低減速炉稠密流路におけるスペーサ効果基礎実験 : 中性子ラジオグラフィ法によるフローパターンと沸騰熱伝達の同時測定(FR2 計測制御の最先端)

概要

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Fundamental thermal-hydraulic experiments have been conducted within a tube for flow boiling characteristics study in the Reduced-Moderation Water Reactor (RMWR) core. It is necessary to study geometrical influences of fuel rod spacer on critical power are necessary to study for thermal-hydraulic design development of RMWR core rod assemblies. In the present study, the effects of spacer thickness and spacer-rod clearance on high-void fraction two-phase flow and heat transfer are focused. Neutron radiography(NRG) technique was used to visualize high-void fraction two-phase flow with sufficient spatial resolution. Thermo-couples on the inside surface of a heated tube were used. An electrically heated tube was used with a plate blockage of various thickness under atmospheric pressure conditions. The results show that the boiling heat transfer inside the spacer was higher than downstream, far from the spacer. The forced flow deviation by the spacer was observed with the NRG visualization.
社団法人日本機械学会の論文
2004-06-22