A detailed knowledge of the heat transfer and fluid flow behaviour around fuel pins in a nuclear fuel cluster is important to determine the thermal performance of a nuclear reactor. Spacers play an important and essential role of spacing out the nuclear fuel pins in the cluster and preventing the formation of hotspots. They also eliminate the possibility of fuel damage due to flow induced vibrations in the slender fuel pins. From the point of view of reactor thermal-hydraulics, they enhance momentum transfer (reflecting in greater pressure drop) and heat transfer. Additionally, they also enhance the mixing inside the fuel cluster. In this paper, the enhancement in mixing due to spacers has been experimentally quantified for a particular fuel cluster geometry. Such data are important for thermal-hydraulic prediction of flow and heat transfer inside rod bundles.
The ability of existing subchannel codes to predict the mixing phenomena and spacer effect has also been gauged in this work.
As outcomes of the study, firstly the downstream distance up to which the spacer effect is significant was quantified. Secondly, suitability of subchannel analysis code to predict the observed mixing is assessed.