During the normal cycle of a pressurized water reactor, boron concentration is reduced in the core until fuel burns up. A stretch out of the normal cycle is however possible afterwards, provided primary coolant temperature is reduced. In those stretch out periods, nuclear operators want to keep constant thermal power exchanged in the steam generator, in order to preserve its performances. Under that constraint, the required reduction in primary coolant temperature involves both a decrease of secondary cooling system pressure and an increase of tube bundle vibrations. Since neither pressure nor vibrations should exceed some given thresholds in order to preserve component integrity, the reduction of primary coolant temperature has to be limited. Nuclear plant operators thereafter need an operating diagram, i.e. a diagram that provides minimum allowed primary coolant temperature versus power rate. In that context, we propose a method to derive such a diagram, by combining, on the one hand a code for simulating primary and secondary fluid flows in steam generators and, on the other hand, a software that allows one to predict fluid elastic tube bundle instabilities. That method allows one to take into account both tube fouling and plugging. It is now used by French utility “Electricite´ De France”, in order to check or supplement the analysis that are provided by steam generator manufacturers.
- Pressure Vessels and Piping Division
Influence of Steam Generator Tube Bundle Vibrations on the Operating Diagram of a Nuclear Plant During Stretch Out
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Adobes, A, Pillet, J, David, F, & Gaudin, M. "Influence of Steam Generator Tube Bundle Vibrations on the Operating Diagram of a Nuclear Plant During Stretch Out." Proceedings of the ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference. Volume 4: Fluid Structure Interaction, Parts A and B. Vancouver, BC, Canada. July 23–27, 2006. pp. 147-156. ASME. https://doi.org/10.1115/PVP2006-ICPVT-11-93239
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