Rod bundle blowdown heat transfer tests simulating pressurized water reactor loss of coolant accident conditions. Final report

An experimental program has been conducted to investigate transient critical heat flux (CHF) in pressurized water reactor (PWR) cores during a hypothetical loss-of-coolant accident (LOCA). Twenty-seven (27) powered blowdown transients were performed in the test program. Loop and test section parameters were selected to simulate predicted average reactor core blowdown thermal-hydraulic conditions. All experiments were performed with a 25 (5 x 5 square array) rod bundle test section having a uniform axial heat flux profile. The rod bundle had an active heated length of 12.5 feet. The test results are discussed, including the experimentally determined times and locations of CHF.