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研究生： 陳宇民
Chen, Yu-Min
論文名稱： 量化核三廠功率提升後中破口冷卻水流失事故之安全餘裕與爐心熔損頻率
Effect of Power Uprate on Safety Margin and Core Damage Frequency in a MBLOCA
指導教授： 李敏
Lee, Min
陳詩奎
Chen, Shih-Kuei
口試委員: 梁國興
Liang, Kuo-Shing
馮玉明
Ferng, Yu-Ming
陳紹文
Chen, Shao-Wen
學位類別： 博士
Doctor
系所名稱： 原子科學院 - 核子工程與科學研究所
Nuclear Engineering and Science
論文出版年： 2018
畢業學年度： 106
語文別： 中文
論文頁數： 65
中文關鍵詞： 安全度評估 、人因失效機率 、不準度分析 、功率提升 、安全餘裕 、爐心熔損頻率
外文關鍵詞： Probabilistic safety assessment, Human error probability, Uncertainty analysis, Power uprate, Safety margin, Core damage frequency
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    • 本論文研究參考Nuclear Energy Agency所屬的Committee on the Safety of Nuclear Installations (CSNI) 領導之Safety Margin Action Plan (SMAP) 以及Safety Margin Application and Assessment (SM2A)計畫，所發展的安全餘裕縮減 (Safety Margin Reduction，SMR) 與爐心熔損頻率（Core Damage Frequency，CDF）量化的方法論;此方法論通稱為Risk Informed Safety Margin Characterization (RISMC);量化台灣電力公司核能三廠功率提升對SMR與CDF的衝擊，期望能夠減少過去傳統分析方法加諸的保守性，更準確的量化SMR與CDF。台灣電力公司核能三廠採用美國西屋公司設計的三迴路壓水式反應器。分析事故為中破口冷卻水流失事故（Medium Break Loss Of Coolant Accident，MBLOCA）。本研究利用 RELAP5-3D/K 程式量化功率提升前後之安全餘裕的差異，將程式輸入參數的不準度納入考量後，估算事故發生後之最高燃料棒護套溫度 (Peak Cladding Temperature, PCT) 的機率分布函數 (Probability Density Function, PDF)，量化功率提升所造成的安全餘裕縮減。
    功率提升亦會改變爐心熔損頻率，結合安全度評估方法 (Probabilistic Safety Assessment, PSA)，SM2A提供了量化CDF改變的方法論。本研究找出中破口冷卻水流失事故肇始事件中，對於CDF影響最大的人為誤失事件，設定人為動作執行時間的機率分佈，結合對REALP5-3D/K 結果有重大影響的參數的機率分布，以蒙地卡羅取樣，進行多次程式計算，依事件樹頂端事件的成功準則決定該人為誤失事件發生的機率，並重新計算爐心熔損頻率。估算結果顯示，功率提升後安全餘裕縮減了4.2%，MBLOCA肇始事件的爐心熔損頻率提高16%。

    Nuclear Energy Agency of OECD has developed a methodology in the Safety Margin Action Plan (SMAP) and in the Safety Margin Application and Assessment (SM2A) to quantify the reduction of safety margin and the increase of core damage frequency (CDF) due to power uprate. The methodology combines the probabilistic and deterministic approaches of safety analyses. In the present study, the SMAP methodology is adopted to quantify the reduction of safety margin of a 5% power uprate of Maanshan nuclear power plant (NPP) of Taiwan Power Company. The sequence selected to demonstrate the methodology is medium break cold leg loss of coolant accident. An evaluation model code, RELAP5-3D/K is used in the LOCA analysis. The plant specific probabilistic safety assessment (PSA) model is used to determine the accident scenario analyzed. Uncertainty of the predicted peak cladding temperature (PCT) during the accident is quantified using the rank statistics. Phenomenon Identification and Ranking Table (PIRT) for parameters that are important for the medium break LOCA analyses and their distribution are identified. Two sets of calculations are performed to determine PCT of the 95th percentile with 95% of confidence at 102% and 105% power cases. When the best estimated value is determined non-parametrically, the safety margin before SPU is 10.17 and after SPU is 10.42. The margin increases slightly after SPU. From these results, it can be concluded that the impact of SPU on the safety margin of PCT is very insignificant. For CDF, there were two operator actions in the accident mitigation for the sequence selected. These actions were the emergency cooldown and depressurization (CND) of RCS and the lineup the high-head safety recirculation (HHSR) when RWST was empty. A new approach was proposed to quantify the failure probability of the operator actions. The time of operator to execute the action was included in the uncertainty analyses of code simulations. The branch probability of the event tree heading involved the operation action was determined based on the success criteria of the heading. The simulation results showed that the CND failure probability increased by 23% after the power uprate. The sequence CDF increased by 16% after the power uprate.

    摘要...............................................................i 英文摘要...........................................................ii 誌謝..............................................................iii 目錄...............................................................iv 表目錄.............................................................vi 圖目錄............................................................vii 第一章緒論...........................................................1 1.1 前言............................................................1 1.2 PRA發展歷史......................................................1 1.3 研究目的.........................................................4 1.4 論文架構.........................................................5 第二章 文獻回顧......................................................6 2.1 SMAP方法論介紹與文獻回顧..........................................6 2.2 SM2A方法論介紹與文獻回顧.........................................10 第三章 RELAP5程式介紹...............................................13 3.1 RELAP程式發展...................................................13 3.2 RELAP5-3D/K模式介紹.............................................13 3.3 RELAP5-3D/K核三廠輸入檔介紹.....................................14 3.4功率提升介紹.....................................................15 第四章 安全餘裕縮減.................................................19 4.1 安全餘裕縮減分析之方法論.........................................19 4.1.1 安全餘裕概念介紹............................................19 4.1.2 Aleatory不準度.............................................20 4.1.3 Epistemic不準度............................................22 4.2 安全餘裕分析結果與討論...........................................25 4.2.1 中破口冷卻水流失事故暫態案例分析..............................25 4.2.2 燃料護套最高溫度的分佈情形與95/95th統計結果...................27 4.2.3 量化功率提升後的安全餘裕縮減.................................28 4.2.4 探討高壓注水系統與輔助飼水系統的影響..........................28 4.2.5 重要度分析..................................................29 4.3 安全餘裕縮減結論................................................36 第五章 爐心熔損頻率.................................................38 5.1 爐心熔損頻率分析之方法論.........................................38 5.1.1 事故序列選擇與探討..........................................38 5.1.2 如何決定人為動作的分佈.......................................40 5.1.3 中破口冷卻水事故中的不準度分析................................42 5.2 爐心熔損分析結果與討論...........................................46 5.2.1 中破口冷卻水流失事故現象討論.................................46 5.2.2 不準度分析結果..............................................50 5.2.3 量化多重人為動作下的爐心熔損頻率..............................55 5.3 爐心熔損分析結果與討論...........................................58 第六章 結論.........................................................59 參考文獻............................................................61 縮寫表.............................................................64

    [1] “Safety Margins Action Plan -– Final Report”, NEA/CSNI/R(2007)9.
    [2] “Safety Margin Evaluation – SMAP Framework Assessment and Application”, NEA/ CSNI/R, (2011)3.
    [3] https://www.nrc.gov/about-nrc/regulatory/risk-informed/history.html
    [4] D.R. Bradley, “Overview of PSA Development and Risk-Informed Application in the U.S.”, Level 2 PSA and SAMG Workshop Beijing, China, 2016.
    [5] S.K. Chen, “The History Perspective of PRA”, PRA II Course Lecture, NTHU NES, 2017.
    [6] “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory”, NRC/60FR 42622, 1995.
    [7] C.C. Chao,“The History and Methodology of PRA”, PRA I Course Lecture, NTHU NES, 2017.
    [8] C. Queral, J. Montero-Mayorga, J. Gonzalez-Cadelo, G. Jimenez. “AP1000 Large-Break LOCA BEPU analysis with TRACE code”, Annals of Nuclear Energy 85 (2015) 576-589.
    [9] T.H. Liang, K.S. Liang, C.K. Cheng, B.S. Pei, E. Patelli, “Risk-Informed Analysis of the Large Break Loss of Coolant Accident and PCT Margin Evaluation with the RISMC Methodology”, Nuclear Engineering and Design 308 (2016) 214-221.
    [10] R.R. Sherry, J.R. Gabor, S.M. Hess, “Pilot Application of Risk Informed Safety Margin Characterization to a Total Loss of Feedwater Event”, Reliability Engineering and System Safety 117 (2013) 65–72.
    [11] T.W. Kim, V.N. Dang, M.A. Zimmermann, A. Manera, “Quantitative Evaluation of Change in Core Damage Frequency by Postulated Power Uprate: Medium-Break Loss-of-Coolant-Accidents”, Annals of Nuclear Energy 47 (2012) 69-80.
    [12] D.H. Lee, H.G. Lim, H.Y. Yoon and J.J. Jeong. “Improvement of the LOCA PSA Model using a Best-Estimate Thermal-Hydraulic Analysis”, Nuclear Engineering and Technology, Vol.46 No.4 Aug. 2014.
    [13] D.A. Dube, R.R. Sherry, J.R. Gabor, S.M. Hess. “Application of Risk Informed Safety Margin Characterization to Extended Power Uprate Analysis”, Reliability Engineering and System Safety 129 (2014) 19-28.
    [14] D.R. Karanki, T.W. Kim, V.N. Dang, “A Dynamic Event Tree Informed Approach to Probabilistic Accident Sequence Modeling: Dynamics and Variabilities in Medium LOCA”, Reliability Engineering and System Safety 142 (2015) 78-91.
    [15] F. D’Auria, C. Camargo, O. Mazzantini, “The Best Estimate Plus Uncertainty (BEPU) Approach in Licensing of Current Nuclear Reactors”, Nuclear Engineering and Design 248 (2012) 317–328.
    [16] A. Petruzzi, M. Cherubini, M. Lanfredini, F. D’Auria, and O. Mazzantini, “The BEPU Evaluation Model with RELAP5-3D for the Licensing of the Atucha-II NPP”, Nuclear Technology, Vol. 193 (2016) 113–160.
    [17] “Appendix K to Part 50 – ECCS Evaluation Models”, U.S. NRC, March, 2013.
    [18] “Nuclear Plant Power Uprate Safety Analysis Preliminary Report”, Taiwan Power Company, June, 2009.
    [19] “Living Probabilistic Risk Assessment of Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, Vol. 1, 3-1~3-18, Dec 1995.
    [20] “Emergency Operator Procedure 570.00 : Reactor Scram or Safety Injection version 10”, Taiwan Power Company, July, 2011.(Chinese).
    [21] “Emergency Operator Procedure 570.04 : Loss of Coolant at Reactor Vessel or Secondary Side version 5”, Taiwan Power Company, August, 2011.(Chinese).
    [22] “Emergency Operator Procedure 570.06 : Cooldown and Depressurization version 5”, Taiwan Power Company, July, 2011.(Chinese).
    [23] “Living Probabilistic Risk Assessment of Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, Vol. 4, C1-21, Dec 1995.
    [24] “Best-Estimate Analysis of the Large-Break Loss of Coolant Accident for Maanshan Unit 1 and 2 Nuclear Power Plant using the ASTRUM Methodology”, Westinghouse, June 22, 2009.
    [25] Joseph, L. Leva, “A Fast Normal Random Number Generator”, ACM Transactions on Mathematical Software, Vol.18, No.4, Dec.1992, 449-453.
    [26] S. Lee and H.J. Kim, “Prediction of Loop Seal Formation and Clearing During Small Break Loss of Coolant Accident”, Journal of the Korea Nuclear Society, Vol. 24, No.3, Sep. 1992.
    [27] “Probabilistic Risk Assessment Maanshan Nuclear Power Station Unit 1 – Final Report”, Atomic Energy Council, Executive Yuan, Republic of China, 1987.
    [28] “Living Probabilistic Risk Assessment Maanshan Nuclear Power Plant “, Institute of Nuclear Energy Research, 1995, (Chinese).
    [29] G.W. Parry, A.J. Spurgin, P. Moieni, A. Beare, “An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment “ , EPRI, June 1992.
    [30] “BEMUSE Phase V Report – Uncertainty and Sensitivity Analysis of a LB-LOCA in ZION Nuclear Power Plant ”, NEA/CSNI/R, 2009, 13.
    [31] S.W. Lee, B.D. Chung, Y.S. Bang and S.W. Bae, ”Analysis of Uncertainty Quantification Method by Comparing Monte-Carlo Method and Wilks’ formula”, Nuclear Engineering and Technology, Vol.46, No.4, 2014.
    [32] T.J. Liu, ”Reflux Condensation behavior in a U-tube steam generator with or without noncondensables”, Nucluear Engineering and Design, 204, 211–232.
    [33] ”Emergency Operator Procedure 570.42 : Short of Core Cooling version 5”, Taiwan Power Company, Sep. 2014, (Chinese).

    [34] ”Emergency Operator Procedure 570.43 : Core Cooling Ability Reduce version 5”, Taiwan Power Company, Sep. 2014, (Chinese).
    [35] ”Maanshan Nuclear Power Station 1&2 – Final Safety Analysis Report”, Ch.6, 1998.
    [36] ”Regulatory Guide 1.174”, U.S. NRC, 2011.
    [37] “10 CFR 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors”, U.S. NRC, 2017.
    [38] “10 CFR 50.46c Requirements for emergency core cooling systems for light-water nuclear power reactors”, U.S. NRC, 2018.
    [39] R. Szilard, “Applying RISMC Methods, Tools, and Data to Enhance Safety and Economics through Industry Application Demonstrations”, Idaho National Laboratory, Fourth International Conference of Nuclear Power Plant Life Management, 2017.
    [40] “10 CFR 50.69 Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors”, U.S. NRC, 2017.

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