One of the ways to reduce costs in modern nuclear power stations is to increase the life time of the steam generator. Much research is being carried out to gain a better understanding of the mechanical and thermal loads, which are generally overestimated in the interest of safety. One of the main technical problems is the feedwater of the reactor at a relatively cold (40 °C) temperature into the hot steam generator (270 °C). The connection zone, i.e. where the cold pipe is connected to the steam generator, is protected by an annulus inside the feedwater nozzle to reduce the thermal stresses. We have to identify the thermo-hydraulic behavior in that zone in order to accurately assess thermal information. This will give reliable boundary conditions for thermo-mechanical calculations. The configuration of the flows is a superposition of well known elementary fluid mechanic problems which interact strongly. No databases are available for this typical configuration, so we need to qualify the flow before using CFD modelisations. Therefore, a specific experimental testing bench was developed. In this paper, we focus our attention on the understanding of the flow in the one eyed cavity. PIV measurements allows us to identify the flow behavior. Our hypothesizes are demonstrated by a frequencies analysis, and finally confirmed with a numerical model. We also present and discuss the impact on a low thermal loading.
- Pressure Vessels and Piping Division
Thermo-Hydraulic Analysis in a Assembly Representative of a Feedwater Nozzle
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Boissiere, X, Laine, C, & Doubliez, L. "Thermo-Hydraulic Analysis in a Assembly Representative of a Feedwater Nozzle." Proceedings of the ASME 2003 Pressure Vessels and Piping Conference. Design and Analysis Methods and Fitness for Service Evaluations for Pressure Vessels and Components. Cleveland, Ohio, USA. July 20–24, 2003. pp. 113-119. ASME. https://doi.org/10.1115/PVP2003-1936
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