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Home / Publications / Journals / Nuclear Technology / Volume 197 / Number 2

Experimental Investigation of Void Drift in Simulated Subchannels of a Natural-Circulation Pressure Tube–Type BWR

M. P. Sharma, A. K. Nayak

Nuclear Technology / Volume 197 / Number 2 / February 2017 / Pages 158-170

Technical Paper / dx.doi.org/10.13182/NT15-127

First Online Publication:January 16, 2017
Updated:February 16, 2017

The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube–type, heavy water–moderated, and boiling light water–cooled natural-circulation–based reactor. The fuel bundle of AHWR contains 54 fuel rods arranged in three concentric rings of 12, 18, and 24 fuel rods. This fuel bundle is divided into a number of imaginary interacting flow passages called subchannels. Transition from a single-phase-flow condition to a two-phase-flow condition occurs in the reactor rod bundle with increase in power. Prediction of the thermal margin of the reactor has necessitated the determination of intersubchannel mixing due to void drift. Void drift is due to redistribution of the non-equilibrium void fraction to attain an equilibrium void fraction. This redistribution occurs in the reactor rod bundle until it reaches the state of equilibrium void fraction. Hence, it is vital to evaluate void drift between subchannels of AHWR rod bundles.

In this paper, experiments were carried out to investigate the void drift phenomena in simulated subchannels of AHWR. The size of the rod and the pitch in the test section were the same as those of the actual rod bundle in the prototype. Three subchannels are considered in 1/12th of the cross section of the rod bundle. Water and air were used as the working fluid, and the experiments were carried out at atmospheric condition without the addition of heat. The void fraction in the simulated subchannels was varied from 0 to 0.8 under various ranges of superficial liquid velocities. The void drift between the subchannels was measured. The test data were compared with existing models in the literature. It was found that the existing models could predict the measured equilibrium void fraction in the rod bundle of the reactor within the range +8% to −14%.

 
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