This work is part of a general study on the behavior of Zircaloy clad-UO₂ fuel elements during a loss-of-coolant accident to a water-cooled industrial power reactor. The fuel pins have an internal pressure due to fission gases and may or may not be exposed to steam during the thermal excursion. The energy from the nuclear heating (nuclear afterheat) plus heat from the metal-water reaction may cause the reactor core to experience rapid heat-up to about 1400 C before emergency core-cooling systems become effective in reducing the core temperatures to below the normal operating temperatures. Analysis of such an incident requires an understanding of the mode of failure and the effects of internal pressure, rate of heating, and oxidation.

The temperature of failure for Zircaloy cladding material, heated in an inert atmosphere, was determined for different heating rates for tension specimens of sheet stock under constant load and for tubes with constant internal pressure. The temperature of failure (that is, 34 percent expansion for the tubes, and rupture for the tension specimens) was found to be heating-rate sensitive with higher failure temperatures occurring at higher heating rates. An analytical expression relating applied stress, the heating rate, the true strain at failure, and temperature of failure was derived through analysis of the data from these tests. The activation energy for second-stage creep in Zircaloy-4 obtained from this analysis is 78.5 kcal/mole.

Pressurized tests in a steam atmosphere were made at pressures from 0.040 to 0.141 kg/mm² (57 to 200 psig) at heating rates from 0.1 to 300 C/s. Steam oxidation caused an increase in the temperature of failure. In addition, at low heating rates and low internal pressures the mode of failure changed from ductile failure (34 percent expansion) to brittle failure (burst). High pressures and high heating rates showed failure behavior similar to that observed in an inert atmosphere.