報告書番号

T98064

タイトル（和文）

大口径管内気液二相流評価手法の高度化（その４） -上昇流部入口境界条件による流動特性への影響-

タイトル（英文）

Improvement of Estimation Method for Two-Phase Flow in a Large Diameter Pipe (Part 4) -Effect of the Inlet Boundary Condition of the Upward Flow Section on Flow Characteristics-

概要 （図表や脚注は「報告書全文」に掲載しております）

口径100ｍｍ以上の大口径管内気液二相流動特性は、詳細な実験データの絶対的な不足により充分に把握されていない。その為、高精度な予測が要求される原子炉流動解析の大口径管内気液二相流への適用性が充分に検討できない状態にある。そこで本研究においては大口径管内気液二相流動特性を解明する為に、内径155mmの垂直管における水・蒸気気液二相流動の多次元計測を行い、各相分率、流速、気泡径等の径方向分布データを取得した。その結果上昇流部入口境界条件の違いにより未発達領域におけるドリフト速度が大きく影響を受けることがわかった。また局所平均気泡径の相関式ををボイド率、圧力の関数として導出し、本実験で得られた気泡径データを既存の相関式よりも大幅に精度良く予測できることを示した。

概要 （英文）

In nuclear power plants, many pipes are subject to steam-water two-phase flow, and these pipes have a wide range of lengths and hydraulic diameters. For rational design and performance estimation of the plants, the two-phase flow in these pipes must be predicted with high accuracy. At present, data from fully developed two-phase flow experiments of small-diameter pipes is mainly used to construct the correlations of the nulear reactor safety analysis code. However, it is known empirically that the characteristics of the two-phase flow are affected by the hydraulic diameter of the pipe. Hence, the validity of the applicability of the existing correlations to the two-phase flow in large-diameter pipes should be confirmed. To ensure this validity, the flow structure of the two-phase flow in large-diameter pipes are required to be understood in detail. In the present study, steam-water two-phase flow in a large-diameter vertical pipe (inner diameter:D=155mm) was investigated. Lateral distribution data of phase volume fraction, gas velocity and bubble diameter was obtained. The effects of the inlet boundary conditions were also observed. The void profile was shown to converge into a core-shaped distribution under every inlet boundary condition. The drift velocity in the developing region was considerably affected by the inlet boundary condition. The correlation of the Sauter mean bubble diameter was proposed as a function of the local void fraction and the pressure. The proposed correlation showed high accuracy compared to that used in best-estimate codes of nuclear reactor safety analysis.

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