本研究根據核能研究所(Institute of Nuclear Energy Research, INER)之核能三廠(Maanshan Nuclear Power Plant, MNPP)功率運轉活態安全度評估研究內容，並依照電廠之緊急運轉程序書(Emergency Operating Procedures, EOP)模擬電廠在發生蒸汽產生器管束破裂事故(Steam Generator Tube Rupture, SGTR)時的電廠狀況。本研究利用RELAP5-3D/K程式模擬蒸汽產生器管束破裂事故下，針對運轉員執行高壓注水(High-Head Safety Injection, HHSI)、緊急降溫降壓(Emergency Cooldown and Depressurization, Emergency CND)、和燃料更換水儲存槽(Refueling Water Storage Tank, RWST)再補水(replenishment)等人為緩解措施的時間，並將程式輸入參數的不準確度(Uncertainty)納入考量，計算事故發生後之最高燃料棒護套溫度(Peak Cladding Temperature, PCT)，並依此量化前述三項措施的人為失誤機率(Human Error Probability, HEP)。

本研究利用人為認知可靠度(Human Cognitive Reliability, HCR)模式，設定人為緩解措施執行時間的機率分佈，結合對RELAP5-3D/K結果有決定性影響的參數之機率分佈，以蒙地卡羅取樣(Monte Carlo Sampling)，並進行多次程式運算。運算的結果依照事件樹(Event Tree)頂端事件的成功準則為基準，判斷人為緩解動作發生失誤的機率。並以此結果重新計算事故序列之爐心熔損頻率(Core Damage Frequency, CDF)。根據模擬結果，總計124組案例均沒有失敗導致爐心溫度上升至1,200°F以上，顯示運轉人員應該有足夠的時間判斷與執行緩和措施，人為失誤機率僅剩執行時的誤失。

This study is based on the research conducted by the Institute of Nuclear Energy Research (INER) on the safety assessment of power operation activities at the Maanshan Nuclear Power Plant (MNPP). It simulates the plant's condition in the event of a Steam Generator Tube Rupture (SGTR) accident, following the Emergency Operating Procedures (EOP) of the power plant. The RELAP5-3D/K program is used to simulate the SGTR accident and evaluate the plant's response, including the actions taken by operators such as High-Head Safety Injection (HHSI), emergency cooldown and depressurization (emergency CND), and replenishment from the refueling water storage tank (RWST). The study also considers the uncertainties in input parameters and calculates the Peak Cladding Temperature (PCT) of fuel rods after the accident, quantifying the Human Error Probability (HEP) associated with the aforementioned mitigation measures.

The study employs the Human Cognitive Reliability (HCR) model to establish the probability distribution of the execution time for human mitigation actions. Monte Carlo sampling is used to combine this with the probability distribution of parameters that have a decisive impact on RELAP5-3D/K results. Multiple program calculations are performed. Based on the success criteria of the top event in the Event Tree, the probability of human error in the execution of mitigation actions is determined. This result is then used to recalculate the Core Damage Frequency (CDF) of accident sequences. According to the simulation results, in all 124 cases, there were no failures leading to a core temperature rise above 1,200°F, indicating that operators have sufficient time to assess and implement mitigation measures, with the remaining probability of human error being associated with missed opportunities during execution.

摘要 i
英文摘要 ii
目錄 iii
表目錄 v
圖目錄 vi
第一章 緒論 1
1.1前言 1
1.2研究目的 1
1.3論文架構 2
第二章 文獻回顧 3
2.1 核電廠安全度評估(PSA)發展歷史 3
2.1.1人為可靠度評估方法 6
2.1.2人為認知可靠度 7
2.2蒸汽產生器管束破裂事故介紹 9
2.3韓國韓蔚4號機組SGTR事故模擬與緩解措施分析 10
2.3.1 MARS程式輸入檔驗證 11
2.3.2 事故模擬 12
2.3.2.1低功率、停機條件下之模擬 13
2.3.2.2全功率條件下之模擬 17
第三章 RELAP5程式介紹 22
3.1 RELAP程式發展 22
3.2 RELAP5-3D/K模式介紹 22
3.3 RELAP5-3D/K核三廠輸入檔介紹 23
第四章 蒸汽產生器管束破裂事故序列模擬與分析 27
4.1蒸汽產生器管束破裂事故描述 27
4.2肇始事件事件樹 28
4.3蒸汽產生器管束破裂事故基準事件模擬 30
第五章 爐心熔損頻率再量化 35
5.1人為可靠度評估方法論 35
5.1.1事故序列選擇與探討 35
5.1.2如何決定人為動作之分布 35
5.1.3蒸汽產生器管束破裂事故之不準度分析 39
5.2爐心熔損分析結果與討論 42
5.2.1蒸汽產生器管束破裂事故現象討論 42
5.2.2不準度分析結果 46
5.2.3 其它結果討論 52
第六章 與其他事故比較 55
6.1其他事故事件樹介紹 55
6.2重新量化結果比較 57
第七章 結論與未來建議 58
7.1 結論 58
7.2未來建議 59
參考文獻 60

[1] “Living Probabilistic Risk Assessment Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, 1995, (Chinese).
[2] Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR]. United States: N. p., 1975. Web. doi:10.2172/7134131.
[3] S.K. Chen, “The History Perspective of PRA”, PRA II Course Lecture, NTHU NES, 2017.
[4] C.C. Chao, “The History and Methodology of PRA”, PRA I Course Lecture, NTHU NES, 2017.
[5] “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory”, NRC/60FR 42622, 1995.
[6] “Emergency Operator Procedure 570.00 : Reactor Scram or Safety Injection version 13”, Taiwan Power Company, February, 2016, (Chinese).
[7] “Emergency Operator Procedure 570.10 : Steam Generator Rupture version 7”, Taiwan Power Company, February, 2016, (Chinese).
[8] Jungjin Bang, Gi Hyeon Choi, Dong-Wook Jerng, Sung-Won Bae, Sunghyon Jang, Sang Jun Ha, “Analysis of Steam Generator Tube Rupture Accidents for the Development of Mitigation Strategies”, Nuclear Engineering and Technology 54 (2022) 152-161.
[9] B.D. Chung, K.D. Kim, S.W. Bae, J.J. Jeong, S.W. Lee, M.K. Hwang, C. Yoon, MARS Code Manual Volume I: Code Structure, System Models, and Solution Methods, Korea Atomic Energy Research Institute, 2010.
[10] Jungjin Bang, Gi Hyeon Choi, Dong-Wook Jerng, Sung-Won Bae, Sunghyon Jang, Sang Jun Ha, Analysis of steam generator tube rupture accidents for the development of mitigation strategies, Nuclear Engineering and Technology, Volume 54, Issue 1, 2022
[11] Y.-S. Son, J.-Y. Shin, H.-G. Lim, J.-H. Park, S.-C. Jang, Thermal-hydraulic calculations using MARS code applied to low power and shutdown probabilistic
safety assessment in a PWR, Nucl. Eng. Des. 235 (15) (2005) 1571-1581. [12] Khnp, Hanuel 3,4 Final Safety Analysis Report, 1998.
[13] A.P.R. Khnp, 1400 Standard Safety Analysis Report Chapter 5, 2002.
[14] “Appendix K to Part 50 – ECCS Evaluation Models”, U.S. NRC, March, 2013.
[15] “Maanshan Nuclear Power Station 1 & 2 – Final Safety Analysis Report”, Ch. 6, 1998.
[16] “Probabilistic Risk Assessment Maanshan Nuclear Power Station Unit 1 – Final Report”, Atomic Energy Council, Executive Yuan, Republic of China, 1987.
[17] “Living Probabilistic Risk Assessment Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, Vol. 1, Ch. 3, 77, Dec 1995.
[18] “Living Probabilistic Risk Assessment Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, Vol. 4, Appendix C, Ch. 2, 1-4, Dec 1995.
[19] “Living Probabilistic Risk Assessment Maanshan Nuclear Power Plant”, Institute of Nuclear Energy Research, Vol. 4, Appendix C, Ch. 2, 4-7, Dec 1995.
[20]Dhillon BS (2007) Human reliability and error in transportation systems. Springer, New York, pp 43–54
[21]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.
[22]“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.
[23]“BEMUSE Phase V Report – Uncertainty and Sensitivity Analysis of a LB-LOCA in ZION Nuclear Power Plant”, NEA/CSNI/R, 2009, 13.
[24]Joseph, L. Leva, “A Fast Normal Random Number Generator”, ACM Transactions on Mathematical Software, Vol. 18, No. 4, Dec 1992, 449-453.
[25]S.W. Lee, B.D. Chung, Y.S. Bang, S.W. Bae, “Analysis of Uncertainty Quantification Method by Comparing Monte-Carlo Method and Wilk’s Formula”, Nuclear Engineering and Technology, Vol. 46, No. 4, 2014.
[26] Chen, Yu-Min, “Effect of Power Uprate on Safety Margin and Core Damage Frequency in a MBLOCA”, 2018
[27] Wang, Ta-Chun, “Analysis of SBLOCA in a Pressurized Water Reactor: Quantification of Human Error Probability and Thermal Hydraulic Analysis of Cooldown and Depressurization”, 2022
[28] B.D Chung, K.D. Kim, S.W. Bae, J.J. Jeong, S.W. Lee, M.K. Hwang, C. Yoon, MARS Code Manual Volume I: Code Structure, System Models, and Solution Methods, Korea Atomic Energy Research Institute, 2010.
[29] A.D Swain & H.E. Guttman, “Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Application,” NUREG/CR-1278, 1984.
[30] 1975. "Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR]". United States.