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炉心内複雑二相流挙動に関する数値解析

Numerical analysis of complicated two-phase flow behavior in nuclear reactor cores

高瀬 和之; 吉田 啓之 ; 玉井 秀定; 小瀬 裕男*; 青木 尊之*; Xu, Z.*

Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*; Aoki, Takayuki*; Xu, Z.*

次世代型軽水炉の燃料集合体を対象にして、開発中であるシミュレーションを主体とした炉心熱設計手法を使って複雑な二相流挙動の予測評価を行った。使用したコードは著者らが開発している界面追跡法を改良したTPFITと商用コードとして世界的に有名であるFLUENTである。TPFITは気液二相流現象の高精度予測が可能であり、FLUENTは非構造格子による複雑流路形状の解析が可能である。本研究によって、燃料棒表面を流れる液膜の挙動,燃料チャンネルを流れる気泡の挙動,スペーサ領域で飛散する液滴の挙動などの詳細を定量的に把握することが可能になり、熱設計精度向上に関しての見通しが得られた。

Three-dimensional large-scale numerical simulations were carried out to predict the complicated water-vapor two-phase flow characteristics in a fuel bundle of an advanced light water reactor. Conventional analysis methods with a two-fluid model need composition equations and empirical correlations based on the experimental data. Therefore, it is difficult to obtain high prediction accuracy when experimental data are nothing. Then, a new two-phase flow analysis method was proposed and the TPFIT code was developed. This paper describes the predicted liquid film, bubbly and droplet flow behavior in the simulated fuel channels with the TPFIT code, and the predicted two-phase flow behavior around a curved fuel rod with the FLUENT code which is one of the most famous commercial code. From the present results, the high prospect was acquired on the possibility of development of the thermal design procedure of the advanced nuclear reactors by large-scale simulations.

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