International audienceIn this paper, the recent efforts for the qualification of the thermal-hydraulic system code CATHAREfor nuclear research reactors are presented. Even if the code has been extensively validated forcommercial pressurized water reactors, additional work was needed due to the unique design andoperating conditions of research reactors. In fact, in the core region, the coolant usually flows at lowpressures in narrow channels which are arranged in a parallel configuration. Such an arrangement maybe subject to the flow excursion instability that can lead to boiling crisis in some of the channels.The study is based on a database of flow excursion experiments in vertical narrow rectangular channelswith gap sizes between 1.4 and 3.23 mm. The thermal-hydraulic conditions are representative of theones in research reactors and the coolant flows either up- or down-ward. The experimental proceduresof these experiments consist in either reducing the mass flux or increasing the heat flux to the channel(keeping all other parameters constant) until the flow excursion occurs.The system code CATHARE is used to simulate the test sections and the experimental procedures arereproduced. The available pressure drops measured over the heated channel are compared to thesimulations and good predictions are obtained both in single- and in two-phase conditions. It is alsoshown that the whole flow excursion phenomenon and, in particular, the mass flux or the heat flux atwhich the onset of flow excursion instability occurs can be reproduced by CATHARE in a satisfactoryway
