Rod Bundle Heat Transfer Facility Two-Phase Mixture Level Swell and Uncovery Test Experiments Data Report (NUREG/CR-7218, Volume 2)

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Publication Information

Manuscript Completed: July 2016
Date Published: September 2016

Prepared by:
L.E. Hochreiter, F-B. Cheung, T.F. Lin, D.J. Miller, B.R. Lowery

The Pennsylvania State University
University Park, PA 16802

Prepared for:
Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

K. Tien, Project Manager

NRC Job Code N6154

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Abstract

A series of two-phase level swell and uncovery experiments have been performed in the US Nuclear Regulatory Commission/ Penn State Rod Bundle Heat Transfer Test (RBHT) Facility. A total of 75 experiments were performed in a quasi-steady state manner in which the inlet flooding rated into the RBHT rod bundle was slowly decreased in steps and the two-phase mixture level in the bundle was allowed to decrease. In several of the experiments the top region of the rod bundle became uncovered and the heater rod temperatures were significantly above the saturation temperature.

The range of conditions investigated in the experiments were: pressure, 0.138 to 0.414 Mpa (20 to 60 psia); Inlet subcooling 11.1 to 69.4 degrees K (20 to 125 degrees F); Inlet injection temperature 334 to 393 degrees K (142 to 247 degrees F); Peak linear power 0.492 to 1.31 kw/m (0.15 to 0.4 kw/ft); and Inlet flooding rate 2.54 to 40.64 mm/s (0.1 to 1.6 in/s).

A one-dimensional energy balance was used to calculate the saturation location in the bundle as well as the local fluid quality. The resulting calculations were used to estimate the single and two-phase friction and acceleration pressure drop components such that the differential pressure measurements could be corrected and used to estimate the local void fraction distribution along the heated bundle. The two-phase mixture level or dryout locations were also determined from the heater rod thermocouple response as the local heat transfer changed from boiling to steam cooling. The resulting data can be used to assess the void fraction models and heat transfer models in the Nuclear Regulatory Commission advanced safety analysis computer codes.