本論文研究參考阿斯科-班德略斯核能協會與加泰羅尼亞理工大學能源技術研究所針對位於西班牙的阿斯科核能發電廠之洩充操作流程，結合核能研究所之核能三廠功率運轉活態安全度評估相關研究內容，並根據電廠之緊急運轉程序書模擬電廠遭遇暫態事故時之電廠狀態。分析事故為喪失廠外電源事故。本研究利用RELAP5-3D/K程式模擬電廠在肇始事件為喪失廠外電源事故並喪失二次側熱移除能力的情形下，針對運轉員洩沖操作及高壓注水再循環兩緩解行為之人為操作部分，將程式輸入參數的不準度納入考量，評估事故發生後之最高燃料棒護套溫度，量化其人為失誤機率。

本研究結合安全度評估方法，設定人為動作執行時間的機率分佈，結合對RELAP5-3D/K的結果有重大影響的參數之機率分佈，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失誤事件發生的機率，並重新計算事故序列之爐心熔損頻率。估算結果顯示，模擬的124組案例均沒有失敗，換言之主導該兩項人為操作之失誤主因並非時間上的不足，而是牽涉到運轉員的訓練程度，重新量化結果顯示該事故序列的爐心熔損頻率降低了91.3%。

In the present study the Human Error Probability (HEP) of operator mitigation actions in a Loss of Offsite Power (LOOP) sequence of a Pressurized Water Reactor (PWR) are quantified based on the results of the simulation of system thermal hydraulic code, RELAP5-3D/K. The surrogate plant in the analysis is the Maanshan Nuclear Power Station of Taiwan Power Company. The plant employs a Westinghouse designed 3-Loop PWR with rated power of 2830 MWt. The mitigation actions of a LOOP sequence as identified in the plant specific Probabilistic Safety Assessment (PSA) are analyzed. These actions are the bleed and feed (B&F) of the primary loop when any of the steam generators is almost out of liquid water and the lineup the high-head safety recirculation (HHSR) when reactor water storage tank (RWST) is about to empty. In the plant specific PSA model, HEPs are quantified using Human Cognitive Reliability (HCR) model and HEP of these two actions are 2.24 × 10-2 and 2.25 × 10-1 , respectively. Uncertainty analyses of code calculations are performed. There are 19 parameters in the Phenomenon Identification and Ranking Table (PIRT) and 124 sets of input data were generated using the Monte Carlo Sampling technique. The times for operators to execute B&F and HHSR were included as one of the parameter in the uncertainty analysis. The times to execute the mitigation actions are assumed to be log-normally distributed. The parameters of the probability density function (PDF) are determined based on the numerical values of parameters in HCR model of the plant specified PSA model. The HEPs are then determined based on uncertainty analysis, e.g. the number of samples that fail to meet the success criteria in the conventional PSA model. The simulation results showed that none of the cases is failed, which means the traditional PSA overestimates the HEP of these two human actions. The Sequence CDF decreased by 91.3%.

摘要---------------------------------------------i
英文摘要-----------------------------------------ii
致謝---------------------------------------------iii
目錄---------------------------------------------iv
表目錄-------------------------------------------vi
圖目錄-------------------------------------------vii
第一章 諸論---------------------------------------1
1.1 前言------------------------------------------1
1.2 核電廠安全度評估發展歷史------------------------2
1.3 研究目的---------------------------------------5
1.4 論文架構---------------------------------------6
第二章 文獻回顧-------------------------------------7
2.1 西班牙阿斯科核能發電廠之洩充操作研究與文獻回顧-----7
第三章 RELAP5程式介紹-------------------------------18
3.1 RELAP程式發展----------------------------------18
3.2 RELAP5-3D/K模式介紹----------------------------20
3.3 核能三廠輸入檔介紹------------------------------20
第四章 喪失廠外電源事故之模擬與分析-------------------23
4.1 喪失廠外電源事故之基準事件模擬-------------------23
4.2 人為可靠度分析----------------------------------28
4.2.1 人為可靠度量化方法----------------------------28
4.2.2 人為認知可靠度--------------------------------30
第五章 爐心熔損頻率---------------------------------32
5.1 爐心熔損頻率分析之方法論------------------------32
5.1.1 事故序列選擇與探討----------------------------32
5.1.2 如何決定人為動作之分布------------------------33
5.1.3 喪失廠外電源事故之不準度分析-------------------37
5.2 人為失誤機率結果與討論--------------------------39
5.2.1 運轉員洩充操作之現象討論-----------------------39
5.2.2 不準度分析結果--------------------------------43
5.2.3 重要度分析------------------------------------44
5.2.4 洩充操作之最大允許時間-------------------------51
5.3 結果與討論--------------------------------------53
第六章 結論-----------------------------------------54
參考文獻--------------------------------------------56

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