Heat transfer to water from a vertical tube bundle under natural-circulation conditions

The natural circulation heat transfer data for longitudinal flow of water outside a vertical rod bundle are needed for developing correlations which are used in best estimate computer codes to model thermal hydraulic behavior of nuclear reactor cores under accident or shutdown conditions. The heat transfer coefficient between the fuel rod surface and the coolant is the key parameter required to predict the fuel temperature. Because of the absence of the required heat transfer coefficient data base under natural circulation conditions, experiments were performed in a natural circulation loop. A seven tube bundle having a pitch to diameter ratio of 1.25 was used as a test heat exchanger. A circulating flow was established in the loop, because of buoyancy differences between its two vertical legs. Steady state and transient heat transfer measurements were made over as wide a range of thermal conditions as possible with the system.