Void Fraction in Spent Fuel Pool under Loss of Cooling Accident

Abstract

Thermohydraulic behavior in spent fuel pool is quite important in evaluating safety of a nuclear reactor under accidental conditions. Particularly, accurate prediction of void fraction in spent fuel pool is indispensable for evaluating cooling characteristics of spent fuel. In view of these, experimental and analytical studies were carried out for void fraction in spent fuel pool. The experiment was performed to measure the heat-up and void fraction transient during the postulated SFP accident. In this experiment, a simulated 7x7 BWR rod bundle that consists of 49 heater rods, 7 spacer grids and upper tie-plate was used. The measured data was compared with the some drift-flux correlations under the low pressure and the low flow rate condition related to SFP accident.