Nuclear Technology / Volume 158 / Number 2 / May 2007 / Pages 219-228
Technical Paper / Nuclear Reactor Thermal Hydraulics / dx.doi.org/10.13182/NT07-A3837
Critical heat flux (CHF) in rod bundles is a parameter of great importance for the thermal-hydraulic design and safety analysis of advanced light water reactors. An experimental investigation has been conducted for the 19-rod hexagonal test bundles with a tightly spaced nonsquare arrangement of heater rods. The parametric effects on the CHF were examined for the heated length, the unheated rod, and the nonuniform axial power shape. As a result, a pertinent CHF correlation has been developed on the basis of the bundle cross-sectional averaged conditions. The available CHF database for rod bundles with square and nonsquare rod pitches was employed for the assessment of representative CHF correlations that were applicable to the round tubes and rod bundles. The database covered a wide range of operating conditions and test bundle geometries that are applicable to advanced light water reactors. The prediction accuracy of the CHF correlations was evaluated on the basis of the local thermal-hydraulic conditions calculated by a subchannel analysis code.