本研究使用最佳化熱水流估算程式TRACE成功建立除役期間之金山電廠用過燃料池分析模式，並運用此模型分析除役期間用過燃料池遭遇喪失冷卻水(大破口)且喪失正常冷卻系統假設事故下之熱水流現象，並探討國際報告NEI-0612所建議之救援措施在此事故下之有效性。除此之外，此分析方法與模型亦帶入台灣電力公司所發布之除役計畫、國際上用過燃料池熱水流分析報告(包含NUREG-1738與NSIR-2015-001)之假設與邊界條件分別進行模擬，藉由模擬結果的再現性與比對，驗證該分析方法之有效性與模型可信度。
　　首先本研究使用美國核管會標準審查方案報告ASB 9-2內之衰變熱功率計算公式作為基礎，且根據不同週期之冷卻時間進行計算，以評估金山核能電廠現況之總衰變熱，並將該衰變熱表建立至模型中，以模擬後續嚴重事故，除此之外，亦根據該公式預測未來十年內核能一廠用過燃料池之總衰變熱，以供相關單位參考。在後續事故中，考慮到冷卻水流失事故之模擬結果與破口大小、位置皆有密切相關性，在基本案例中以初始洩露量500gpm以及破口高度約為燃料池底部上方0.93公尺進行模擬。藉由嚴重事故分析之結果，可以估算運轉方在發生事故後能擁有多少餘裕時間準備救援措施。除了基本案例分析以外，也針對初始破口流量，以及鋯水反應氧化計算方式進行靈敏度分析。
　　綜合以上所述之模擬結果，用過燃料池在遭遇喪失冷卻水(大破口)且喪失正常冷卻系統嚴重事故下，仍可擁有長達62小時之餘裕時間可準備後續救援措施，即使於破口流量最嚴重之案例中(約為700gpm)，也仍有59.59小時之準備時間。值得一提的是，由於此模型中沿用了國際上熱水流分析報告之結論與方法論，包含絕熱、燃料束與束之間無熱傳、空氣對流條件較差等保守條件，此模擬結果將具有一定保守度。除此之外，由於金山電廠已停機逾四年，其總衰變熱與剛停機時之衰變熱已不可同日而語，根據後續救援措施之有效性分析中可知，在此餘裕時間內，若能成功執行救援措施，燃料護套之溫度亦能良好地控制在低溫，於整個事故過程皆無升溫至1088.7K。而根據計算鋯水反應氧化熱釋放率關係式之不同，其模擬結果也將有所差異，在靈敏度分析案例中，Baker-Just correlation相對於Cathcart-Pawel correlation之計算結果較為保守，在TRACE模擬中，其開始明顯發生鋯水反應之時間點亦有所提前，在最終達到1088.7K之時間點約提前3小時。

In this research, I used the thermal hydraulic code, TRACE which is developed by US NRC, and coupling with SNAP (a graphical interface code) to establish a Chinshan spent fuel pool model and compare the results with decommission plan announced by Taiwan Power Company (Taipower). Based on this model, the severe accident including station blackout (SBO), loss of coolant water accident (LOCA) had been analyzed in this study. Besides, the capability of the mitigation strategy which is recommended by the NEI-0612 report to deal with the sever accident is evaluated in this research. In order to proof the accuracy of this methodology in establishing the heat transfer model, the same assumption of decommission plan and other analysis report of spent fuel pool (including NUREG-1738 and NSIR-2015-001) are used to rebuild a same model with the same boundary conditions. The simulation results with the same assumption shows that the prediction of TRACE are consistent with the results of the decommission plan and analysis reports. And these results implies the TRACE/SNAP model that I established in this research has a respectable accuracy.
　　First of all, this study used the calculation formula of decay heat in the ASB 9-2 standard review plan report of the US Nuclear Regulatory Commission as the basis to discuss the decay heat of the spent fuel rod and calculate it according to the cooling time of different cycles to evaluate the total decay heat of Chinshan Nuclear Power Plant spent fuel pool, and brought this decay heat table into the model to simulate subsequent serious accidents. In addition, according to the formula, the total decay heat of the spent fuel pool is also predicted in the next ten years.
　　The analysis results show that even if the extreme accident that SBO and LOCA accident occurred simultaneously, the temperature of spent fuel cladding will not exceed 1088.7K for about three days. It is due to the total decay heat is very low compare with the decay heat just after nuclear power plant shutdown. However, as long as we take the mitigation strategy to spray the water above the top of the fuel in such server accident within three days, the spent fuel pool can restore to a safe state.

摘要 i
ABSTRACT iii
致謝 v
表目錄 ix
圖目錄 x
名詞縮寫表 xiii
符號表 xiv
第一章 緒論 1
1.1研究動機與方法論 1
1.2論文架構 2
第二章 文獻回顧 4
2.1國內外用過燃料池安全分析 4
2.2 TRACE模擬程式文獻回顧 14
2.3用過燃料束之衰變熱計算方法文獻回顧 20
第三章 模擬程式與電廠介紹 27
3.1 金山電廠 27
3.2 除役期間金山用過燃料池 27
3.2.1 用過燃料池簡介 27
3.2.2 事故救援設備介紹 28
3.3 TRACE熱水流分析程式 29
3.3.1 熱水流理論介紹 30
3.3.2 降幅水位計算模型 32
3.3.3 熱傳導模型理論介紹 34
3.4 SNAP 圖形化介面程式 35
第四章 建立除役期間用過燃料池分析模式 36
4.1 TRACE分析模式建立 36
4.2 用過燃料池分析模式驗證 51
4.2.1 國際安全分析文獻再現性 51
4.2.2 金山核能電廠除役計畫熱水流分析再現性 60
4.3 金山核能電廠除役期間燃料棒衰變熱計算 64
第五章 金山核能電廠暫態案例安全分析 70
5.1 除役期間用過燃料池喪失長期冷卻能力與冷卻水流失事故 70
5.1.1 基本介紹與初始條件 70
5.1.2 案例分析結果 70
5.2 除役期間用過燃料池發生嚴重事故且實施後續救援措施 74
5.2.1 NEI-0612 救援措施介紹 74
5.2.2 案例分析結果 77
5.3 靈敏度分析 79
5.3.1 初始破口流量案例分析 79
5.3.2 氧化熱計算公式案例分析 82
第六章 結論與未來建議 85
6.1 結論 85
6.2 未來建議 86
參考資料 88

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