Diverse boiling phenomena occur during the operation of light-water reactors. Their understanding is necessary to guarantee a safe service and to avoid unstable operating modes. For example, the comportment of the coolant could either be subcooled boiling during normal operation or even critical boiling during the occurrence of a disturbance. Besides, boiling effects also appear on the secondary loop of the steam generator.
The boiling process allows significantly higher heat transfer rates compared to the single-phase convection. But this heat transport can be suddenly decreased when the limit of the critical heat flux (CHF) is reached. The occurrence of the boiling crisis leads generally to severe damage of the facility components and has to be avoided during reactor operation. Until today, there is no reliable method predicting this phenomenon based on universally valid correlations. A substantial benefit for the reactor safety research would be a prediction method which is based on the solution of the transport equations for the two-phase flow of water and steam.
There exist many correlations based on observations in experiments or theoretical reflections which try to explain the occurrence and the development of the critical heat flux. Unfortunately, they cannot be combined to one complete model as they are counter-predicting effects or are set up on different physical effects. For example, the ‘Near Wall Bubble Crowding Model’ [Kandlikar, S. G., 2011] postulates the decrease of the liquid flow to the wall due to turbulence with increasing heat flux as bubbles will concentrate near the wall. Whereas the ‘Interfacial Lift-Off Model’ [Galloway, J., Mudawar, I., 1993] predicts pseudo-periodic ‘wetting-fronts’ which cause the agglomeration of steam leading to the CHF as these zones lift off from the wall. Using the COSMOS-L test facility, IKET at KIT tries to contribute to analyzing the different existing theories and to examine specific phenomena like flow pattern or void distribution for flow boiling
