[en] VTT Technical Research Centre of Finland took part in the recently ended European HPLWR2- project, in which a SCWR concept called the High Performance Light Water Reactor (HPLWR) was studied. VTT participated in the project by calculating preliminary safety analyses using two simulation tools, APROS and TRAB-3D/SMABRE, both of which were developed in-house. In order to support simulation of thermal hydraulics at supercritical pressures, these codes were modified by adding new constitutive equations suitable for the supercritical-pressures conditions, extending and refining the steam tables used to calculate the thermophysical properties of water, and by developing a numerical scheme which handles the liquid-to-supercritical-to-gas phase transition in a sound manner. In this paper, an analysis of a large-break loss of coolant accident in the High Performance Light Water Reactor, calculated with APROS, is presented. The simulation is calculated with the one- dimensional separate two-fluid (6-equation) thermal-hydraulic flow model, and the neutronic behavior of the nuclear reactor is modeled using a three-dimensional neutron diffusion model. The simulation model, which represents the reactor concept with a three-pass core, includes the reactor pressure vessel, and a part of the steam line until the main steam line isolation valve. In the neutronic model, each fuel assembly cluster is modeled separately, but on the thermal hydraulic side only three core-flow channels, each corresponding to a single pass through the reactor core, are modeled due to computation time requirements. The accident is initiated by a 1 x 100 % break in one of the four steam lines, between the pressure vessel outlet and the main steam line isolation valve. Decreasing pressure at the pressure vessel outlet initiates the reactor scram and closure of the isolation valves. Reactor inlet flow is kept constant until the feed-water tank runs empty. After the closure of the main steam line isolation valves, all the water flowing out of the pressure vessel runs through the break orifice into the containment. Low head safety injection, an active safety component, starts to inject cold water at the reactor inlet as the pressure has decreased sufficiently. The presented analysis suggests that the reactor core of the HPLWR can be kept sufficiently cooled-down in the case of a large break LOCA in the main steam line using the specified safety systems. Some uncertainty to the simulation results is caused by the constitutive equations used at supercritical pressures, but they have very limited effect on the overall results in a simulation case, where the pressure drops to subcritical conditions at a very early stage. Also the current model with only three thermal hydraulic core channels is sufficient for this kind of simulation, where the reactor is brought to decay heat very early on to the simulation. Proper modeling of flows in each fuel cluster would be needed for more elaborate analyses, and for example analyses of reactivity initiated accidents. (author)