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研究生: 虞崇瑞
Yu, Chong-Ruei
論文名稱: 馬鞍山核三廠圍阻體的安全性分析
Safety Analysis of Nuclear Power Plant Containment at Maanshan
指導教授: 胡宣德
Hu, Hsuan-Teh
學位類別: 碩士
Master
系所名稱: 工學院 - 土木工程學系
Department of Civil Engineering
論文出版年: 2017
畢業學年度: 105
語文別: 中文
論文頁數: 97
中文關鍵詞: 預力混凝土圍阻體極限耐壓能力ABAQUS地震力
外文關鍵詞: pre-stressed concrete containment, ultimate pressure strength, ABAQUS, dynamic implicit, seismic analysis
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  • 核能發電一直在世界上占有在相當重要的一環,目前在世界上所佔的發電百分比有10.8%,而台灣由於天然資源的缺乏,核能發電更是不可或缺的電力來源,民國100年台灣發電來源,有12.4%來自核能發電,因此現階段來說核能發電在目前的台灣依然是不可或缺的一部分,核能發電除了比再生能源發電,提供較為穩定、高效率的發電模式,也不像火力發電產生大量二氧化碳,造成溫室效應的問題,與節能減碳的趨勢不符合,但核能雖然經濟便利,也有其最主要的缺點,那就是輻射污染的問題,當核反應爐冷卻功能失效的時候(Loss-of-Coolant-Accident)導致內壓上升或者是強烈地震的情況,要如何確保核電廠圍阻體結構是否安全,以及輻射是否會外洩等問題產生。
    因此本文利用有限元素法軟體ABAQUS,來對馬鞍山核電廠的圍阻體進行分析,本文主要可分為兩個部份,第一部分是模擬其極限耐壓能力,第二部份受地震力的影響及破壞模式。由於在耐壓分析的部份由於已有關於馬鞍山核電廠耐壓分析的相關研究,因此第一部份主要在嘗試以不同分析方法並且加以驗證本文模形之正確性,並且對於其內部承壓模式的基本變形模式以及破壞模式做探討,第二部份則是利用PGA=0.65g的集集地震歷時,以了解設計PGA為0.4g的馬鞍山核電廠圍阻體的運動行為及破壞情形…等,並且探求馬鞍山為阻體的自然震動頻率加以比較,加以檢驗馬鞍山發電廠的安全性是否無虞。

    Nuclear power generation has been occupied a very important part of the current share of the world's power generation, having 10.8% of total power generation in the world. Due to the lack of natural resources in Taiwan, nuclear power generation is an indispensable source of electricity, however; nucler power generation has a vary important shortcoming, that is, radiation pollution problem. When the nuclear reactor cooling function is failure (Loss-of-Coolant-Accident), it will lead to an increase in internal pressure to possible explosion.
    Therefore, this paper uses the finite element model ABAQUS to analyze the containment of Maanshan nuclear power plant. This paper can be divided into two parts. The first part is to simulate the ultimate pressure resistance, and the second part is the behavior affected by the earthquake. Since the part of the pressure analysis has been already been studied in our research team, the first part is mainly trying to differentiate the method and verify the correctness of the model, and for its basic deformation to do the discussion. The second part is the use the chichi earthquake which PGA is 0.65g to simulate, in order to understand the behavior of maanshan containment that its design is for PGA 0.4g. In addition, the natural vibration frequency of Maanshan has been simulated and then compared to the the simulation that containment under seismic load, for making sure the Maanshan containment is safe .

    摘要 i 誌謝 v 目錄 vi 表目錄 x 圖目錄 xi 第1章 緒論 1 1.1 研究動機 1 1.2 研究目的 2 1.3 本文內容及架構 2 第2章 核能發電廠與圍阻體的背景介紹 4 2.1 核能發電廠背景介紹 4 2.2 圍阻體的基本介紹 7 2.3 馬鞍山圍阻體的構造 8 第3章 圍阻體的材料行為 11 3.1 鋼筋 11 3.2 預力鋼鍵 13 3.3 混凝土 14 3.3.1 混凝土的單軸材料行為 14 3.3.2 混凝土雙軸行為 18 3.3.3 混凝土的三軸行為 19 3.3.4 混凝土的材料組合率 21 3.4 鋼櫬材料特性 25 第4章 ABAQUS對圍阻體之數值模擬 27 4.1 Concrete Smeared Cracking 27 4.1.1 混凝土的開裂 27 4.1.2 混凝土抗壓彈塑性論 29 4.1.3 混凝土開裂檢測 34 4.2 Concrete Damaged Placticity(CDP) 39 4.2.1 單軸循環行為 42 4.2.2 塑性行為 42 4.2.3 拉力行為 43 4.2.4 壓力行為 43 4.3 圍阻體模型的建立 44 4.3.1 單位系統 44 4.3.2 圍阻體模型簡述 45 4.3.3 材料參數 51 4.3.4 分析步驟與邊界條件 53 4.3.5 施加的地震歷時 54 第5章 分析與討論 57 5.1 預力分析 57 5.2 受壓分析 58 5.3 圍阻體自然震動頻率分析 63 5.4 地震歷時分析 66 5.4.1 底板鋼性及非鋼性的比較 66 5.4.2 等效塑性應變分布圖 67 5.4.3 樓板反應譜 70 5.4.4 最大加速度分布圖 78 5.4.5 最大位移分布圖 79 5.4.6 最大剪力應變分布圖 81 5.4.7 破壞情形 81 5.4.8 C3D8R及C3D8I的比較 84 第6章 結論與建議 86 6.1 結論 86 6.2 未來建議 87 參考文獻 88 附錄 核電廠的設計圖摘錄 92

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