[en] The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program was to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations were tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The BWR-type bundles consisted of 18 fuel rod simulators (heated and unheated rods), an absorber blade of steel containing eleven absorber rods filled with boron carbide powder. The larger bundle CORA-18 contained the same number of absorber rods but was made up of 48 fuel rod simulators. All BWR bundles were surrounded by a zircaloy shroud and the absorber blades by a channel box wall on each side, also made of zircaloy. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence was simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. In all experiments bundle destruction started in the upper region (axially) with melting of the absorber blade and the absorber rod cladding at about 1250 C by interaction of boron carbide and steel. After destruction of the channel box walls this melt attacked the zircaloy fuel rod cladding and started to interact with the UO₂ pellets. The test bundles also resulted in severe oxidation of the following components made of zircaloy: shroud, cladding, and grid spacers at the central and upper positions. Relocated absorber melt formed extended blockages at lower elevations of the bundles. This distribution of absorber material could in the type of reactor accident described with subsequent flooding of the partially destroyed reactor core with unborated water lead to local recriticalities. There was no difference in the behavior of the large bundle CORA-18 compared to the BWR test bundles of regular size, i.e. CORA-16 and CORA-17. Quenching (flooding) of a degraded BWR-type bundle (CORA-17) exhibits identical behavior as observed in the PWR-type quenching experiments: The bundle results in locally enhanced zircaloy/steam reaction causing a renewed temperature rise, an additional meltdown of materials, and an additional strong hydrogen generation. (orig.)