Scaling criteria for modeling natural- and forced-convection loops

Nuclear reactor safety regulations have required extensive thermal-hydraulic testing of simulated reactor systems and components. In view of the inherent difficulties associated with full-scale testing, scale models for prototype systems have been extensively used to predict the behavior of nuclear reactor systems during normal and abnormal operations as well as under accident conditions. Several studies have been performed to establish similarity relations between a prototype and scale model. It is the purpose of the present study to develop scaling criteria for a forced and natural circulation loop under single- and/or two-phase flow conditions, and to apply the criteria to obtain the preliminary conceptual design parameters for the B and W 2 x 4 loop system. The 2 x 4 loop scaled system contains representative components of all thermal-hydraulic systems considered important in performing tests to obtain data representative of the response of the prototype plant