- 學位論文
核一廠嚴重事故序列之研究
A Study on Severe Accident Sequence Analyses for Chin-Shan Nuclear Power Station
摘要
本研究主要針對核一廠所採用之沸水式反應爐，應用MAAP 4.0.4程式進行參數檔與輸入檔的建立以及電廠全黑、冷卻水流失、預見暫態未急停事故的安全分析。研究重點係針對核一廠於嚴重核子事故下之爐心熔毀及圍組體反應等狀況，以MAAP 4.0.4程式分析事故之嚴重性，並將事故中現象之差異作一番探討。 全黑事故事故的模擬中係假設所有廠區(On-Site)及廠外(Off-Site)的交流電皆無法供應，除了爐心隔離冷卻系統外，所有的注水系統均無法執行注水的功能；在冷卻水流失事故的模擬中係假設所有爐心緊急冷卻系統皆關閉，而主蒸汽隔離閥及飼水泵之開關，則由程式自行計算關閉時間。破口位置係設定於再循環管路，破口面積則係假設六個面積，分別為0.1 ft2、0.3 ft2、0.5 ft2、0.7 ft2、1.0 ft2、2.1795 ft2。 而於預見暫態未急停事故的模擬中係假設反應爐無法急停，所有控制棒皆無法自動或是以手動方式插入，同時備用硼液控制系統亦失效，導致事故因爐心持續產生大量能量而快速惡化。最後致使反應爐及圍阻體失效排氣，分裂產物因而外釋。 在預見暫態未急停事故中，同時也針對操作員動作加以模擬，期望於操作員的救援動作下，惡化之情形能得以舒緩。
並列摘要
The purpose of this study is to evaluate the postulated severe accident scenarios - such as station blackout, loss-of-coolant accidents (LOCAs), and anticipated transient without scram (ATWS) - for the Chin-Shan Nuclear Power Station using the Modular Accident Analysis Program (MAAP) version 4.0.4. For these accident scenarios, the behaviors of reactor core and containment, and the release of fission products were analyzed. In addition, the phenomena associated these scenarios were discussed. The station blackout scenario assumed that the plant lost all its on-site and off-site power, leading to loss of all coolant injection capabilities, except the reactor core isolation cooling (RCIC) system that is driven by the steam provided by the reactor. For the LOCA scenarios, all coolant injection systems were assumed to be lost and the break location was assumed to be at the piping connecting recirculation pump to the reactor vessel, with the break sizes of 0.1, 0.3, 0.5, 0.7, 1.0, and 2.1795 (double-ended, guillotine-type break) ft2. For the ATWS scenario, the reactor scram was assumed to be not available, due to the failures of automatic and manual control rod insertion as well as the stand-by liquid control system. In this scenario, the reactor core became degraded rapidly due to the elevated core power generated. For these types of scenarios, actions taken by the operators were analyzed to determine their impacts on the progression of the accidents. Without adequate core cooling and/or containment heat removal, the reactor core heated up, melted, and then relocated to the vessel bottom head. In the meantime, substantial amount of hydrogen resulting from the metal-water action in the core region was generated. Due to the decay heat associated with the core debris (or so-called corium), the molten corium continually heated up and melted through the bottom of the vessel. The molten corium that located at the lower drywell again heated up, interacted with the concrete, and generated additional non-condensable gases. The gases pressurized the wetwell gas space, leading to venting of the containment through the hard-pipe vent. Following containment venting, the fission products were released to the environment. Results of this study indicated that the progressions of the accident scenarios were affected by the availability of the coolant injection systems and the containment heat removal systems, and the reactions taken by the operators. In addition, the models implemented in the MAAP 4.0.4 compared to those of the MAAP 3B had significant effects on the timing of the failure of the core plate and the melt-through of the vessel bottom head. Furthermore, the values used in the decontamination factor had a major impact on the amount of the release of the fission products following containment venting.