The results of an experimental study of the integral hydraulic characteristics of two-phase flow in the mode of natural circulation in vertical channels are presented. A comparative analysis is done and determined are the limits of applicability of computational techniques used in the design and validation of safety systems of water-cooled nuclear power units, based on a comparison of the calculated dependences with the experimental results. Recommendations for the selection and refinement of the closing relations for friction pressure losses are proposed.
Keywords: vertical channel, two-phase flow, natural circulation, steam void, closing relations, coolant systems, friction, nuclear power unit.
Article language: Russian. Pp. 1–10. DOI: 10.26277/SECNRS.2019.94.4.002.
