Loss-of-coolant-accident (LOCA) and anticipated transient without scram (ATWS) calculations have been performed for the two Kernforschungszentrum Karlsruhe advanced pressurized water reactor reference designs (a homogeneous reactor with p/d = 1.2 and a heterogeneous reactor), for a homogeneous reactor with a tighter fuel rod lattice (p/d = 1.123), and for a reference pressurized water reactor (PWR). The calculations have been performed with the Ispra version of the code RELAP5/MOD1. New correlations have been introduced in the code to account for the core geometry, which is different from that of a PWR. The results of the calculations show that during the LOCA the fuel rod cladding hot spot temperatures in the seed of the heterogeneous reactor reach values ∼250°C higher than the corresponding temperatures for a PWR. The results also show that during the ATWS the pressure inside the primary circuit exceeds the maximum allowable pressure in the case of the homogeneous reactor with p/d = 1.123. Based on the present calculations, only the homogeneous reactor with p/d =1.2 appears to be acceptably safe. Of course, these results need experimental confirmation.