Improvement of CHF correlation for research reactors using plate-type fuels

In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities in their core and are cooled down by a downward core flow. A core flow reversal from a steady-state 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, a critical heat flux (CHF) under a countercurrent flow limitation (CCFL) or a flooding condition including a zero flow condition are important to determine safety margins of fuel against CHF during a core flow reversal. The present authors have proposed a CHF correlation scheme for the thermal hydraulic design and safety analysis of research reactors such as JRR-3, based on CHF experiments for both upward and downward flows including CCFL or the flooding condition including the zero flow condition using vertical rectangular channels heated from both sides. When the CHF correlation scheme was proposed, the subcooling effects for CHF correlation under a CCFL condition had not been considered because of a conservative evaluation. A too conservative evaluation is not always appropriate for the design of research reactors such as a high performance research reactors, because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effects for CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on existing CHF experimental results. Effects of axial power profiles for CHF under the CCFL conditions were also studied. (author)