STP633: Diameter Increases in Steam Generating Heavy Water Reactor Zircaloy Cans under Loss-of-Coolant Accident Conditions

    Rose, KM
    Section leader and metallurgist, Fuel Element Laboratories, United Kingdom Atomic Energy Authority, Salwick Preston, Lancashire,

    Hindle, ED
    Section leader and metallurgist, Fuel Element Laboratories, United Kingdom Atomic Energy Authority, Salwick Preston, Lancashire,

    Pages: 12    Published: Jan 1977


    Abstract

    In the steam generating heavy water reactor (SGHWR) the fuel rod temperature rise following a loss-of-coolant accident (LOCA) is controlled by a water spray from within the rod bundle. To allow this emergency water spray to operate as intended, the fuel rod cladding diameter increase during the LOCA is kept to less than 5 percent. This gives rise to the need for a means of estimating the diameter increase of cans during temperature excursions up to 1200°C under varying internal pressure. The Water Reactor Fuel Can Swelling (CANSWEL) code has been developed for this purpose, and there are well known codes available which deal with the CANadian Deuterium Uranium (CANDU) and light water (LW) reactors. The CANSWEL code satisfactorily predicts the can diameter increase in conditions where the absorption of oxygen by zirconium does not affect the result significantly but does not deal yet with exposure to high temperature in steam. This paper describes simple equations which represent the behavior of Zircaloy-2 cans below the transformation temperature and in the β-phase in the absence of oxygen. It goes on to consider some phenomena which cause the β-phase equation to be modified when the diameter increase occurs in steam.

    Keywords:

    zirconium, zirconium alloys, heavy water reactors


    Paper ID: STP35562S

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP35562S


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