[en] The PMK-2 experimental facility at the KFKI-AEKI, Budapest, is a full pressure, scaled down model of the primary and partly the secondary circuit of the Paks Nuclear Power Plant. This NPP is equipped with four VVER-440/213-type reactors. Such plants are slightly different from PWRs of usual design and have a number of special features as 6-loop primary circuit, horizontal steam generators, loop seal in hot and cold legs, setpoint pressure of passive safety injection tanks (SIT) higher than secondary pressure, etc. The PMK-2 was primarily designed for investigating operational and off-normal transient processes, as well as small-break loss of coolant accidents of Paks NPP. The volume and power scaling ratios are 1:2070. Due to the importance of gravitational forces in both single- and two-phase flow the elevation ratio is 1:1 except for the lower plenum and pressuriser. The six loops of the plant are modelled by a single active loop. Transients can be started from nominal operating conditions. The pressuriser (PRZ) is connected to the lower part of the hot leg as in the reference system. The core model consists of 19 electrically heated rods. The main circulating pump of the PMK-2 serves to produce the nominal operating conditions and to simulate the flow coast-down following pump trip. The horizontal design of the VVER-440 steam generator is modelled by horizontal heat transfer tubes between hot and cold vertical collectors in the primary side. The emergency core cooling systems including the SITs. High and low pressure injection systems of the Paks NPP are also modelled. The first design of the PMK-NVH facility only modelled the primary circuit of plant. This version was used until 1990. The PMK-2 facility is an upgraded version (first of all by addition of a controlled secondary heat removal system) extending the capability of the test loop to modelling transient processes evoked by initiating events in the secondary circuit or including accident sequences in support of accident management (AM) procedures. During the 15 operational years - from May 1986 onwards with the first of four IAEA Standard Problem Exercise tests - 48 different experiments, including cold and hot leg break LOCA, primary-to-secondary leakage (PRISE), loss of flow, loss of feedwater, disturbances of natural circulation, etc. tests were performed on this integral type test facility. An overview on 11 experiments related to the operational behaviour of horizontal steam generators performed in framework of national research projects IAEA Technical Co-operation Project RER/9/004 (Standard Problem Exercises) and three EU PHARE projects (in co-operation with AEAT, FRAMATOM, SIEMENS, IPSN, GRS, FZR and VVER owner countries) is given in the first part of paper. In the second part results of two types of tests in shutdown condition with RELAP5 post-test analysis may be of interest to the computer simulation of the horizontal SG too - are summarised. (orig.)