進步型沸水式反應爐嚴重事故之研究

A Study on the Severe Accidents for an Advanced Boiling Water Reactor

陳俊翔(Chun-Hsiang Chen)

指導教授 ： 鄧治東

中原大學/工學院/機械工程學系/碩士(2007年)

https://doi.org/10.6840/cycu200700199

摘要

本研究係針對龍門電廠所採用之進步型沸水式反應爐（Advanced Boiling Water Reactor, ABWR），應用MAAP 4.0.4程式進行電廠全黑事故（SBO）、冷卻水流失事故（LOCA）及預期暫態未急停事故（ATWS）安全分析。研究重點係按照緊急運轉程序書（Emergency Operating Procedures, EOPs）之控制程序模擬上述三大事故，以MAAP 4.0.4程式計算和分析事故
之嚴重性，並分析所得之事故序列表中各現象之差異性。

電廠全黑事故（SBO）於模擬上分為兩個案。
個案一：為一般之電廠全黑事故，係假設所有廠區（On-Site）及廠外（Off-Site）喪失所有交流電源，除了爐心隔離冷卻系統（RCIC）外，所有的冷卻系統均無法啟動，於RCIC跳脫後，反應爐爐心水位迅速下降，致使爐心裸露、過熱、熔毀及重置，最後導致圍阻體過壓保護系統（Containment Overpressure Protection System, COPS）開啟進行洩壓排氣。
個案二：係按照緊急運轉程序書進行模擬設定，於RCIC跳脫後，設定由廠外以手動方式注入消防水至反應爐內，以維持反應爐爐心水位（Level 3~Level 8），反應爐因此維持完整狀態而未失效，但圍阻體最終仍因抑壓池上部區間之壓力值達COPS啟動設定點而開啟進行圍阻體洩壓排氣。

冷卻水流失事故（LOCA），係設定於飼水管路位置破口（破口面積設定為0.07 m2），模擬上分為兩個案。
個案一：為一般之冷卻水流失事故，電廠內所有爐心冷卻水系統均依照系統設定，自動地開啟與關閉。此事故於模擬過程中COPS仍因抑壓池上部區間之壓力值達啟動設定點而開啟，進行洩壓排氣，且爐心並未有裸露現象，故並未有任何惰性氣體及放射性物質因COPS啟動而釋出。
個案二：係按照緊急運轉程序書進行模擬設定，使用高壓爐心灌水系統（HPCF）及低壓爐心灌水系統（LPFL）維持爐心水位高度，並開啟兩組餘熱移除系統（RHR）以維持抑壓池水溫低於33℃（306.15 K）。

預期暫態未急停事故（ATWS），係設定因主蒸汽隔離閥（MSIVs）關斷而引起之預見暫態未急停事故，模擬上分為兩個案。
個案一：控制棒皆無法以自動或手動方式插入達到停機動作，電廠內各冷卻系統，均依照系統設定進行自動開啟與關閉。高壓爐心灌水系統失效跳脫後，主系統之壓力值仍大於低壓爐心灌水系統啟動之設定點，而未能注水，故剩下控制棒驅動冷卻泵系統與爐心隔離冷卻系統，但此兩系統無法達到完全吸收爐心反應熱，致使爐心出現裸露，隨後爐心開始熔毀，圍阻體仍因抑壓池上部區間壓力值達COPS啟動設定點而開啟，致使圍阻體過壓保護系統啟動進行排氣洩壓動作，惰性氣體及揮發性較高之分裂產物（如Cs、I、Te等）將以氣體形式自反應爐釋放而出。
個案二：係按照緊急運轉程序書進行模擬設定。按照EOP-582設定當抑壓池水溫無法維持低於33 ℃時，開啟一串餘熱移除系統進行抑壓池水冷卻動作；且設定當乾井溫度超過55.6 ℃（328.75 K）時，開啟一串噴灑系統維持乾井溫度低於55.6 ℃。於爐心水位控制部分，開啟爐心隔離冷卻系統進行爐心注水動作，維持水位於燃料棒頂端上方203.3 cm與TAF之間，且於指令中設定爐心隔離冷卻系統跳脫後，進行反應爐緊急洩壓，並啟動低壓爐心灌水系統進行爐心灌水動作，進而控制爐心水位於燃料棒頂端下方63.5 cm（最低蒸汽冷卻爐心水位高：MSCRWL）與TAF之間。
本文中針對龍門電廠，所進行探討與研究之電廠全黑、冷卻水流失事故及預見暫態未急停事故之三大基本事故之六個案，於結果中顯示，按照緊急運轉程序書以模擬設定之事故序列，與一般事故之序列比較其結果而言，能減緩事故嚴重性，進而維持反應爐爐心之完整性。因此根據本文之結論，對於電廠內運轉員或是維修人員，於事故發生時，依據緊急運轉程序書（Emergency Operating Procedures, EOPs）採取或啟動任何相關維持反應爐及爐心完整性之動作或搶救措施，如此可避免爐心及反應爐出現裸露（Core Uncovered）或是因高溫熔毀，進而導致重置現象出現（Relocation），甚至避免反應爐失效（RPV Failed）。

關鍵字

冷卻水流失 ； 預期暫態未急停 ； 電廠全黑 ； 緊急運轉程序書 ； 核四廠

並列摘要

The purpose of this study is to evaluate the postulate severe accidents scenarios, including Station Blackout（SBO）, Loss of Coolant Accident（LOCA）, and Anticipated Transient Without Scram（ATWS）, of the Lungmen Nuclear Power Plant（LMNPP）. The methodology uses the MAAP 4.0.4 computer code - （the Modular Accident Analysis Program version 4.0.4）, developed by the Fauske Associates, Incorporated（FAI）, under the contract from the Electric Power Research Institute（EPRI）.
Simulations of severe accidents will be modeled and effects of severe accidents among the cases using the plant-specific Emergency Operator Procedures（EOPs） of LMNPP and the cases without them will be compared and discussed.
The scenarios of SBO accident were divided into two cases.
CaseⅠ: assuming that the plant lost all its on-site and off-site powers, leading to loss all of coolant injection systems. The reactor core isolation cooling（RCIC）system, driven by steam from reactor and battery, is the only coolant injection system available.
Case Ⅱ: most of the accident scenario is same as that of the CaseⅠ before RCIC trips. According to the EOPs, the fire-water（AC-Independent Water Addition, ACIWA）begins to inject into reactor core when the RCIC trips.
The scenarios of LOCA accidents were divided into two cases.
CaseⅠ: the break location is at the elevation of the feedwater pipe, with the break size of 0.07 m2 （doubled-ended break） and ensuing loss of coolant in the reactor core. All coolant injection systems switched （open/close） automatically by the setting of plant.
CaseⅡ：most of the accident scenario is same as that of the CaseⅠ, and according to the EOPs, the high pressure core flooder（HPCF） and low pressure core flooder（LPFL）were assumed to be available manually to inject into the reactor core and maintain the water level（between Level 3 and Level 8）.

In addition, two trains of the residual heat removal（RHR）systems were assumed to be available to maintain the suppression pool temperature below 33 ℃（306.15 K per EOP-582 SP/T）.
The scenarios of Loss of ATWS were divided into two cases.
CaseⅠ：the reactor core scram was assumed to be not available, due to the failures of control rods insertion automatically or manually as well as the stand-by liquid control system. All coolant injection systems switched（open/close）automatically by the setting of plant. For this case, the reactor core generated a large amount steam and power, leading to the reactor core
overheating and then melting down
CaseⅡ：most of the accident scenario is same as that of CaseⅠ, and according to the EOPs, RCIC was assumed to be available to maintain the water level of reactor core (between top-of-active-fuel and above 203.3 cm).
For the cases being analyzed, it was assumed that the RHR systems were available to maintain the drywell temperature of below 55.6 ℃ and the suppression pool temperature below 33 ℃. The LPFL system begins to inject
into reactor core and maintain the water level of reactor core (between
TAF and below 63.5 cm) when the RCIC trips.
The EOPs for the LMNPP were used for the postulate severe accidents scenarios, including SBO, LOCA and ATWS. The results obtained from this study indicated that implementation of the EOPs is useful to reduce the degree of severe accidents and maintain the completeness of reactor core.
Especially, the result obtained form the SBO and ATWS cases with EOPs showed that the operator actions had the potential to bring the plant into a safe-stable state.

並列關鍵字

LMNPP ； EOPs ； SBO ； ATWS ； LOCA

參考文獻

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