in certain areas carried out in cooperation with others.

    Zircaloy-2 and 4 turned out to be very reliable materials that fulfilled all requirements for normal operation and likewise the requirements for postulated accidental conditions and for intermediate storage for many years. Optimization of Zircaloy-2 und -4 during, recent years includes both optimization of microstructure and of chemical composition. BWRs and PWRs need differently optimized materials. Today's more demanding operation conditions and discharge burnups required a further optimization of the Zircaloys and for “hot” PWRs even the development of more corrosion-resistant Zr alloys. A significant improvement of PWR corrosion behavior can be achieved with Zr alloys using the alloying elements of Zircaloy with somewhat modified concentration. Sn should be below or at least in the lower range of the ASTM specification range far Zircaloy-4, Fe and Cr should be somewhat higher, and Si should be specified as an alloying element rather than as an impurity. The original paper was published by ASTM International in STP 1295, Zirconium in the Nuclear Industry: Eleventh International Symposium, 1996, pp. 12–32.

    Keywords:

    zirconium alloys, Zircaloy-2/4, irradiation growth, creep, corrosion, in-reactor creep, in-reactor corrosion, precipitates, iodine stress corrosion, neutron irradiation, radiation effects, nuclear application, operating conditions in LWRs


    Author Information:

    Garzarolli, F.
    Siemens AG, Erlangen,

    Stehle, H.
    Siemens AG, Erlangen,

    Steinberg, E.
    Siemens AG, Erlangen,


    Paper ID: MNL12116R

    Committee/Subcommittee: B10.02

    DOI: 10.1520/MNL12116R


    CrossRef ASTM International is a member of CrossRef.

    ISBN10:
    ISBN13: 978-0-8031-7018-6