Critical Heat Flux (CHF) and Cold-Flow Pressure Drop Investigations of R-11 in a Vertical Uniformly Heated 5 x 5 Rod Bundle

Experimental and numerical investigations were performed to determine the critical heat flux (CHF) and pressure drop in a vertical uniformly heated 5×5 rod bundle. R-11 is used as the working fluid due to its low latent heat, low critical pressure and well known properties. The experimental investigation was performed at the Westinghouse Nuclear Fuel PWR Product Technologies Test Laboratory in Columbia, South Carolina. Two types of grid designs were tested with Bundle-1 and Bundle-2 respectively. The matrix of DNB test parameters (water equivalent conditions shown in parentheses) includes exit pressure: 317–445 psia (1800–2400), inlet mass flux: 1.725–3.400 Mlbm/hr-ft² (1.5–2.5 Mlbm/hr-ft²) and inlet temperature: 170–305°F (320–590 Mlbm/hr-ft²). Nominal R-11 conditions for the cold test are 165 psia, and 80°F±10°F. The experimental results were validated using a nuclear thermal hydraulic code VIPRE-01, MOD-02.1. The EPRI-1 correlation was being used to have a gross understanding of the CHF in the Bundles. Subsequently an empirical correlation USC11F was developed with Bundle-1 CHF database which compares favorably with experimental CHF for both the grid designs. Overall mean ratios of experimental to predicted CHF of 0.9984 and 0.9948 with standard deviations of 0.0265 and 0.0338 were obtained for Bundle-1 and Bundle-2 respectively. The predicted cold flow pressure drops also reproduced the experimental values closely. At lower mass flux a slight overprediction was encountered due to the dependability of mixing grid loss coefficient on Reynold’s number.