大破口喪失水流事故是核能電廠最嚴重的設計基準事故，電廠安全系統必須確保在該事故發生時燃料護套溫度維持在安全限值以下，因此，大破口喪失冷卻水事故模擬對於電廠安全分析，或是分析模型的驗證都相當重要。美國核管會(NRC)是世界上重要核能電廠管制機構，現行法規很多都參考美國核管會所提出的安全理念發展而來，早期為了將安全列第一考量，對於LOCA分析使用比較保守的分析方法，大多遵照美國核管會分析規範10CFR50.46 Appendix K 方法論。隨著電腦科技進步以及核能分析程式的演進，核能的熱水流程式大多都能模擬整個事故暫態，並且能夠模擬更複雜的情況，模擬的結果也相較準確，最佳化估算以及不準度分析方法論逐漸流行。 本研究主要目的是建立核三廠破口喪失冷卻水事故最佳化估算分析方法，使用最佳化程式RELAP5/MOD3.3 patch05版本。分析流程為主要兩部分，首先是基本案例分析探討LBLOCA的熱水流現象，以及爐心熱傳現象對燃料護套溫度影響。在了解LBLOCA爐心熱傳機制後，透過不準度分析更進一步考慮電廠實際運轉中的系統參數偏差，探討其偏差對燃料護套溫度估算的影響。

Large Break Loss of Coolant Accident (LBLOCA) is the most severe design base accident. The safety systems have to ensure the peak cladding temperature below the critical limit during LBLOCA. Therefore, the simulation of LBLOCA is important for NPP safety analysis. NRC is one of the most important regulatory agency and also lay down many regulations. In the past time, conservative approach was widely used. Most of the countries followed the 10CFR50.46 Appendix K methodology for the LOCA analysis. Nowadays, in order to make the analysis results close to the physical reality, best-estimate plus uncertainty (BEPU) approach is getting popular. The main objective of this research is to establish a BEPU approach for Maanshan NPP LOCA analysis using RELAP5/MOD3.3. The analysis method will be divided into two part. First of all, a basic LOCA analysis will be performed to learn the thermal-hydraulic phenomenon during LOCA, and learn how the thermal-hydraulic phenomenon influence the cladding temperature. Second, uncertainty analysis will be performed to consider the deviations in the real situation. Also, sensitivity analysis will be done to learn how the system deviations affect the cladding temperature.

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