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 f
low 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 now". Therefore, in the thermal hydraulic desig
n of research reactors, critical heat nux (CHF) under a counter-current flo
w limitation (CCFL) or a flooding condition are important to determine safe
ty margins of fuel against CHF during a core flow reversal.
The authors have proposed a CHF correlation scheme for the thermal hydrauli
c design of research reactors, based on CHF experiments for both upward and
downward flows including CCFL condition. When the CHF correlation scheme w
as proposed, a subcooling effect for CHF correlation under CCFL condition h
ad not been considered because of a conservative evaluation and a lack of e
nough 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 subcooli
ng effect on CHF under the CCFL conditions in vertical rectangular channels
heated from both sides were investigated quantitatively based on CHF exper
imental results obtained under uniform and non-uniform heat flux conditions
. 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. The new corr
elation could be adopted under the conditions of the atmospheric pressure,
the inlet subcooling less than 78 K, the channel gap size between 2.25 to 5
.0 mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 1
74 which were the ranges investigated in this study.