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Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1815 - 1822, 1997/00
In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as "Loss of the primary coolant flow". Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and nonuniform heat flux condition. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed.
Tokuhiro, Akira; ; Kimura, Nobuyuki
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1712 - 1723, 1997/00
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Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, 0 Pages, 1997/00
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; Brear, D. J.; Tobita, Yoshiharu; Morita, Koji
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1340 - 1348, 1997/00
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Tobita, Yoshiharu
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1357 - 1364, 1997/00
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Morita, Koji; Tobita, Yoshiharu; ;
Proceedings of 8th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), Vo.3, p.1332 - 1339, 1997/00
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