Analysis of a Subchannel Void Distribution in a 8x8 Rod Bundle Using the MATRA Code

Subchannel analysis codes have been widely adopted in the design calculation for an LWR core since it provides reasonably accurate results on the flow and enthalpy distributions inside the rod bundles with a pertinent computing time. A subchannel analysis code MATRA is under development at KAERI which is purposed to analyze the thermal hydraulic characteristics in rod bundles for several kinds of nuclear reactors; PWR, integral-type advanced reactor, supercritical light water reactor, and a gas cooled reactor. The applicability of the subchannel codes based on the mixture equations has usually been examined at subcooled or low quality conditions. Recently, the MATRA code was extended to higher quality conditions by employing the void drift model. The subchannel exit quality and mass velocity were evaluated for 3x3 or 4x4 test bundles. In this study the subchannel void distribution data obtained from the 8x8 BWR fuel bundles has been evaluated by the MATRA code. The experimental data was collected from a OECD/NRC benchmark program