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    An Experimental Study of the Deformation of Zircaloy PWR Fuel Rod Cladding Under Mainly Convective Cooling

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    In a loss-of-coolant accident (LOCA) the temperature of the cladding of some fuel rods may rise to levels [650 to 810°C (1202 to 1490°F)] where the ductility of Zircaloy is high (∼80 percent). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of pressurized water reactor (PWR) Zircaloy cladding tubing 450 mm (18 in.) long under internal pressure had shown that strains of over 50 percent could be produced over considerable lengths (greater than 20 tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel assembly. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss, believed to be more representative of a fuel assembly following a loss of coolant.

    Zircaloy-4 cladding specimens 450 mm (18 in.) long were filled with alumina pellets and tested at temperatures between 630 and 915°C (1166 and 1679°F) in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa (100 to 1600 lb/in.2). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the stream flow.

    The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38 ± 0.17 MPa (200 ± 25 lb/in.2). It is though that this results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa (145 lb/in.2). These results are compared with those from multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a LOCA are outlined.


    zirconium, nuclear industry, loss of coolant, pressurized water reactors, fuel cans, deformation, Zircaloy-4, distribution, swelling, very high temperature, convective heat transfer, strain anisotropy, superplastic deformation

    Author Information:

    Hindle, ED
    Principal Scientific Officers, United Kingdom Atomic Energy Authority, Springfields Nuclear Laboratories, Preston, Lancs

    Mann, CA
    Principal Scientific Officers, United Kingdom Atomic Energy Authority, Springfields Nuclear Laboratories, Preston, Lancs

    Committee/Subcommittee: B10.02

    DOI: 10.1520/STP37059S