Experiments on counter-current flow limitation in a PWR hot leg model


Experiments on counter-current flow limitation in a PWR hot leg model

Lucas, D.; Beyer, M.; Pietruske, H.; Szalinski, L.

Counter-Current Flow Limitation (CCFL) is of importance for PWR safety analyses in several accident scenarios connected with loss of coolant events. To investigate the characteristics of this phenomenon and to establish a database suitable for CFD-model development and validation a 1:3 scaled flat model of a German Konvoi reactor hot leg including a part of the steam generator inlet chamber was built up. The test section was operated in the TOPFLOW pressure chamber to allow optical observation of steam-water flows at pressure levels up to 50 bars. The test series comprises air-water tests at 1 and 2 bar as well as steam-water tests at 10, 25 and 50 bar. The gas was injected from the tank simulating the reactor vessel while the water enters into the steam generator (SG) separator tank. A steel sheet divides the SG separator tank in two parts. The first one serves as completion of the SG inlet chamber from the water side and the second is necessary to keep the level in the first part constant, using the steel sheet as level drain. This allows to investigate CCFL under steady state conditions. During the experiments the flow structure was observed along the hot leg model using a high-speed camera and web-cams. In addition pressure was measured at several positions along the horizontal part and the water levels in the reactor simulator and steam generator simulator tanks were determined. From the measurements flooding curves basing on the Wallis parameters for gas and liquid were obtained. The results show a slight shift of the curves in dependency on pressure. In addition a slight decrease of the slope was found with increasing pressure. Additional investigations on the frequencies of liquid slugs were done.

Keywords: two-phase flow; counter-current flow limitation; flooding; experiment; hot leg

Involved research facilities

  • TOPFLOW Facility
  • Contribution to proceedings
    The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), 03.-08.09.2017, Xi’an, China
  • Lecture (Conference)
    The 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17), 03.-08.09.2017, Xi'an, China

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