Conference Agenda

Post-Critical Heat Flux (Post-CHF) is one of the most complex two-phase phenomena significantly affecting the coolability of nuclear fuel during a loss of coolant accident (LOCA) in light water reactors. Wall heat transfer characteristics in inverted annular film boiling (IAFB), inverted slug film boiling (ISFB), and dispersed flow film boiling (DFFB) regimes have been widely investigated in the literature to develop heat transfer models/correlations to predict the peak cladding temperature among other parameters. However, lack of comprehensive data necessary for validating physical assumptions made during modeling of the IAFB/ISFB/DFFB regimes has led to limited predictive capabilities of existing models and correlations. In this study, a series of quasi steady-state IAFB, ISFB, and DFFB experiments were performed in the Post-CHF Heat Transfer (PCHT) test facility at the University of Michigan that employs a direct hot-patch technique to stabilize the quench fronts in a tubular test section made of Incoloy 800H with an inner diameter of 12.95 mm. Experimental conditions spanned over a relatively broad range to investigate the effects of the liquid subcooling, mass flux, and system pressure on heat transfer in those regimes. Detailed test section wall temperature was acquired using thermocouples and the void fraction in the test section was measured using a gamma densitometer and an X-ray radiography system. In addition, the predictive capabilities and limitations of the existing models for those regimes were evaluated using the acquired film boiling experimental data. Based on the model benchmark results, improvements were proposed to enhance the model accuracy for IAFB/ISFB/DFFB wall heat transfer.

[1] This abstract is intended for Dr. Sun’s American Nuclear Society Thermal Hydraulics Division Technical Achievement Award (ANS THD TAA) Lecture.

This abstract was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed in this paper, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the U.S. Nuclear Regulatory Commission.