[en] Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes TRAB and SMATRA, have been examined from two points of view. First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In the study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. (60 refs., 11 figs., 4 tabs.)