EFFECT OF SURFACE STATE AND STRESS ON AN OXIDATION OF THE ZIRCALOY-2 ALLOY

Zr alloys are widely used as materials for nuclear fuel pellets in the nuclear industry. In the case of the LOCA or RIA happen, a temperature may locally reach high values. Even if the high temperature maintains shortly, the zirconium oxides may become permeable, absorb hydrogen appearing in cooling water from decomposition reaction and crack because of formation and brittle failure of hydrides. Such model cannot so far take into account that high-temperature oxidation of Zr claddings may also result in an appearance of significant intrinsic stresses, which cause the faster descaling and cracking of the layer, thus making a hydrogen entry into the bulk more rapid. It also neglects the real surface state of commercial Zr, which may possess the surface damage and cracks. The present research was aimed at examination of an appearance and degradation of the oxide layers in the Zircaloy-2 nuclear material on round as-received specimens. The tests were performed by an oxidation of the alloy in the air at temperatures ranged between 350 and 900°C for time 0.25 or 0.5 h, corresponding to conditions of the nuclear accident. The cracks observed at relatively low temperature followed by the oxide descaling can be explained regarding already existing cracks and high intrinsic stresses appearing on the surface of small diameter rods and tubes.