In order to ensure the compactness and high-power density of a nuclear power reactor, the research on tight-lattice fuel bundle, with a narrow gap distance between fuels, has been highlighted. Recently, KAERI (Korea Atomic Energy Research Institute) has been developing dual-cooled annular fuel to increase a significant amount of the reactor power in OPR1000 (Optimized Power Reactor), a PWR (Pressurized Water Reactor) optimized in the Republic of Korea. The dual-cooled annular fuel is configured to allow a coolant flow through the inner channel as well as the outer channel. To introduce the dual-cooled annular fuel to OPR1000 is aiming at increasing the reactor power by 20% and reducing the fuel-pellet temperature by 30%, as compared to the cylindrical solid fuel, without a change in reactor components. In such a case, due to larger outer diameter of a dual-cooled annular fuel, the dual-cooled annular fuel assembly exhibits a smaller P/D (Pitch-to-Diameter ratio) than the conventional cylindrical solid fuel assembly. In other words, the dual-cooled annular fuel array becomes the tight-lattice fuel bundle configuration, and such a change in P/D can significantly affect the thermal-hydraulic characteristics in nuclear reactor core.
In this paper, the pressure drop and flow pulsation in tight-lattice rod bundle were investigated. As the test sections, the tight-lattice rod bundle of P/D = 1.08 was prepared with the regular rod bundle of P/D = 1.35. The friction factors in P/D = 1.08 appeared smaller than those in P/D = 1.35. For P/D = 1.08, the twist-vane spacer grid became the larger pressure loss coefficients than the plain spacer grid. In P/D = 1.08, the flow pulsation, quasi-periodic oscillating flow motion, was visualized successfully by PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques. The peak frequency and power spectral density of flow pulsation increased with increasing the Reynolds number. Our belief is that this work can contribute to a design of nuclear reactor with tight-lattice fuel bundle for compactness and power-uprate and a further understanding of the coolant mixing phenomena in a nuclear fuel assembly.
OS8-3 Feasibility Study on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles (1) : 37-rod bundle tests and subchannel analysis