Transient-temperature burst tests were performed on both unirradiated tubing and irradiated fuel rods of Zircaloy at a variety of heating rates and internal pressures. Base-line tests, performed on unirradiated boiling-water reactor size tubing over a range of initial pressures at 600°F from 50 to 1000 psig and heating rates from 10 to 100°F/sec, showed that minimum circumferential strains were obtained in the 400 to 600 psig pressure range for all heating rates. At lower and higher pressures, depending on heating rate, circumferential strains of up to 125% were found. The strain minimum was associated with rupture occurring in the two-phase α + β region of the Zircaloy as it was heated. Wall thickness variation was shown to have a large effect on the amount of strain produced. Similar tests were performed in a hot cell facility on both comparison tubing and irradiated tubing in pressurized- and boiling-water reactor sizes. Ductility minima were found in the intermediate pressure ranges of these tests, in agreement with the base-line results. No effects directly attributable to irradiation occurred in these tests. Although lower strains were found, the specific causes could not be defined because of experimental differences between the base-line and hot cell tests and the relatively low neutron exposures.