Heat transfer regimes for a flow of water at supercritcal conditions in vertical channels

• Published in: Thermal engineering 2017-11, Vol.64 (11), p.849-855
• Main Authors: Deev, V. I., Kharitonov, V. S., Churkin, A. N., Baisov, A. M.
• Format: Article
• Language: English
• Subjects: Channels; Criteria; Critical point; Engineering; Engineering Thermodynamics; Heat and Mass Transfer; Heat and Mass Transfer and Properties of Working Fluids and Materials; Heat transfer; Heat transfer coefficients; Mathematical analysis; Nuclear reactors; Reactor cores; Water flow
• ISSN: 0040-6015; 1555-6301
• DOI: 10.1134/S0040601517110039

Heat transfer regimes observed in experiments with water at supercritical conditions flowing in vertical channels of various cross-sections (such as round pipes, annulus, or rod bundles) are analyzed. In accordance with the established practice, the normal and the deteriorated heat transfer regimes were singled out as the basic regimes specific for heat carriers with highly variable properties. At the same time, it has been established that most published experimental data on supercritical pressure water heat transfer along the length of test sections demonstrate combined (or transient) heat transfer regimes. The features can be presented as a superposition of characteristics of the above-mentioned basic regimes. The combined regimes are not stable in certain ranges of water flow conditions in which sudden transitions between the basic regimes can occur. A system of similarity criteria governing heat transfer rate in the vicinity of the critical point is examined. As applicable to cores of water-cooled reactors, due to a small hydraulic diameter of cooling channels, buoyancy forces acting in these channels are negligible as compared with the inertia effects caused by thermal acceleration of the flow and viscous force. This concept yields two integrated criteria whose use in the correction factors for the basic heat transfer equation, which we proposed previously for the normal regimes, adequately (with an error of 20–25%) describes the specific of the heat transfer coefficient in the normal, deteriorated, and combined regimes. A system of equations is proposed for design calculation of heat transfer in channels of nuclear reactors cooled with supercritical pressure water.