Abstract
The implementation of passive safety systems in nuclear reactors provides the opportunity to enhance the nuclear safety. On the other hand, an accurate and reliable prediction of the heat removal behavior is not ensured because the operating conditions of certain types of passive systems like containment cooling systems differ from the validity ranges of the established heat transfer correlations. Therefore, a generic and detailed investigation is still necessary for passive systems.
Against this background, the test facility GENEVA was erected at Technische Universität Dresden in 2012. Since the commissioning, generic experiments concerning the system and stability behavior of this facility, which emulate a low-pressure and low-flow (LPLF) natural circulation system, were provided. Nevertheless, the investigation of the heat transfer behavior remained an open issue. On this account, the instrumentation in the heat transfer region inside GENEVA was improved to gather the necessary temperature and void fraction profiles.
The performed experiments provide a generic and wide database concerning boiling in a LPLF natural circulation systems. Within this paper, the development of the wall and bulk fluid temperature as well as the axial and center line void fraction profile in a slightly inclined tube for different heat flow rates are discussed. Furthermore, flow patterns could be identified on behalf of the void fraction measurements. To conclude the experimental analysis, the development of the heat transfer coefficient was estimated.
These experimental data provide the basis for a simulation with the lumped-parameter thermal-hydraulic code ATHLET and serve as validation reference. However, the comparisons between the experimental and computational results show insufficient agreements. Mainly, the simulation misses the saturation point of the experiments, which leads to great differences of the void fraction values. Moreover, inaccuracies appear as well with the heat transfer coefficient.
The experimental and computational results that are discussed in this paper provide the basis for the advancement not only of heat transfer correlations but also of flow pattern maps within the range of low pressure natural circulation system. In summary, this investigation contributes to the general purpose to enhance nuclear safety by providing an accurate and reliable prediction of the heat removal capacity of passive systems.
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