An Experimental Facility for Two-Phase Flow Regime Identification for a Nuclear Reactor’s Reliability Studies

In nuclear reactors like India’s 700 MWe Pressurised Heavy Water Reactors, a Water-Steam Mixture is present in the Primary Heat Transport system. The WSM results in different flow patterns in the PHT. The patterns are better known as Flow Regimes. The FRs are classified as Only-water (single liquid phase), Bubbly, Slug, Churn and Annular. The Flow Regimes are responsible for thermal hydraulics of the system affecting heat transfer from the Primary Heat Transport and are responsible for certain safety aspects. For example, Slug Flow Regimes has more pressure drop and produces severe pressure oscillations which may cause harm to the system components. Thus, Flow Regime Identification in the Primary Heat Transport, would be an effective tool to ensure the adequacy of the heat transfer and support the reactor safety. For this purpose, an Experimental Facility (EF) for FR simulation and FRI has been evolved with an intention to extend the laboratory results to the actual Pressurised Heavy Water Reactors. Pressure sensor based FRI techniques are considered as the pressure measurements are possible with the least penetrations in the Primary Heat Transport's flow lines and the signals can be extended to a distance from the measurement points. In the Pressurised Heavy Water Reactors, the FLs with different internal diameters are provided. This is also a design parameter. The major components of the EF are replaceable test sections, air and water flow measurements and controls, an air–water mixer to generate various FRs, calibrated pressure and flow sensors, a video facility for the FRs’ capture, and a data acquisition system. This manuscript describes in detail the features of the EF. A set of experimental results for the FRI are also included.