[en] An experimental program was carried out, where heat transfer and hydraulic resistance were measured in a facility that simulated a fuel assembly of a Soviet VVER-440 nuclear power reactor. The experiments were performed in a test loop with four different rod bundles that consisted of 19 full height, directly electrically heated rods. Most of the tests were conducted close to nominal conditions of the simulated reactor. Heat transfer crisis of dryout type was measured at mass fluxes from 15 to 40% of the nominal flow of the reactor, when inlet temperature of flow and heat flux were varied in the neighbourhood of the nominal reactor values. Occurrence of heat transfer crisis was independent of heat flux in the covered large steam qualities and small mass fluxes. No crisis correlation predicted this phenomenon, but the GE and the Hsu-Beckner formulas obtained on an average similar values to those of the experiments. In slow flow decay transients the crisis appeared later than in the stationary experiments. Single phase heat transfer coefficient in the rod bundles was found to be close to the Dittus-Boelter correlation. Two-phase heat transfer coefficient was smaller than expected, but the measurements were disturbed by surface fouling. Single phase friction of turbulent flow was the same as had earlier been obtained in similar conditions. Two-phase friction was measured up to large steam qualities beyond dryout in small mass fluxes. Local minimum of friction was observed around the crisis point. The Baroczy curves predict two-phase friction best of the correlations that were tested against the experimental data. (author)