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The first fuel rod failure experiment in Trrjisient Reactor Test Facility (TREAT) was performed with a sevenrod bundle of 27 in. long Zircaloy-clad U0 2 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 uu2 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. The 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%. The center rod was preirradiated to 650 Mwd/MT in the Materials Testing Reactor (MTR) to build up an inventory of fission products and to determine irradia­ tion eiiects on fuel rod failure characteristics. No irradiation effect was seen from this low-level irradia­ tion on the swelling and rupture characteristics. A form of fuel rod failure propagation apparently occurred when chips of fragmented U0 2 from the rupture hole in the center irradiated rod contacted one of the outer rods resulting in stress concentrations and rupture of the outer rod at the point of contact. The release of gaseous fission products 1 3 3 X e and 8 Kr from the irradiated center rod was in the 0.1% to 0.2% range. The release of volatile fission products *MetaIs and Ceramics Division.

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