馬鞍山電廠為三迴路壓水式反應器之核能電廠，其100%爐心熱功率為2822MWt。主蒸汽管路破裂事故（Main Steam Line Break, MSLB）為馬鞍山電廠終期安全分析報告（Final Safety Analysis Report, FSAR）核能電廠設計基準事故（Design Basis Accident, DBA）中一項非常重要的安全分析。當電廠發生主蒸汽管路破裂事故時，核能電廠之安全系統必須確保核能系統有足夠的能力以維持電廠保持安全餘裕。本研究所使用之程式為美國核能管制委員會（United State Nuclear Regulatory Commission, U.S. NRC）資助下由美國愛德荷國家工程實驗室（Idaho National Engineering and Environmental Laboratory, INEEL）所發展的熱水流安全分析程式RELAP5/MOD3.3（Reactor Excursion and Leak Analysis Program）和圖形化介面程式SNAP（Symbolic Nuclear Analysis Package）。在過去，針對核能電廠設計基準事故或嚴重事故時較常以保守的分析方法進行分析，以模擬核能系統內之一次側壓力、二次側壓力、燃料護套溫度、水位等結果。隨著計算機的進步與熱水流安全分析程式之演進，國際上開始逐漸使用最佳估算分析與不準度分析來進行嚴重事故暫態分析，不僅能夠模擬更複雜的熱水流現象，並且可以模擬出較為準確的結果。本研究首先建立RELAP5/MOD3.3 MSLB分析模式以探討各熱水力結果於MSLB事故基本案例下對於電廠之影響，並探討沖放係數與臨界流模式對於結果之靈敏度分析。緊接著以SNAP圖形化界面搭配DAKOTA不準度安全分析程式進行不準度分析，探討於主蒸汽管路破裂事故下邊界條件與模型不準度參數偏差下對於熱水流現象的影響。最後在不準度分析結果中，利用皮爾森相關係數理論對選取之不準度參數估算其相關程度。結果顯示模型不準度參數對於結果有著顯著的影響。本研究成功建立了RELAP5/MOD3.3耦合圖形化介面程式SNAP與不準度分析程式DAKOTA進行MSLB事故下之不準度分析，並透過Python進行數據後處理，以便往後不準度研究者快速且正確地擷取結果。

　　The Maanshan nuclear power plant is a three-loop PWR. Its rated core thermal power is 2822 MWt. The Main Steam Line Break accident（MSLB）is one of the Design Basis Accidents（DBAs）in Maanshan Final Safety Analysis Report（FSAR）. When MSLB occur, the safety systems in nuclear power plant must be secure adequate safety margin. In this thesis adopts RELAP5/MOD3.3, an advanced thermohydraulic code developed by Idaho National Engineering and Environmental Laboratory（INEEL）, combined with SNAP, an effective and convenient CGI code to establish Maanshan Nuclear Power Plant RELAP5/SNAP/DAKOTA model for MSLB safety analysis. In the past time, conservative analysis approach was widely used. Most of the countries followed the 10 CFR 50.46 Appendix K methodology for the safety analysis. Today, in order to bring the analysis results closer to physical phenomena. The Best-Estimate Plus Uncertainty （BEPU） analysis approach is becoming more popular. The MSLB accident research will be divided into two part, each of which presents the results relating to one of the research questions. First part of the thesis research the physical phenomena which emerged from the basic MSLB accident analysis. Second part will complete uncertainty concern will be performed in this research. In addition to the boundary condition uncertainty, model uncertainty related to two phase flow model in RELAP5 code are further considered. RELAP5 code had developed seven interfacial drag coefficient uncertainty factors, including bubbly flow drag, slug flow drags, annular-mist flow drags, dispersed flow drag, re-flood drag and two-phase friction and form loss. With randomly adjusting this developed two-phase flow drag factors, the model uncertainty can be also considered. After realizing the effect of these two-phase flow model uncertainties, a completed simulation model which considers both the boundary condition and model uncertainty will be developed. As a result, the predictions of the simulation model may become closer to the real world. At last, with the RELAP5/MOD3.3, DAKOTA and SNAP interface were combined successfully to perform the MSLB uncertainty analysis. In addition, this thesis adopts Python to capture uncertainty data at the same time.

摘要 ii
Abstract iii
誌謝 iv
目錄 v
表目錄 viii
圖目錄 ix
緒論 第一章 1
1.1 前言 1
1.2 研究動機與目的 1
1.3 核能安全分析之必要性 2
1.4 論文架構 3
第二章 文獻回顧與研析 4
2.1 馬鞍山核電廠簡介 4
2.2 國際間MSLB案例安全分析研究研析 4
2.3 不準度分析方法論 6
2.3.1 不準度分析理論 6
2.3.2 決定不準度參數 7
2.3.3 不準度區間和參數分布之選擇 7
2.3.4 模擬次數 8
2.3.5 隨機抽樣方法 9
2.3.5.1蒙地卡羅抽樣（Monte-Carlo Sampling）[14] 9
2.3.5.2 拉丁超立方抽樣 （Latin Hypercube Sampling）[15, 16, 17] 10
2.4 相關係數理論 10
2.4.1 皮爾森積差相關係數理論 10
2.4.2 斯皮爾曼等級相關係數 11
第三章 分析程式介紹 24
3.1 RELAP5/MOD3.3熱水流分析程式 24
3.1.1 RELAP5程式簡介 24
3.1.2 RELAP5/3.3雙相流統御方程式 24
3.1.3 RELAP5臨界流模式 27
3.1.4 熱傳模式 31
3.2 SNAP圖形化介面程式 33
3.3 DAKOTA程式[41] 34
第四章 RELAP5/MOD3.3 MSLB模式建立 47
4.1 核能系統[11] 47
4.2 熱水流組件與熱結構組件 49
4.3 電廠控制元件 50
4.4 反應器功率計算 51
第五章 馬鞍山核電廠RELAP5/MOD3.3主蒸汽管路破裂事故基本案例分析 58
5.1 主蒸汽管管路破裂事故案例 58
5.2 MSLB初始條件假設 59
5.3 MSLB基本案例分析結果 60
5.4 靈敏度分析 61
5.4.1 沖放係數靈敏度分析 61
5.4.2 臨界流模式靈敏度分析 62
5.5 主蒸汽管路破裂事故結論 63
第六章 最佳化估算於不準度分析 76
6.1 不準度分析 76
6.2 不準度分析結果 78
第七章 結論與建議 93
7.1 結論 93
7.2 建議 94
參考文獻 95
附錄 99

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