Untersuchung horizontaler, zweiphasiger Rohrströmung unter besonderer Berücksichtigung des Entrainment

Modeling the thermo-hydraulic behavior of nuclear facilities is an important aspect of reactor safety research. Highly intermittent flow regimes which may occur during loss of coolant accidents can lead to high levels of stresses on the structure materials and to unfavorable feedback to the neutronic behavior of the reactor core.

Strong spatial and temporal scale spreads in the regimes pose major challenges to the numerical modeling of the phenomena. In order to meet the requirements, a numerical method was developed which combines the advantages of the scale averaging Euler-Euler two-fluid model with respect to the numerical effort in large computational domains with the ability of the volume-of-fluid method to describe stratified flows quantitatively accurate. The process is characterized by high flexibility with regard to the dynamic transition processes between disperse and stratified flow regions.