本研究使用由美國桑迪亞國家實驗室研發之嚴重事故程式MELCOR模擬馬鞍山電廠的全黑結合熱端管路潛變破裂事故，以及圍阻體噴灑系統有效性分析。本研究所使用之馬鞍山電廠MELCOR模式為目前最新之2.2版，在版本更新的過程同時更改與新增了數項設定：衰變熱計算方式、蓄壓槽注水設定、管路潛變破裂計算模組、圍阻體噴灑系統，並將更新後的模式利用於民國90年發生之馬鞍山電廠318事件之電廠數據進行驗證比對，隨後再使用已經過驗證之馬鞍山電廠MELCOR2.2模式模擬全黑結合熱端管路潛變破裂事故與分析圍阻體有效性。分析結果顯示在電廠全黑且僅有ACC補水的情形下熱端管路會於全黑後6.5小時發生管路潛變破裂，9.2小時熔融燃料融穿爐底，圍阻體於70.4小時失效。圍阻體噴灑救援則分別於兩個時間點開始：熱端管路破裂、反應爐壓力槽失效，分析結果顯示噴灑能夠有效降低圍阻體壓力，幫助維持圍阻體完整性，但卻會產生更多氫氣與氧氣，造成更多的氫氣燃燒，可能損壞廠內設備，故須謹慎判斷使用時機。

This study focuses on the analysis of a hypothetical ELAP accident with hot leg creep rupture and the mitigation capability of containment spray system by using MELCOR 2.2. The severe accident code MELCOR is developed by Sandia National Laboratories. This study uses the input model that has verified by simulating the SBO accident happened in 2001. The aim of this study is to tell the difference between ELAP without hot leg creep rupture accident and with hot leg creep rupture and verify the mitigation capability of containment spray. In conclusion, hot leg creep rupture happened earlier than RPV failure which made mass of hydrogen release into containment at 6.5 hrs. And because of MCCI, containment failure happened at 70.4 hrs. Two cases of containment spray proved that the mitigation can prevent the containment failure. However, using spray increases the relative concentrations of the hydrogen and oxygen and makes the containment reach combustible conditions and burn.

中文摘要 i
英文摘要 ii
致謝 iii
目錄 v
表目錄 viii
圖目錄 ix
名詞縮寫表 xii
第一章 緒論 1
1.1 研究動機與方法 1
1.2 分析程式介紹 2
1.3 論文架構與模擬案例簡介 3
第二章 文獻回顧 5
2.1 MELCOR程式驗證 5
2.2 潛變破裂相關研究 5
2.3 國內輕水式反應器嚴重事故安全分析與發展 6
2.4 國外輕水式反應器嚴重事故安全分析 7
第三章 熱端管路潛變破裂事故分析方法論 13
3.1馬鞍山電廠MELCOR2.2模式 13
3.1.1 潛變破裂現象與MELCOR設定 13
3.1.2 圍阻體噴灑組件 15
3.1.3衰變熱模擬 15
3.1.4蓄壓槽注水組件 16
3.2 馬鞍山電廠MELCOR2.2模式驗證 17
3.2.1馬鞍山電廠318全黑事件簡介 17
3.2.2馬鞍山電廠318全黑事件模擬方法 18
3.2.3馬鞍山電廠318事件模擬結果 19
3.2.4馬鞍山電廠318事件模擬之結論 20
第四章 電廠全黑造成之熱端管路潛變破裂事故模擬 33
4.1 馬鞍山電廠假設性全黑事故描述 33
4.2 馬鞍山電廠假設性全黑事故分析MELCOR2.2模式 33
4.2.1馬鞍山電廠MELCOR2.2全黑事故模擬方法 33
4.2.2馬鞍山電廠MELCOR2.2版本更新調整 34
4.3馬鞍山電廠假設性全黑事故模擬結果 35
4.3.1未計算潛變破裂之熱水流結果 35
4.3.2熱水流模擬結果 36
4.3.3燃料遷移與氫氣行為 37
4.3.4圍阻體狀態分析 38
4.4 SNAP動畫模式 39
4.5小結 39
第五章 圍阻體噴灑有效性分析 67
5.1圍阻體噴灑救援措施介紹 67
5.2圍阻體噴灑救援模擬方法 67
5.3救援措施模擬結果 68
5.4小結 69
第六章 結論與建議 76
6.1結論 76
6.2建議 76
參考文獻 78
附錄A 發表之國際研討會論文 82

[1] P. Gauntt, D.K., Jeff Cardoni, Jesse Phillips, Andrew Goldmann, Susan Pickering, Matthew Francis, Kevin Robb, Larry Ott, Dean Wang, Curtis Smith, Shawn St.Germain, David Schwieder, Cherie Phelan, “Fukushima Daiichi Accident Study (Status as of April 2012),” Sandia National Lab, 2012.
[2] Sandia National Lab, “MELCOR Computer Code Manuals Vol.3: MELCOR Assessment Problems, Version 2.1.7347,” 2015
[3] S.A. Chavez, J.L. Rempe, “Finite Element Analysis of a BWR Vessel and Penetration under Severe Accident Conditions,” Nuclear Engineering and Design, 1994, p. 413-435.
[4] Sergey Galushin, Pavel Kudinov, “Sensitivity Analysis of the Vessel Lower Head Failure in Nordic BWR using MELCOR Code,” PSAM14, 2018.
[5] Yoshihito Yamaguchi, Jinya Katsuyama, Yoshiyuki Nemoto, Yoshiyuki Kaji, Hiroyuji Yoshida, Yinsheng Li, “Development of Failure Evaluation Method for BWR Lower Head in Severe Accident; High Temperature Creep Test and Creep Damage Model,” Mechanical Engineering Journal, 2017.
[6] Y. Liao, K. Vierow, “MELCOR Analysis of Steam Generator Tube Creep Rupture in Station Blackout Severe Accident,” Nuclear Technology, 2005.
[7] SandiaNL, “State-of-the-Art Reactor Consequence Analysis Project Volume 2: Surry Integrated Analysis,” NUREG/CR-7110, 2012.
[8] Frederick W. Brust, R. Iyengar, M. Benson, Howard Rathbun, “Severe Accident Condition Modeling in PWR Environment: Creep Rupture Modeling,” ASME Pressure Vessels and Piping Conference, 2013
[9] Fabio Giannetti, Fulvio Mascari, A. Naviglio, “Station Blackout Transient Analysis for a PWR like Design by Using the MELCOR code,” 2014
[10] Martina Adorni, Albert Malkhasyan, “Bel V Model Develop and Assessment with MELCOR 2.2,” 10th EMUG, 2018.
[11] Te-Chuan Wang, S. J. Wang, “Analysis of BWR Large Break LOCA with MELCOR1.8.5,” Institute of Nuclear Energy Research, 2005.
[12] 李煒婷, “使用MELCOR程式模擬福島一場三號機事故與放射性物質外釋量之分析,” 清華大學核子工程與科學研究所, 2014.
[13] 張賀嵎, “ 核三廠MELCOR程式輸入檔的建立與電廠全黑嚴重事故分析,” 清華大學核子工程與科學研究所, 2014.
[14] 張靖, “馬鞍山電廠MELCOR2.1模式建立與嚴重事故分析,” 清華大學核子工程與科學研究所, 2017.
[15] 蔣宇, “MELOR2.2程式與沸水式反應器嚴重事故分析之模式建立與應用,” 核子工程與科學研究所, 2018.
[16] Tae Woon Kim, Jinho Song, Vo Thi Huong, Dong Ha Kim, Bo Wook Rhee, Shripad Revankar, “Sensitivity Study on Severe Accident Core Melt Progress for Advanced PWR Using MELCOR Code,” Nuclear Engineering and Design, 2014, p. 155-159.
[17] Thi Huong Vo, Jin Ho Song, “An Analysis of Containment Responses during a Station Blackout Accident,” Journal of Nuclear Science and Technology, 2017
[18] Martina Adorni, Anis Bousbia Salah, Jacques Vlassenbroeck, “Natural Circulation (Interruption) Analysis with MELCOR 2.2 during Asymmetric Cooldown Transients,” NUREG/IA-0507, 2019
[19] Kukhee Lim, Yongjin Cho, Byung Jo Kim, Jeong Seong Lee, In Chul Ryu, “In-Vessel Retention Analysis Using MAAP and MELCOR for a High-Power Reactor,” CSARP / MCAP, 2016.
[20] Yong Jin CHO, “New Severe Accident Research Activities in KINS,” CSARP/MCAP, 2018.
[21] Eunho Kim, “KINS Research Plam for Ex-Vessel Corium Coolability Issue,” C SARP/MCAP, 2018.
[22] Tina Ghosh, “State-of-Art Reactor Consequence Analysis (SORCA) Surry Uncertainty Analysis (UA),” CSARP, 2016.
[23] Retsu Kojo, Akitoshi Hotta, “MELCOR Analysis for Spent Fuel Pools,” CSARP, 2018.
[24] Hao Min Sun, Yasutere Sibamoto, Yuria Okagaki, Taisuke Yonomoto, “Current Status of Aerosol Related Research at JAEA in 2018,” CSARP, 2018
[25] Sandia National Lab, “MELCOR Computer Code Manuals Vol.: Primer and Users’ Guide, Version 2.2.9541,” 2017.
[26] Sandia National Lab, “MELCOR Computer Code Manuals Vo2.: Reference Manual, Version 2.2.9541,” 2017.
[27] 黃凱群, “以TRACE程式進行馬鞍山電廠全黑事故分析與斷然處置策略之探討,” 清華大學核子工程與科學研究所, 2013
[28] 原子能委員會, “核三廠一號機三月十八日喪失廠內外交流電源事件調查報告綜合摘要,” 2001
[29] Larry Humphries, “MELCOR Code Development Status EMUG 2019,” Sandia National Lab, 2019.
[30] American Nuclear Society Standards Committee Working Group ANS-5.1, “American National Standard for Decay Heat Power in Light Water Reactor,” American Nuclear Society, 1979.
[31] 馬鞍山電廠訓練教材
[32] Taiwan Power Company, “Maanshan Nuclear Power Station Final Safety Analysis Report (FSAR),” 1987.
[33] 行政院原子能委員會核能管制處, “「核三廠圍阻體分析方法論」安全評估報告,” 2014
[34] N. Andrews, C. Faucett, S. Belon, C. Bouillet, H.Bonneville, “MELCOR-ASTEC Crosswalk of the Accident at Fukushima-Daiichi Unit 1: Phase I Analysis,” Sandia National Lab, 2019
[35] 鄭惟遠, “馬鞍山電廠MELCOR2.2計算書,” 核能與新能源教育研究協進會, 2019
[36] J. Freixa, V. Martinez-Quiroga, F. Reventos, “Core Exit Temperature Response during an SBLOCA Event in the Asco NPP,” NUREG/IA-0498, 2018