Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydraulic system codes. These coupled codes can be used for the analysis of the whole reactor system. In the framework of the international association Atomic Energy Research (AER) on VVER Reactor Physics and Reactor Safety, two benchmarks for these code systems were defined. The reference reactor is the Russian VVER-440. The response of the reactor core to a symmetric and an asymmetric main steam line break should be investigated. So, different aspects of the coupling could be tested. As an additional feature, the participants had to use their own nuclear data. Each of these benchmarks was calculated by five different code systems. The comparison of the received solutions for the symmetric case shows good agreement in the evolution of the thermal hydraulics. When the core power reestablishes after recriticality, differences between the single solutions develop, mainly connected with the use of different nuclear data. Because of the increased complexity of the calculations, in the second benchmark differences between the thermal-hydraulic behavior in the single calculations were observed, additionally. These differences have their main origin in the behavior of the secondary side. The results of both benchmarks show the safety potential of the VVER-440 reactor. Even under very conservative conditions, no fuel rod failure was determined by the calculations, and the reactor was transferred into a subcritical final state.