Nuclear Technology / Volume 147 / Number 2 / August 2004 / Pages 191-201
Technical Paper / Thermal Hydraulics
In a severe accident of a light water reactor (LWR), heat transfer models in a narrow gap between superheated core debris and reactor pressure vessel (RPV) are important for evaluating the integrity of the RPV and emergency procedures. Newly developed heat transfer models are discussed that take into account both the local heat flux on a heated surface, which is characterized by the boiling regime, and the average critical heat flux (CHF) on a heated surface, which is restricted by countercurrent flow limitation (CCFL), including the effect of an inclination angle of the gap. The models were incorporated into the mechanistic detailed code RELAP/SCDAPSIM/MOD3.2. The local heat flux was applied to the outer surface of the debris and the inner surface of the RPV wall. The average CHF was evaluated through the CCFL phenomenon at each junction in the gap. For the assessment, an analysis of Japan Atomic Energy Research Institute's ALPHA test was performed. The calculated peak temperature response of the vessel showed good agreement with the experimental data. It was validated that the new models effectively simulate the coolability in a narrow gap, which could be an effective means of cooling the vessel wall and thereby preventing RPV failure, as was demonstrated in the TMI-2 accident.