本論文之研究目的就是以圍阻體內主蒸汽管路破管為例，使用美國核管會近期主要發展的熱水流安全分析程式TRACE（TRAC/RELAP Advanced Computational Engine）程式以及圖形介面化程式SNAP (Symbolic Nuclear Analysis Package) 分析案例並將其結果與台灣第一座進步型沸水式電廠，龍門電廠的終期安全分析報告 ( Final Safety Analysis Report, FSAR) 進行比較。除此之外，本分析中也將圍阻體內主蒸汽管路破管進行CHF (Critical Heat Flux) 計算方式靈敏度分析以及一維水位追蹤 (Level Tracking) 和一維及三維水位追蹤交叉之靈敏度分析，另外也對通道組件細網格切割和Choke flow參數進行靈敏度分析。模擬結果在破口流率、壓力槽壓力、燃料護套溫度、水位以及緊急爐心冷卻系統注水等表現在龍門電廠圍阻體內主蒸汽管路破管案例之熱水力現象都可以合理的呈現；燃料護套最高溫度為619oC與龍門電廠終期安全分析報告相比約低50oC；靈敏度方面發現CHF計算中，雙相計算上建議使用AECL-IPPE、AECL-IPPE w/ Biasi Correlation及AECL-IPPE w/ CISE-GE Correlation三種方法；水位追蹤靈敏度分析上無論單嘗試一維水位追蹤或是一維及三維兩者交叉水位追蹤皆發現並無具體的不同。

The main purpose of this study is to simulate the main steamline break inside-containment LOCA of Lungmen ABWR by using TRACE under the SNAP interface and then to compare the results of this model by referring to Lungmen FSAR. The most important parameter, peak cladding temperature, calculated from TRACE is 619oC, which is 50 oC lower than that in FSAR. The results of this TRACE model also indicates reasonable thermal-hydraulic phenomena and presents good agreements with FSAR for the main steamline break inside-containment LOCA. Furthermore, sensitivity study of critical heat flux flag, level tracking and choke flow model are also included. The sensitivity studies of critical heat flux flag show that the "Convection Only" option is not recommended, and the other three selections, AECL-IPPE、AECL-IPPE w/ Biasi Correlation and AECL-IPPE w/ CISE-GE Correlation, are proposed. Besides, the sensitivity studies of level tracking show no appreciable significance exists among different models.

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