[en] The United States Nuclear Regulatory Commission has initiated an international sponsored severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper discusses the results of the second experiment, Test SFD 1-1. The Severe Fuel Damage Test 1-1 (SFD 1-) was designed to simulate the fuel heatup and damage and fission product release in the upper half of a 3000-MW(t) commercial pressurized water reactor during a hypothesized small break loss-of-coolant accident without emergency core coolant injection. The SFD 1-1 transient was performed by adjusting the fission power and steam flow in a 32 rod bundle of typical light water reactor 17x17 type fuel rods to produce an initial temperature increase of 0.44 K/s