An Experimental Analysis on Nusselt Number of Natural Circulation Flow in Transient Condition Based on the Height Differences between Heater and Cooler

Abstract

A better understanding on the phenomenon of natural circulation flow for cooling systems is necessary prior to improving the safety of nuclear power plant, not only in normal operation but also in accident conditions. One way to understand this phenomenon is by analyzing the Nusselt number in various geometrical dimensions through experimentation. The purpose of this study is to understand natural circulation phenomenon in transient condition by varying height differences between heater and cooler. To achieve this purpose, an experiment apparatus called NC-Queen was developed and arranged to enable three variations of height differences between heater and cooler, i.e., 1.4 m, 1.0 m, and 0.3 m. It is made of a stainless steel tube with a diameter of 1 inch, arranged in rectangular shape 6.4 m in length, and uses water as coolant. The initial temperature of the heater was set at 90 °C. The Nusselt number was obtained by calculating the flow rate as a function of transient temperature. The results confirm that height differences affect thermal properties and flow region based kinetics characteristics of water. In initial condition, decreasing height difference from 1.4 m to 1.0 m resulted in flow rate reduction of 16.7 %, while decreasing height difference from 1.4 m to 0.3 m resulted in a 39.1 % flow rate reduction. In final condition, the flow rate reductions were 75 % and 82.6 %, respectively. Meanwhile, in initial condition, the Nusselt number for height difference reduction from 1.4 m to 1.0 m and from 1.4 m to 0.3 m decreased by 30.5 % and 74.6 %, respectively, while for final condition, the Nusselt number decreased by 11.9 % and 67.4 %, respectively. The new constants in relationship between Nusselt number and the height difference are a = 20.06 and b = 0.56. The dominance of turbulent flow provides a good safety margin with indications of the large amount of heat released.

Computational fluid dynamics analysis of the natural circulation system in vertical heater horizontal cooler (VHHC) molten salt loop

“…Analysis of the safety aspects is crucial before physically building a nuclear reactor to deal with possible accidents due to the failure of the operation of technical components or accidents due to natural disasters. Like the nuclear incident in March 2011 at Fukushima Daiichi, Japan, one of the causes was the failure of active cooling devices due to a lack of electricity supply due to the earthquake and tsunami [1]. Motivated by this, the safety of nuclear power plants with the use of passive safety features has become a method that has begun to be widely developed and applied [2].…”

Experimental and simulation approach of the loop geometry effect on the natural circulation system of the advanced nuclear reactor

“…The cause of the nuclear incident in March 2011 at Fukushima Daiichi, Japan, was the failure of an active cooling device due to a lack of electricity supply that occurred as a result of earthquake and tsunami 1,2 . Motivated by this, the safety of nuclear power plants using a passive safety system has become a method that has been developed and applied widely 3,4 .…”