FINAL REPORT ON THE FIRST FUEL ROD FAILURE TRANSIENT TEST OF A ZIRCALOY-CLAD FUEL ROD CLUSTER IN TREAT

摘要

The first fuel rod failure experiment in Transient Reactor Test Facility (TREAT) was performed with a seven-rod bundle of 27 in. long Zircaloy-clad U02 fuel rods in flowing steam atmosphere. A water reactor loss-of-coolant accident was simulated by operating TREAT reactor at con-stant power for 20 sec so that fission heat in the U02 pellets caused the Zircaloy cladding temperature to rise 72°F/sec to a maximum of approximately 1800°F. The fuel rods were initially pressurized with helium between 115 and 215 psia (77°F) to simulate accumulated fission gas.nThe Zircaloy cladding swelled and ruptured resulting in 48% blockage of the bundle coolant channel area at the location of maximum swelling. The average rod maximum circumferential swelling was 36%. Calculations related the hoop stress and ultimate strength at the onset of rapid expansion. The ideal gas law was used to calculate the rate of cladding expansion from measured rod tempera¬ture and internal pressure. Metallographic examination revealed ductile ruptures and significant oxygen pickup. Zirconium-steam reaction was 0.2%.