Experimental Use of Coolant Flow Meter

ABSTRACT

In this study, experiments were performed at different power levels to monitor some thermo-hydraulic parameters like core mass flow rate, coolant velocity, mass flux, density and Reynolds number in the core of NIRR-1. Two probes with thermocouples were used in the experiments to measure the bulk temperature of the coolant in the coolant channel. The results were compared with the calculations performed by theoretical analyses. A data acquisition system (DAS) was developed to monitor the operational parameters in real-time. Information displayed was recorded on hard disk in a historical database. Flowmeter application during reactor, conditions and the results of measurement are presented. Details of the modified flowmeter to be used in BN 600 reactor, that enables its verification during reactor operation by the correlation method have been briefly treated. These experiments confirm the efficiency of natural circulation in removing the heat produced in the reactor core by nuclear fission.