餘熱移除(RHR)為反應爐緊急爐心冷卻系統(ECCS)之一支系統，當電廠發生喪失冷卻水事故(Loss of Coolant Accident, LOCA)時，高壓注水系統無法維持反應爐水位，將啟動低壓注水系統-餘熱移除系統(RHR)，以供給反應爐冷卻水維持水位，目的將爐心產生的衰變餘熱移除，避免爐心溫度過高發生危害。
依據案例核電廠之終期安全分析報告可知，影響爐心熔毀風險貢獻最大之事故序列中，消防補水系統 (ACIWA) 失效為導致爐心熔毀的最終主因，而在ACIWA系統中，又以RHR之管線破裂為各元件中之風險增加當量值最高者。
因此本文以案例核電廠ACIWA系統中之RHR管線為研究對象，先以有限元素軟體ABAQUS分析RHR管線中接觸較為複雜，且可能造成管內液體洩漏的法蘭接頭，並能與法蘭接頭四點彎矩載重試驗結果相互擬合，驗證法蘭接頭模型之正確性。接著將此接頭用於RHR管線系統模型內，以多組地震歷時輸入至模型，並以管線系統振動台試驗驗證管線系統模型之正確性。本論文最終目的在建立RHR系統以地表運動為參數之耐震易損性曲線，以Reed and Kennedy 所提出之易損性分析方法，討論易損性分析方法中所需考慮之變數以及如何選取，並配合數值模型分析結果，最後將易損性曲線繪出，作為往後地震機率式風險評估之參考依據。
分析結果顯示此案例電廠RHR管線系統耐震性能良好，因地震產生損壞繼而導致核電廠發生危害之機率非常低，並非影響核電廠安危之關鍵設備。

The residual heat removal (RHR) is one of the systems of the emergency core cooling system (ECCS). When the loss of coolant accident (LOCA) happens, the high pressure water injection system cannot maintain the water level in reactor and will activate low pressure water injection system - residual heat removal system (RHR) in order to supply the cooling water for reactor to maintain the water level. The purpose is to remove the residual heat generated by the reactor core and avoid the hazard induced due to high core temperature.
According to the final safety analysis report of the sample nuclear power plant (NPP), in the accident sequence that has the greatest contribution to inducing reactor core meltdown, the ac-independent water addition system (ACIWA) is the main cause. In addition, in the ACIWA system, risk achievement worth (RAW) of rupture of the RHR piping is the highest among all components in NPP.
Therefore, this dissertation takes the RHR piping in the ACIWA system of the sample NPP as the research object. Firstly, the finite element software ABAQUS is used to analyze the flange joints in the RHR piping, which are more complicated in analysis procedure and may cause liquid leakage in the piping. Then the results of 4-point bending test are used to verify the correctness of the flange joint model. This joint is further used in the RHR piping system shaking table model, and the correctness of the model is verified by shaking table test.
The ultimate goal of this dissertation is to establish the fragility curve of the RHR system and discuss the variables supposed to be considered and how to select them in the fragility analysis method proposed by Reed and Kennedy. Combined with the numerical model analysis results, the fragility curve is finally drawn, which serves as the reference for seismic probabilistic risk assessment (SPRA)..
The analysis results show that the RHR piping system in the sample NPP has good seismic performance, and the probability of damage caused by earthquakes is very low, which is not a key equipment affecting the safety of sample NPP.

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