Experiments were carried out with a vertical rectangular channel simulating a subchannel of the upgraded JRR-3 fuel element, in order to investigate the validity and the error of the correlations predicting the superheat a t the onset of nucleate boiling. These correlations were used in the core thermal-hydraulic design of the upgraded JRR-3. AS the results, the following were made clear : @ T h e existing Bergles-Rohsenow correlation gives a good prediction for the relationship of heat flux vs. superheat at the onset of nucleate boiling, with the error of about 1 K against the lower limits of the measured superheat. 0 There are no significant differences in the characteristics of the relationship of heat flux U S . superheat a t the onset of nucleate boiling between upflow and downflow. @ There are no significant differences in the histories of relationship of heat flux us. superheat from the forced convection single-phase flow to the subcooled boiling between increasing heat flux and decreasing heat flux, with little overshoot of superheat a t the onset of nucleate boiling both in the upflow and in the downflow.
This paper presents the outline of the core thermohydraulic design and analysis of the research reactor JRR-3, which is to be upgraded to a 20MWt pool-type, light water-cooled reactor with 20% low enriched uranium (LEU) plate-type fuel. For the condition of normal operation, the upgraded JRR-3 core is planned to be cooled by two cooling modes of forcedconvection a t high power and natural-convection a t low power. The major feature of core thermohydraulics is that a t the forced-convection cooling mode the core flow is a downflow, under which fuel plates are exposed to a severer condition than an upflow in cases of operational transients and accidents. The core thermohydraulic design was, therefore, done f o r the condition of normal operation so that fuel plates may have enough safety margins both against the onset of nucleate boiling (ONB) not to allow the nucleate boiling anywhere in the core and against the departure from nucleate boiling (DNB) . T h e safety margins against ONB and DNB were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the ONB, and the minimum DNB ratio (ratio of DNB heat flux to the maximum heat flux) was evaluated to be about 2.1, which gives a sufficient margin against the DNB. T h e core thermohydraulic characteristics were also clarified for the natural-convection cooling mode.
The differences in the single-phase forced-convection heat transfer characteristics between upflow and downflow were investigated experimentally with a narrow vertical rectangular channel. The objectives of the experiment were to investigate in both laminar and turbulent flow regions the applicability of existing correlations to and the effects of buoyant force on the heat transfer characteristics in the narrow vertical rectangular channel, which is simulating a subchannel of 2.25 mm in gap and 750 mm in length in the fuel element of the research reactor, JRR-3 to be upgraded at 20MWt. As the results, it was revealed that (1) by use of equivalent hydraulic diameter, existing correlations are applicable to a channel as narrow as 2.25mm in gap for turbulent flow though the precision and critical Reynolds number are different among the correlations, and (2) in the laminar flow, the difference in heat transfer characteristics arises between upflow and downflow with Reynolds number less than about 700 and Grashof number larger than about 1,000, giving smaller Nusselt number for downflow than for upflow as the effect of buoyant force. New heat transfer correlations for channel heated from both sides are proposed as lower limits for upflow and downflow, respectively, in the laminar flow.
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