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

摘要

The second fuel rod failure experiment in the Tran¬sient Reactor Test Facility (TREAT) was performed with a seven-rod bundle of 27-in.-long, Zircaloy-clad UO2 fuel rods in a flowing steam atmosphere. A water-reactor loss-of-coolant accident was simu¬lated by operating the TREAT reactor at constant power for 30 sec so that fission heat in the UO2 pellets caused the Zircaloy cladding temperature to rise 80°F/sec to a maximum of approximately 2400°F. The fuel rods were initially pressurized with helium to between 65 and 75 psia (77°F) to simulate accumulated fission gas.nThe Zircaloy cladding swelled and ruptured. The amount and distribution of swelling could result in the blockage of 91% of the bundle coolant channel area of a Boiling Water Reactor (BWR) at the location of maximum swelling. The average rod maximum circumferential swelling was 60%. Metallographic examination revealed ductile ruptures and significant oxygen pickup. Zirconium-steam reaction was 1.1%.