雙相流現象廣泛於運用在各類工程包含化工、能源等其他工業領域，在核工領域實際運用於燃料棒不正常升溫，燃料棒溫度上升至沸騰溫度，壁面溫度與水因沸騰產生氣泡，周圍沸騰狀態對於燃料棒與冷卻水之情況，本實驗室亦致力於採用計算流體力學(Computational Fluid Dynamics，CFD)模擬程式，模擬各類存在於工業之雙相流現象，並藉由實驗數據去修正模擬結果，此方法亦被國際間接受。
本論文著重在實驗部分，目前雙相流現象為難以利用模擬程式預測，故製作此實驗採即有效數據，觀察環形管內之氣泡。除了使用攝影機以影像方式紀錄氣泡之破裂與結合影像，採用雙探針量測方法(Double-sensor probe)，利用電子訊號的方式加以處理，可得到徑向分布的各點局部空泡分率(Local void fraction)、氣泡大小(Bubble diameter)、表象速度(j_f和j_g)，進而計算出介面面積濃度(Interfacial area concentration)，並與攝影機記錄互相比較、驗正，為CFD雙相流模擬的基準(Benchmark)，作為本實驗室模擬研究資料庫，並修正模擬結果。

Bubbly flow phenomenon is the most common flow pattern in nuclear plant system. In this paper, the experiment was designed to observe the fluid phenomenon surrounding nuclear rod bundle, and developed the measurement method in an annulus pipe measured by double sensor probe.
In recent years, double sensor probe is one of the measuring techniques for the bubbly flow. It can measure local void fraction and bubble velocity. After mathematical statistical calculation, interfacial area concentration and sauter mean diameter were obtained. A total of 9 data consisted of four superficial liquid velocities, 0.31, 0.41, 0.51, and 0.61 m/s, and four superficial liquid velocities, 0.02, 0.03, 0.05, and 0.13 m/s. The obtained data will be compared and used to develop to the two phase model for Computational Fluid Dynamics

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