For safety and economic reasons, one of the greatest concerns in modern nuclear power stations is to increase the service life of the steam generator. Enormous work has been undertaken to obtain better knowledge of the mechanical and thermal loads, which are generally overestimated to ensure maximum safety. One of the technical problems is that the reactor is fed with relatively cold (40 °C) water into the hot steam generator. The connection zone, i.e. where the cold water pipe is connected to the steam generator, is protected by an annular cavity inside the feed water nozzle to reduce thermal stresses. We have to identify the thermo-hydraulic behaviour in that zone in order to get more accurate thermal information for the boundary condition in the thermo-mechanical calculations for this connection. No database is available for this configuration model, so we need to qualify the flow before using CFD modeling. Therefore a specific experimental testing bench along with numerical modeling has been developed. The first requirement is that the flow in the annular cavity is representative of the actual one. For this purpose, we used Reynolds and Strouhal similitudes to set up the flow in the model. Attention has been focused on the understanding of the flow in the annular cavity. PIV measurements enable us to have a good understanding of the flow structure. We also present and discuss the results in terms of velocity profiles. The main result is that the flow is unsteady in the cavity and depends on the length and thickness of the annular cavity. Experimental results give us data to validate our numerical approach, which is used to test several configurations.

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