冷卻水流失事故(Loss of Coolant Accident, LOCA)為核電廠安全分析中設計基準事故(Design Basis Accidents, DBA)之一，其有可能造成圍阻體系統失效的，而圍堵不住放射性物質使其外釋到周遭生物圈中。基此考量，本研究之首重工作為針對核三廠FSAR中之圍阻體安全個案分析，選取主蒸汽管斷管事故與冷端管路冷卻水泵吸水側斷管失水事故為分析重點，確保圍阻體能負荷失水事故發生時所造成之圍阻體內壓力與溫度尖值之衝擊。
本研究採用GOTHIC程式建立核能三廠圍阻體短程分析模式：在短程分析模式中，採用核能三廠之終期安全分析報告(Final Safety Analysis Report, FSAR) [1]所提供的沖放質能數據作為邊界條件。當反應器功率為102%和25%額定熱功率條件之下，進行主蒸汽管斷管以及冷端管路斷管事故等分析與其靈敏度分析，瞭解圍阻體內壓力及溫度時變趨勢及尖值發生點，最後將GOTHIC程式計算結果與FSAR之分析結果進行比對評估，並探討其合理性。
本研究分析結果與FSAR之分析結果在時變趨勢上相當吻合；與FSAR比較在溫度趨勢上，本研究大多變化較為緩和。失水事故分析所得溫度、壓力尖值與其設計值均有餘裕存在。根據本研究分析結果可得知，在單一事故假設下，圍阻體系統結構之完整性在發生主蒸汽斷管事故及冷端管路斷管事故下應可安全無虞。

The loss of coolant accident (LOCA) is an important design basis accidents (DBA) for the nuclear power plants, because it may challenge the containment structure integrity and let the fission products released to the environment. This thesis has investigated the main steam line break and the pump suction rupture of the cold leg for the Maanshan nuclear power plant in which the large dry containments are utilized. The containment must withstand the peak pressure and temperature caused by the LOCA.
Containment is one of the major barriers against the release of radioactive substances to the environment. Therefore, it’s integrality during the LOCA must be ensured. The analyses performed in this study are the main steam line breaks with 102% and 25% rated thermal power, and the pump suction rupture of the cold leg of Maanshan plant. Sensitivity studies are also presented. GOTHIC is used for the containment analyses, and the blowdown mass and energy release data provided by the final safety analysis report (FSAR) of Maanshan plant are used as the boundary conditions. Short-term analyses for the pressure and temperature responses of the Maanshan’s large dry containment are performed in this study.
In conclusions, the calculated results of the GOTHIC analyses are consistent with the FSAR results. The peak values of pressure and temperature are lower than the design values during the DBA LOCAs. The containment integrity during LOCA can be maintained.

1.Taiwan Power Company, “Final Safety Analysis Report”, Maanshan Nuclear Power Station Units 1 and 2, Amendment No. 32, October 1999.
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3.洪振育，「龍門電廠ABWR圍阻體熱水流分析模式建立」，國立清華大學工程與系統科學研究所，碩士論文，民國99年。
4.彭柏皓，「核能二廠Mark-III圍阻體熱水流分析模式建立」，國立清華大學核子工程與科學研究所，碩士論文，民國100年。
5.Electric Power Research Institute, “GOTHIC: Containment Analysis Package User Manual”, version 7.2a, NAI 8907-02 Rev. 17, January 2006.
6.台灣電力公司第三核能發電廠，「壓水式反應器核能電廠訓練教材」，台灣電力公司，民國84年10月。
7.Electric Power Research Institute, “GOTHIC: Containment Analysis Package Technical Manual”, version 7.2a, NAI 8907-06 Rev. 16, January 2006.
8.Electric Power Research Institute, “GOTHIC: Containment Analysis Package Qualification Report”, version 7.2a, NAI 8907-09 Rev. 9, January 2006.