Critical Heat Flux in a Vertical Annulus under Low Upward Flow and near Atmospheric Pressure
As future boiling water reactors (BWR), concepts of evolutional ABWR (ABWR-IER) and natural circulation BWR (JSBWR) have been investigated in order to reduce their construction cost and simplify their maintenance and inspection procedures. One of the promised features of the design of the evolutional ABWR is to reduce the number of internal pumps and to remove the Motor Generation (MG) sets. These design changes may induce boiling transition in the fuel rods of reactor core during a pump trip transient due to the more rapid flow coastdown characteristics than these of the present design. In addition, the understanding of critical heat flux (CHF) is one important subject to grasp safety margin during the start-up for the natural circulation BWR and to establish the rational start-up procedure in which thermo-hydraulic instabilities can be suppressed.<BR>The present study is to clarify CHF characteristics under low velocity conditions. CHF measurements were conducted in a vertical upward annulus channel composed of an inner heated rod and an outer tube made of glass. CHF data were obtained repeatedly under the condition of stable inlet flow to examine statistically their reproducibility. The flow regime was investigated from flow observation and measurement of differential pressure fluctuation. The CHF data are correlated with the flow regime transition. It was clear from the obtained flow pattern and the CHF data that the CHF behavior could be classified into specified regions by the mass flux and inlet subcooling conditions. A CHF correlation was developed and agreed with other researchers' data within acceptable error.
- Journal of nuclear science and technology
Journal of nuclear science and technology 34(6), 559-570, 1997-06-25
Atomic Energy Society of Japan