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Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

高稠密格子水冷却炉心の除熱技術開発の現状,2; 改良界面追跡法による二相流解析技術の開発

吉田 啓之; 玉井 秀定; 大貫 晃; 高瀬 和之; 秋本 肇

Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; 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、稠密炉心の除熱性能を評価する。この一環として本研究では、改良界面追跡法を用いた詳細二相流解析コードTPFITの開発を行っている。本報では、解析コードのベクトル並列化を行い、大規模解析に対応させるとともに、解析コード検証解析や稠密炉心を模擬した体系における解析結果を示す。

The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.

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