Development of analytical procedures of two-phase flow in tight-lattice fuel bundles for Innovative Water Reactor for Flexible Fuel Cycle

超高燃焼水冷却炉に対する稠密燃料集合体内二相流解析手法の開発

吉田 啓之  ; 大貫 晃; 三澤 丈治; 高瀬 和之; 秋本 肇

Yoshida, Hiroyuki; Onuki, Akira; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

An R&D project to investigate thermal-hydraulic performance in the tight-lattice rod bundles of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been in progress at Japan Atomic Energy Agency (JAEA) in collaboration with power companies, reactor vendors, and universities since 2002. The FLWR can realize favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used because they increase the conversion ratio by reducing the moderation of neutrons. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. Information about the effects of the gap width and grid spacer configuration on the flow characteristics in the FLWR core is still insufficient. Thus, we are developing procedures for qualitative analysis of thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this study, an advanced two-fluid model is developed to economize on the computing resources. In the model, interface structures larger than computational cells (such as liquid film) are simulated by the interface tracking method, and small bubbles and droplets are estimated by the two-fluid model. In this paper, we describe the outline of this model and the numerical simulations we performed to validate the model performance qualitatively.