| Summary: | The MIT Nuclear Research Reactor (MITR) is the only research reactor in the United
States that utilizes plate-type fuel elements with longitudinal fins to augment heat transfer.
Recent studies on the conversion to low-enriched uranium (LEU) fuel at the MITR,
together with the supporting thermal hydraulic analyses, propose different fuel element
designs for optimization of thermal hydraulic performance of the LEU core. Since
proposed fuel design has a smaller coolant channel height than the existing HEU fuel, the
friction pressure drop is required to be verified experimentally.
The objectives of this study are to measure the friction coefficient in both laminar and
turbulent flow regions, and to develop empirical correlations for the finned rectangular
coolant channels for the safety analysis of the MITR. A friction pressure drop experiment
is set-up at the MIT Nuclear Reactor Laboratory, where static differential pressure is
measured for both flat and finned coolant channels of various channel heights. Experiment
data show that the Darcy friction factors for laminar flow in finned rectangular channels
are in good agreement with the existing correlation if a pseudo-smooth equivalent
hydraulic diameter is considered; whereas a new friction factor correlation is proposed for
the friction factors for turbulent flow. Additionally, a model is developed to calculate the
primary flow distribution in the reactor core for transitional core configuration with
various combinations of HEU and LEU fuel elements.
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