You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.


    The Corrosion Behavior of 2.5Nb Zirconium Alloy

    Published: 0

      Format Pages Price  
    PDF (272K) 15 $25   ADD TO CART
    Complete Source PDF (7.1M) 376 $105   ADD TO CART


    The heat-treatable 2.5Nb zirconium alloy is of interest to nuclear-reactor designers because it has greater strength than the Zircaloys. Its corrosion behavior is affected by its metallurgical structure, the chemical composition of the environment, and the presence of reactor radiation.

    Desirable mechanical properties are obtained by quenching from a temperature high in the (α + β)-phase region (generally 850 to 880 C) and stabilizing the structure by aging in the α-phase region (500 to 575 C). Cold work between quenching and aging enhances some properties, notably corrosion resistance. This is illustrated by results of corrosion experiments in aqueous and gaseous media. The corrosion resistance after different amounts of cold work and aging may be correlated with the sub-microstructure developed, as shown by electron microscopy. Evidence from several independent experiments shows that the alloy is much less sensitive to radiation enhancement of corrosion than is Zircaloy, even under strongly oxidizing conditions, such as neutral boiling water. In many cases the oxidation in-flux is less than that out-of-flux in the same coolant stream.

    The 2.5Nb zirconium alloy has found nuclear application as pressure tubes for heavy-water moderated reactors, both with pressurized water and boiling water coolant, and as fuel cladding in organic coolant. Results of irradiation experiments suggest it would be useful as fuel cladding in boiling-water-cooled reactors and for conditions which exceed the upper temperature limit of Zircaloy cladding.


    zirconium alloys, corrosion, environmental tests, nuclear reactor, irradiation effects, mechanical properties, tests, evaluation

    Author Information:

    LeSurf, J. E.
    Leader, Corrosion Group, Fuels and Materials Division, Applied Materials Research Branch, Chalk River Nuclear Laboratoryies, Atomic Energy of Canada Limited, Chalk River, Ont.

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP43834S