Evaluation of subcooled critical heat flux correlations using the PU-BTPFL CHF database for vertical upflow of water in a uniformly heated round tube
- Purdue Univ., West Lafayette, IN (United States)
A simple methodology for assessing the predictive ability of critical heat flux (CHF) correlations applicable to subcooled flow boiling in a uniformly heated vertical tube is developed. Popular correlations published in handbooks and review articles as well as the most recent correlations are analyzed with the PU-BTPFL CHF database, which contains 29,718 CHF data points. This database is the largest collection of CHF data (vertical upflow of water in a uniformly heated round tube) ever cited in the world literature. The parametric ranges of the CHF database are diameters from 0.3 to 45 mm, length-to-diameter ratios from 2 to 2484, mass velocities from 0.01 {times} 10{sup 3} to 138 {times} 10{sup 3} kg/m{sup 2}{center_dot}s, pressures from 1 to 223 bars, inlet subcoolings from 0 to 347 C, inlet qualities from {minus}2.63 to 0.00, outlet subcoolings from 0 to 305 C, outlet qualities from {minus}2.13 to 1.00, and CHFs from 0.05 {times} 10{sup 6} to 276 {times} 10{sup 6} W/m{sup 2}. The database contains 4,357 data points having a subcooled outlet condition at CHF. A correlation published elsewhere is the most accurate in both low- and high-mass velocity regions, having been developed with a larger database than most correlations. In general, CHF correlations developed from data covering a limited range of flow conditions cannot be extended to other flow conditions without much uncertainty.
- DOE Contract Number:
- FG02-93ER14394
- OSTI ID:
- 449569
- Journal Information:
- Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 2 Vol. 117; ISSN 0029-5450; ISSN NUTYBB
- Country of Publication:
- United States
- Language:
- English
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