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Abstract
Experimental investigations on critical heat flux (CHF) are mostly on vertical channels involving high mass fluxes and high system pressures. Reported studies on CHF in horizontal flow channels under low-pressure, low-flow (LPLF) conditions are limited. Understanding CHF is essential in the design and operation of heat exchangers and heat-generating devices including fuel channels of nuclear reactors. The present work investigates CHF in horizontal tubes for low steady flow at atmospheric pressure conditions. Appearance of a “red hot” spot on the test section is considered to be the occurrence of critical heat flux condition in this study. Present data could not be predicted using the reported method of applying a correction factor for the vertical lookup table data. A correlation using the experimental data is developed incorporating the fluid-to-fluid modeling parameters for the prediction of CHF in horizontal channels under LPLF conditions. Numerical study using thermal hydraulic system code RELAP5 suggests liquid film dryout as the mechanism of CHF occurrence in the present investigations.
NOMENCLATURE
| Bo | = | Bond number ( |
| CHF | = | critical heat flux (W/m2) |
| d | = | inner diameter (mm) |
| F | = | modified Froude number |
| g | = | acceleration due to gravity (m s−2) |
| G | = | mass flux (kg/m2-s) |
| hfg | = | enthalpy of vaporization (J/kg) |
| H | = | enthalpy (J) |
| K | = | correction factor or product of modified Froude number and square root of Reynolds number |
| L | = | length (m) |
| P | = | pressure (Pa) |
| q′′ | = | heat flux (W/m2) |
| Re | = | reynolds number (Gd/μ) |
| T | = | temperature (°C) or ratio of radial turbulent force to the buoyancy force |
| v | = | velocity (m/s) |
| X | = | Thermodynamic quality |
| XLM | = | Martinelli parameter |
| Z | = | Ohnesorge number ( |
Greek Symbols
| α | = | void fraction |
| ρ | = | density (kg/m3) |
| μ | = | viscosity (Pa-s) |
| ν | = | kinematic viscosity (cm2/s) |
| σ | = | surface tension (N m−1) |
Subscripts
| CHF | = | critical heat flux |
| f | = | liquid phase |
| g | = | vapor phase |
| h, hor | = | horizontal |
| in | = | inlet |
| min | = | minimum |
| max | = | maximum |
| v, ver | = | vertical |
Additional information
Funding
Notes on contributors
Parackal K. Baburajan
Parackal K. Baburajan is currently Scientific Officer at the Nuclear Safety Analysis Division of Atomic Energy Regulatory Board, Mumbai, India. He graduated in mechanical engineering from the Institution of Engineers (India) in 1995 and received a master of technology from Indian Institute of Technology, Bombay, in 2000, and is presently earning a Ph.D. at Indian Institute of Technology, Bombay. His research interests are nuclear reactor thermal hydraulics, severe accidents analysis of PHWRs, and experimentalstudies on severe accidents.
Govind Singh Bisht
Govind Singh Bisht graduated with a B.E. (mech. eng.) from Mumbai University in 2006. He is currently earning his M.Tech. in the Department of Energy Science and Engineering at IIT Bombay. His research interest is two-phase flow and heat transfer.
Avinash J. Gaikwad
Avinash J. Gaikwad, is the present Head, Nuclear Safety Analysis Division (NSAD), AERB. He is a chemical engineering gold medalist from Pune University. He is from the 32nd batch of BARC Training School and is presently pursuing a Ph.D. in the area of development of interleaving configuration for main heat transport system in advanced channel type BWRs. He has been working at the reactor safety division of BARC for more than two decades (1989 to 2012). His areas of specialization include thermal hydraulics for probabilistic safety assessment, passive system reliability, severe accident studies, deterministic safety analysis, in-house computer code development, and safety analysis using RELAP5/SCDAP. He has more than 90 publications in different international journals and conferences to his credit. He has been extensively involved in safety analysis for different reactor systems like AHWR, PHWR, PWR, BWR, and ADSS, including passive system reliability analysis and advanced process control studies.
Satish K. Gupta
Satish K. Gupta has been a distinguished scientist at Atomic Energy Regulatory Board, a visiting professor at IIT Kanpur, and a distinguished professor at IIT Guwahati. His current interests are multidimensional heat transfer in nuclear fuels, and heat removal in primary heat transport systems of water reactors using natural circulation.
S. V. Prabhu
S. V. Prabhu is currently a professor in the Department of Mechanical Engineering, Indian Institute of Technology, Bombay, India. He graduated with a B.E. (mech. eng.) first class with distinction from Mysore University in 1988, a master of technology from National Institute of Technology, Surathkal, in 1991, and a Ph.D. from Indian Institute of Technology, Bombay, in 1998. His research interests are flow metering heat transfer studies involving jet impingement, fire dynamics, renewable energy (hydrokinetic turbines and wind turbines), gas turbine blade cooling, and melting and solidification of phase-change materials (PCM) and metals.