- 學位論文
進步型沸水式核能電廠嚴重事故分析
A Study on the Severe Accidents Associated with the Nuclear Power Plant With Advanced Boiling Water Reactor
摘要
雖然核能工業發展行之有年，技術也越見成熟，但核能嚴重事故還是略有所聞。自1979年美國三哩島事件(Three Mile Island)， 至2011年日本福島(Fukushima)第一核電廠事故，核能事故的發生使得民眾對核能電廠漸失信心，為了能提高核能電廠運轉之安全性，核能相關研究單位也對於核能電廠進行更深入的研究，因而開發出許多核能分析軟體。 本論文使用Fauske &Associates, Inc. (FAI)所開發之嚴重事故分析程式MAAP (Modular Accident Analysis Program, MAAP)及美國能源部所屬聖地亞(Sandia)國家實驗室之研究團隊所開發的MELCOR程式，對於龍門電廠進行嚴重事故個案分析。 本論文分析之個案為電廠全黑事故與預期暫態未急停事故，於嚴重事故中MELCOR在計算上較為嚴謹，熔渣留置在未完全破裂之反應槽槽內，再延遲掉落甚至完全留置於槽內而未掉落，此類熔渣行為也較符合物理現象；而於預期暫態未急停事故中，較明顯之差異為爐心水位高度變化，爐心水位高度與爐心功率成正比，爐心功率的差異進而影響到其他現象的發展。 藉由不同程式的比較，更可驗證程式模擬的準確性，以利提供電廠於發生事故初期就做迅速的處置，提升核能電廠之安全性。
並列摘要
Although the nuclear power industry has been in existence and has been developed for many years, with its technology becoming more and more mature over the years, accidents did happen. Ever since the accident at Three Mile Island in 1979, until the compound accident at Fukushima, Japan in 2011, occurrences of severe nuclear accident gradually make people lost their confidence in the nuclear power plants. In order to improve and to evaluate the safety related issues of nuclear power plants, vendors together with numerous institutes for nuclear energy research have done thorough researches in nuclear power plants and have developed many software packages for nuclear safety analyses. This study used the MAAP software which was developed by Fauske & Associates Inc. (FAI) and MELCOR software which was developed by Sandia National Laboratories in the U.S. to analyze postulated severe accidents at the Lungmen Nuclear Power Plant. The cases being analyzed were station blackout (SBO) and anticipated transient without scram (ATWS) accident scenarios. For the study of severe accident, the predominant difference between MAAP and MELCOR is the modeling of the behavior of the core when the core becomes molten. The former assumes that the entire core collapses and migrates to the bottom of the vessel when the core plate fails. However, the latter models the fact that part of the low-power core materials still remain in the core when the core plate fails. For the study of ATWS case, modeling of the reactor power is proportional to the reactor water level, and the difference in reactor power certainly affects the numerous phenomena occurred in the accident. By comparing the results obtained from the two severe accident evaluation software packages used, adequacy of each software was determined. Furthermore, procedures related to the timings and the amounts of cooling water provided were evaluated to determine whether these factors were adequate for the enhancement of the safety of the nuclear power plant under study.