Pressurized water reactor blockage—prediction of laminar flow and temperature distributions following a loss of coolant accident

Abstract

During the unlikely occurrence of a loss of coolant accident (LOCA) in a pressurized water reactor, it has been postulated that the fuel rod cladding may swell due to the combination of a pressure differential across the cladding and the increased temperature levels in the core. In this event, adjacent fuel rods may ‘balloon’ until they make contact with their neighbours, leading to a reduction in subchannel flow area, subchannels of highly noncircular cross-section, and worsening heat transfer in the blocked region of the core. A particularly important measure in predicting an upper limit on the severity of this core blockage is the peripheral variation of the wall temperature for a representative blocked subchannel. This paper is concerned with the prediction of the temperature variations in the duct wall and the fluid for fully developed laminar flow. A single subchannel is modelled as a four-cusped duct, bounded by a conducting wall of constant thickness, which is subject to uniform heat flux from the fuel. Results for this idealized problem are presented for different values of the thickness and thermal conductivity of the cladding. The important wall temperature distributions have been calculated for superheated steam to cover fluid flow conditions which might be envisaged during a LOCA. Here it must be observed that two-phase flow effects are unlikely to lead to worse heat transfer than can be predicted for single-phase steam cooling. Thus, the predicted temperature variations represent an upper bound for the low Reynolds number end of the Reflood phase in the LOCA.