A burst criterion model has been developed to calculate the burst data of Zircaloy claddings of a pressurized water reactor in a loss-of-coolant accident. It is assumed by this model that the time of burst is reached when the local stress equals the limiting burst stress. The burst stress is assumed to depend on the temperature and oxygen concentration of the Zircaloy.
With the histories of temperature and pressure known, the strain, stress, and oxygen content can be calculated by integration of the creep equation and correlation of the oxidation kinetics. Once the time of burst has been calculated, all burst data, that is, strain, stress, temperature, pressure, and oxygen content, are determined.
To verify the burst criterion model, single-rod transient burst tests in steam were performed using fuel rod simulators with indirect electric heating. The test parameters covered the following ranges: internal overpressure—10 to 140 bar, heating rate—1 to 30 K/s.
The burst temperatures and burst strains predicted by the burst criterion model are in good agreement with the test data.
Under severe fuel-damage (SFD) conditions a combined external and internal oxidation of the fuel-rod cladding occurs due to a reaction with steam or oxygen on the outside surface and the UO2 fuel on the inside surface. These reactions result in the formation of oxygenstabilized α-Zr(O) phases, ZrO2, and a (U,Zr) alloy. The reaction kinetics have been studied with isothermal and transient temperature experiments above 900°C. The tests were performed using short LWR fuel-rod sections in an (Ar + 25 vol% O2) environment. The numerical model PECLOX, which solves the Fick and Stefan equations, has been developed. It predicts the formation, growth, and disappearance of the various interaction layers and the corresponding oxygen profiles as functions of temperature and time.
Burst tests with indirectly heated fuel rod simulators were performed under cooling and temperature conditions postulated for a loss-of-coolant accident (LOCA) of a pressurized water reactor (PWR). The tests were designed to investigate the ballooning behavior of Zircaloy fuel rod claddings and to determine the resulting coolant channel blockage in a rod bundle.
The deformation behavior of the Zircaloy cladding tubes was found to be influenced mainly by local effects due to nonuniform temperature distributions. Single rod and bundle tests exhibited large azimuthal and axial temperature differences. This resulted in relatively small and axially limited strains of the Zircaloy cladding tubes and a tolerable coolant channel blockage in the rod bundle.
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