1979年美國賓州三哩島(Three Mile Island)事故;1986年蘇聯車諾比(Chernobyl)事故;2011年日本福島(Fukushima)第一核電廠事故,於此三次核能電廠嚴重發生重大事故後,人民心中對於核電廠的安全性產生了憂慮感。自1980年起,美國電力研究所(Electric Power Research Institute, EPRI)主導並委託Fauske and Associates, Inc. (FAI)公司開發的嚴重事故分析程式(Modular Accident Analysis Program, MAAP)進行核能安全嚴重事故分析。透過MAAP程全分析研究及了解,能評估並發展準確且及時之應變措施,如此得以減輕甚至消除因發生事故產生對人民、環境上安全的問題。 本論文使用Fauske Associates, Inc. (FAI)所開發之嚴重事故分析程式對龍門電廠進行嚴重事故現象與輻射劑量之分析,MAAP最新版本為MAAP5。分析之個案為大破口冷卻水流失事故(Loss of Core Coolant Accident, LOCA),並對於外釋之輻射劑量進行分析。 本文分為兩大部分進行討論;第一部分(第四章)針對FSAR案例之LBLC-PF-R-N事故序列模擬與輻射劑量以MAAP 5.0程式進行模擬分析,事故肇因為主蒸汽管破裂,並且此案例對於核能電廠之廠內、廠外進行劑量分析。第二部分(第五章)主要是MAAP 5.0程式分析嚴重事故是大破口冷卻水流失事(LOCA),探討嚴重事故主要情境,包括三個項目:LOCA-EOP飼水管路破裂直接消防水系統注水(AC-Independent Water Addition, ACIWA)事故發生後加入緊急操作程序書且延遲注水導致爐心熔毀、反應爐壓力槽失效進行模擬,如此便能了解每個階段事故發生時,探討嚴重事故造成廠內、外與輻射劑量外釋之相互關係,最終助於電廠周圍人民疏散時,將有輻射劑量之數值依據。
After the Three Mile Island accident in the United States in1979, the Chernobyl disaster in Ukraine in1986, and Fukushima accidents in Japan in 2011, people worried about the safety of the nuclear power plants. Nuclear power industries have done more thorough researches in nuclear power plants, and developed many nuclear safety analysis codes. Since 1980 American Electric Power Research Institute (EPRI) has authorized Fauske and Associates, Inc. (FAI) to develop Modular Accident Analysis Program (MAAP). MAAP software was developed to analyze the severe accidents of various nuclear power plants. By using MAAP software, the nuclear power plant accidents can be simulated, and the nuclear core conditions can be predicted. This study uses the MAAP 5 code developed by FAI to analyze severe accident of Lungmen Advanced Boiling Water Reactor (ABWR) nuclear power plant with emphasis on the calculation of radiation dose invlved in the large break Less of Coolant Accdientat (LOCA). There are two parts in the case analysis: Chapter4 is analysis related to Final Safety Analysis Report (FSAR) case LBLC-PF-R-N. Chpter5 discusses analysis including (1) LOCA-EOP Feedwater Line Break with (AC-Independent Water Addition , ACIWA), (2) Core Damage due to delayed coolant injection, and (3)Reactor Pressure Vessel Failure. Difference in dose for each scenario is investigated. Also, sensitivity analysis on the timing of coolant injection is studied.