Researchers have performed many studies to try to understand crud formation on the fuel pin clad surfaces, which has been observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling and precipitation reactions. Crud deposits, may cause an unexpected change in core power distribution known as crud induced power shift (CIPS) or axial offset anomaly (AOA) if boron species accumulate in the deposits. If the crud deposit is thick enough, the potential exists for fuel rod surface dryout. The Westinghouse Advanced Loop Tester (WALT) at the George Westinghouse Science and Technology Center (STC) will be utilized to examine the effects of crud formation on fuel pin clad dryout. This paper describes a single heated rod test facility designed and fabricated by Westinghouse to study the effects of crud deposition at PWR reactor operating conditions. This test loop is a single rod facility with or without grid support structures on the heater rod and can be used for forced convection and sub-cooled boiling tests with and without crud deposition. Sub-cooled boiling experiments at PWR reactor operating conditions are currently being performed at this facility. The single electrical heated rod in this test facility is instrumented with four movable thermocouples to measure the inside wall temperatures at four azimuthal locations within the rod. In addition, there are two fixed thermocouples to measure the inlet and outlet temperatures of the water flowing on the outside of the heated rod such that an overall energy balance (i.e. comparing the heat absorbed by the water coolant to the measured rod electrical power) can be performed on the facility. This paper will present forced convection and boiling heat transfer curves for clean rod surfaces. Comparison with forced convection correlations and sub-cooled boiling correlations are also presented in this paper.

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