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

高稠密格子水冷却炉心に対する二相流解析技術開発

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

日本原子力研究開発機構において開発が進められている超高燃焼水冷却増殖炉の熱設計においては、詳細二相流解析手法により、稠密炉心の助熱性能を評価する。この一環として本研究では、改良二流体モデルを用いた二相流解析コードACE-3Dの開発を行っている。本報では、解析コードの概要と解析結果について述べる。

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is started at Japan Atomic Energy Agency (JAEA) in collaboration with power company, reactor vendors, universities since 2002. The FLWR can attain the 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 to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. In FLWR core, there is no sufficient information about the effects of the gap width and grid spacer configuration on the flow characteristics yet. Then, we started development of qualitative analytical procedures on thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this paper, we describe the outline of the simulation technology and examples of numerical results.