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Proving DRACO will deliver
The United States is now closer than it has been in over five decades to launching the first nuclear thermal rocket into space, thanks to DRACO—the Demonstration Rocket for Agile Cislunar Orbit.
David D. Hall, Issam Mudawar
Nuclear Technology | Volume 117 | Number 2 | February 1997 | Pages 234-247
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT97-A35328
Articles are hosted by Taylor and Francis Online.
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 × 103 to 138 × 103 kg/m2.s, pressures from 1 to 223 bars, inlet subcoolings from 0 to 347°C, inlet qualities from —2.63 to 0.00, outlet subcoolings from 0 to 305°C, outlet qualities from —2.13 to 1.00, and CHFs from 0.05 × 106 to 276 × 106 W/m2. The database contains 4357 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.