Experimental Study on the Effect of Initial Temperature on CHF in a Vertical Annulus Narrow Channel with Bilateral Heated

Abstract

Study on understanding of the complexities of boiling in the narrow channel which was occured in a severe accident on nuclear power plant has been carried out in experimentally using simulation apparatus in order to achieve the safety management capability. Critical Heat Flux (CHF) is one important parameter to control heat during transient accident. The methodology of research is an experiment using experiment apparatus called HeaTiNG-01 test section with modifications in the outside pipe using stainless steel material as the reactor vessel wall simulation. Experiments were conducted by heating the heated rod as a simulation of debris until the desired initial temperature by bilateral heated. Then water with a saturation temperature in atmospheric was poured gravitationally into the narrow channel. Data acquisition system recorded temperature changes in transient during the cooling process. The transient temperature profile in double heating surface and rewetting point (rewet fronts) was characterized. Experiment was conducted at three initial temperature variations i.e. 650^oC, 750^oC and 850^oC and using channel width 1 mm. Experiment data was used to calculate heat flux then to fitting CHF form boiling curve. The results showed that CHF in outer pipe is higher than heated rod, these conditions explain that more heat is released through the outer pipe, so that the heat control can be done from outside the system to reduce the temperature quickly. The average value of CHF for each vertical position 100 mm and 400 mm at outer pipe are 380 kW/m2 and 733 kW/m², and then at the heated rod are 250 kW/m² and 497 kW/m². 

Received: 20 November 2010; Revised: 25 July 2011; Accepted: 08 August 2011