Abstract
In relation to the design of an innovative Water Reactor for FLexible fuel cycle (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap between adjacent fuel rods. In view of the importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study numerically simulated steam-water two-phase cross flow between two modeled subchannels of tight-lattice rod bundle for the FLWR by using a detailed two-phase flow simulation code with an advanced interface tracking method (named TPFIT), statistically evaluated the simulation results, and clarified mechanisms of cross flow for developing a model. The effects of flow pattern, inlet and outlet of mixing section, and gap spacing on cross flow, and the local and general characters of cross flow were extensively investigated. It was confirmed that there exist strong correlation between differential pressure and gas/liquid mixing coefficients. Mechanistically, cross flow results mainly from differential pressure. Liquid cross flow occurs locally and its time lag is negligibly small (less than 2 ms). Gas cross flow, however, occurs across the whole mixing section, and propagates with main stream in the mixing section. The time and space lags are relatively large and can be determined from average velocities in mixing section. The local properties may be negligible as well.
