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    National Tsing Hua University Institutional Repository > 原子科學院  > 核子工程與科學研究所 > 博碩士論文  >  新世代核電廠安全分析軟體(TRACE)在長時間電廠全黑與喪失冷卻水複合型事故核一廠分析模式建立與因應對策之研究


    Please use this identifier to cite or link to this item: http://nthur.lib.nthu.edu.tw/dspace/handle/987654321/83301


    Title: 新世代核電廠安全分析軟體(TRACE)在長時間電廠全黑與喪失冷卻水複合型事故核一廠分析模式建立與因應對策之研究
    Authors: 陳俊宇
    教師: 王仲容
    施純寬
    Date: 2013
    Keywords: 電廠全黑及喪失冷卻水事故
    extended SBO and LOCA
    ECCS
    高壓緊急爐心冷卻系統
    TRACE
    BWR4
    DEG LOCA
    蒸汽驅動注水系統
    Abstract: 從日本福島核電廠事故後,「複合式天然災害導致多重安全系統喪失」是可能發生的,長時間電廠全黑事故會導致現有的低壓緊急爐心冷卻系統喪失功能,本論文探討在電廠全黑時運用衰變熱所產生之蒸汽推動高壓緊急爐心冷卻系統以確保反應爐水位及完整性。因此需要新一代先進、快速且精準的分析程式,廣泛分析各種複合式災害對於核能安全的影響,並對於長時間設備故障會造成事件嚴重程度做更長時間分析,使事故發生現象更清楚呈現,透過此分析後可以有效提供核能發電廠經營者在發生複合式事故初期做更正確的決策。
    TRACE為新世代的最佳估算(Best Estimate)安全分析程式,是美國核管會整合過去發展的分析程式TRAC、RELAP5及RAMONA優點作為未來主要的電廠安全分析程式,具有三維熱水流計算能力、圖形化輸入介面、模組化重要設備及擬態動畫能力,並與大量實驗數據驗證,提供先進、快速且精準的分析工具。
    本論文以福島核電廠同型之核一廠BWR/4做為TRACE模式的建立基礎,從電廠設計資料建立,應用TRACE的各樣模組真實地模擬爐心熱-水流狀態,尤其成功模擬汽水分離器與降流區之間的軸向流動,可更瞭解暫態事件該區域的熱-水流狀態。經與試運轉測試資料及終期安全分析報告(FSAR)驗證,可應用於暫態事件及喪失冷卻水事故分析,利用其快速分析優點,本論文從循環管路雙截斷管喪失冷卻水事故(DEG LOCA)分析,並針對日本福島核電廠類型之複合式天然災害核子事故後,在「複合式電廠全黑及喪失冷卻水事故」維持反應爐水位非常重要的「反應爐心隔離及冷卻系統(RCIC)」蒸汽驅動注水系統,包括無破孔、安全釋壓閥故障及1%, 10%, 100%不同尺寸破孔下採用蒸汽驅動高壓注水的可行替代方案的爐心狀態與策略分析。從本論文研究結果發現在「長時間電廠全黑事故」中反應爐心隔離及冷卻系統(RCIC)注水量僅足以應付反應爐無破孔情況,甚至無法滿足一個安全釋壓閥故障全開時的洩漏量;且反應爐洩壓過程中水位會急遽下降,造成後續注水伴隨的反應爐壓力突升,對於替代性外部低壓注水的有效性大打折扣,增加爐心融毀風險。因此「長時間電廠全黑事故」中善用蒸汽推動高壓注水系統是十分重要的反應爐冷卻對策,希望藉由本論文研究做為未來更完整的複合式緊急安全設備故障事故分析開端,找出潛在的嚴重安全因子,提昇核能安全。
    After the Japanese Fukushima-Daiichi accident, the extreme event beyond the design basis accident is realized to be possible in the combined disasters. The current mitigation strategy of ECCS could fail because the low pressure injection system with electrical pumps will fail in a station blackout accident. To utilize the best of residual steam by the turbine driven pump is a possible alternate mitigation strategy and is analyzed in the paper. An advanced safety analysis tool with fast, accurate and integrated man-machine interface is necessary to analyze more different cases in the extreme accident and provide more safety precautions and more operation strategies for the plant owners. The TRACE code, the latest and advanced best-estimate simulation code, incorporates the four important codes, TRAC-P、TRAC-B、RELAP5 and RAMONA, and a graphic user interface, Symbolic Nuclear Analysis Package (SNAP), to provide a modern thermal-hydraulic analysis tool with fast and integrated inputs, and will become the NRC’s flagship thermal-hydraulic analysis tool in the near future. The TRACE model of Chinshan nuclear power plant with the same BWR/4 reactor of Fukushima-Daiichi NPP is developed, (1) based on the plant design data; (2) consists of different modules to simulate the reactor systems; and (3) analyzes the 3D thermal-hydraulic phenomena through the 3D VESSEL component and more practical thermal-hydraulic phenomena can be analyzed in the downcomer, fuel-assembly reactor core, core bypass, and upper and lower plenum. The Chinshan TRACE model, which has been benchmarked through several transient cases with the Chinshan FSAR report, the start-up data and the transient results of RETRAN data, can be adopted for analyzing both hypothetical transient scenarios and loss-of-coolant accidents, and further more for the alternate mitigation strategies of the extreme accidents of Fukushima-Daiichi type. In this paper, a double-ended guillotine (DEG) break on the recirculation loop is analysis. The Fukushima-Daiichi type accidents, the extended station blackout (SBO) accidents, are evaluated with several scenarios like no break, one SRV stuck open and the various break areas with the 1%, 10%, 100% cross areas of recirculation loop. The current RCIC injection flow rate is not sufficient in a very small break like 1% break area of a recirculation loop and even in the stuck open of a safety/ relieve valve (SRV). The reactor water level will sharply reduce when the reactor pressure is released and result in a fast increase of the fuel temperature. In this situation, the reactor pressure will increase once the coolant being injected that will reduce the effect of the external low pressure injecting system. Thus, the turbine driven pumps, the RCIC pump and the HPCI pump, are one of the important alternate mitigation strategies in the extended SBO. Through this paper, the more advanced analysis on the combined accidents could be performed for the improvement of nuclear safety.
    URI: http://nthur.lib.nthu.edu.tw/dspace/handle/987654321/83301
    Source: http://thesis.nthu.edu.tw/cgi-bin/gs/hugsweb.cgi?o=dnthucdr&i=sGH029713801.id
    Appears in Collections:[核子工程與科學研究所] 博碩士論文

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