本研究的目的為利用 MAAP 5.04 程式，驗證核能研究所發展之「核三廠嚴重事故處理指引」，研究重點為指引中的計算輔助圖 CA-1 的適切性。該圖顯示為注水流量以及釋壓閥釋出流量與反應器壓力槽壓力的關係。
計算輔助圖 CA-1 提供給運轉員在救援措施方面的參考，依電廠狀況判斷注水流量是否足以冷卻且覆蓋爐心，本研究模擬之救援動作為參考緊急運轉程序書內所假設的徵兆條件，當滿足此條件，則應當執行〝洩壓及補水 (Bleed and Feed, B&F)〞，確保當喪失所有蒸汽產生器 (SG) 之二次側熱沈時，可以透過反應器冷卻水系統之洩壓及補水，移除爐心熱能。
本研究模擬之救援案例為使用離心式充水泵 (CCP) 及正排量充水泵 (PDP) 配合開啟動力釋壓閥 (PORV) 進行 10 種應變措施案例分析，以開啟一台 CCP 和一個 PORV 為例，進行 5 種不同救援時刻之靈敏度分析。
針對一次側與二次側的水位、壓力變化作為主要探討項目，觀察模擬結果並以其他參數解釋其物理變化，而氫氣是否產生、放射性物質於反應器壓力槽內釋出為主要探討的化學變化。
為使運轉員在救援措施的執行上有更多的參考，在第四章 4.8 節探討在不同的救援措施以及不同的救援時刻，能夠有效防止各種物理以及化學現象發生 (爐心裸露、氫氣產生、熔渣掉落至反應器壓力槽底部區間、壓力槽失效) 之最遲緊急降壓注水時間點。
最後於第五章中討論 10 個應變措施案例模擬結果之差異，為 RCS 注水再次覆蓋爐心分析結果討論。靈敏度分析的 5 個不同救援時刻案例，為爐心水位案例分析結果討論。然後將 15 種模擬結果之注水能力 (capacity) 以及 PORV 釋出流量繪製於計算輔助圖 CA-1 上，以便於分析與比較。

The purpose of this study is to use the MAAP 5.04 program to verify the " Severe Accident Management Guidelines (SAMGs) of The Maanshan Nuclear Power Plant " developed by the Institute of Nuclear Energy Research (INER). The focus of the study is on the adequacy of the Computational Aid Figure CA-1 of the guidelines. The CA-1 shows the relationship between the injection water flow rate and the discharge flow rate of the power operated relief valves (PORV) and the pressure of the reactor pressure vessel.
Computational Aid CA-1 provides operators a reference in the execution of mitigation action. In a loss of offsite power event, according to the set points specified in the Emergency Operation Procedure (EOP), the mitigation action of "Bleed and Feed (B&F)" is executed by operators. In the B&F operation, the power operated relief valves (PORVs)are manually open and high pressure injection is initiated, The Computation Aid CA-1 helps operators to determine whether the water injection flow rate is sufficient to cool and cover the core under certain the status of reactor coolant system.
The cases analyzed in this study include injection using centrifugal charging pumps (CCP) and positive displacement pump (PDP) with the opening of different number of PORVs. The latest time to execute B&F to prevent core uncover, hydrogen generation, core debris relocation, and vessel failure is identified.

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[6] Sharen Saunders, “PWROG-17005-PRevision 0SECURE User Manual PA-PSC-1397”, Architectural Engineering & Plant Design (AEPD), June, 2017

[7] 王德全、陳承賢、王文興、黃勝帝、許志鵬、羅義竣、余政倫，「核三廠嚴重事故處理指引PWROG SAMG Rev.0計算輔助圖 (CA) 的建立」，核研所，中華民國110年7月

[8] C. Queral, L. Mena-Rosell, G. Jimenez, M. Sanchez-Pereab, J. Gomez-Maganc, J. Hortalba, “Verification of SAMGs in SBO sequences with Seal LOCA. Multipledamage domains”, Universidad Politécnica de Madrid, Spainb Consejo de Seguridad Nuclear, Justo Dorado 11, Spainc NFQ Solutions, Claudio Coello 78, Madrid, Spain, Received 21 October 2014 Received in revised form 12 July 2016 Accepted 13 July 2016 Available online 4 August 2016

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