RPS2: Simulating the Behavior of Zirconium-Alloy Components in Nuclear Reactors

    Coleman, Christopher E.
    Researcher Emeritus, Chalk River Laboratories, Chalk River, Ontario

    Pages: 19    Published: Jan 2010


    Abstract

    To prevent failure in nuclear components, one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: • swelling tests that led to a method for increasing the tolerance of Zircaloy fuel cladding to power ramps • observations of the behavior of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defense against flaw development • contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident The original paper was published by ASTM International in STP 1423, Zirconium in the Nuclear Industry: Thirteenth International Symposium, 2002, pp. 3–19.

    Keywords:

    simulation, zirconium alloys, fuel cladding, pressure tubes, calandria tubes, fuel swelling test, power ramp, leak-before-break, CRACLE, contact conductance, surface modification, heat sink


    Paper ID: MNL12120R

    Committee/Subcommittee: B10.02

    DOI: 10.1520/MNL12120R


    CrossRef ASTM International is a member of CrossRef.

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