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  • 學位論文

進步型沸水式反應爐冷卻水流失事故熱流現象之研究

A Study on the Thermofluidic Phenomena for an Advanced Boiling Water Reactor under the Loss-of-Coolant Accident Conditions

指導教授 : 鄧治東
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摘要


本研究主要針對核四廠所採用之進步型沸水式反應爐,應用RELAP5-YA熱水流分析程式進行評估模式的建立及冷卻水流失事故的安全分析。研究重點係針對反應爐冷卻水系統不同的破口位置及破口尺寸所造成之冷卻水流失事故加以模擬,並將其爐心所達到之燃料護套最高溫度(PCT)及反應爐最低水位加以評估,以分析事故之嚴重性。 冷卻水流失事故分析考慮之破管面積範圍由反應爐底蓋洩水管(0.0219 ft2)至圍阻體外主蒸汽管(4.24 ft2),分別進行七種不同破口位置及其不同破口尺寸,總計共十九組個案之模擬分析。分析結果顯示,對燃料護套最高溫度而言,最嚴重之事故為圍阻體外主蒸汽管路破口,其燃料護套最高溫度為1289.15 ℉,此溫度遠在法規10CFR50.46所定之限值 2200 ℉以下;對反應爐最低水位而言,則最嚴重之事故為高壓注水系統管路破口,反應爐降水區最低水位降至18.4 ft,約在有效燃料頂部以下11.6 ft,然而爐心內部仍為雙相飽和流體所完全覆蓋。

並列摘要


The purposes of this research are directed at Advanced Boiling Water Reactor (ABWR) adopted by the Lung-Men Nuclear Power Station (LMNPS) using the RELAP5-YA methodology to establish an evaluative mode and to analyze the safety of the behavior of the reactor core under the postulated LOCA conditions. The main point of this research is to simulate loss-of-coolant accidents (LOCAs) which are caused by different broken lines positions and areas and also to evaluate how limiting the accident is if the core reaches its peak cladding temperature (PCT) and the minimum water level measured outside the core shroud. In this study, there are 19 cases simulated and analyzed with 7 different breaks’ positions and areas. The range of the breaks’ spectrum is considered from Bottom Head Drain Line (0.0219 ft2) to Main Steamline Outside Containment (4.24 ft2). The analytic results obtained from the PCT evaluations revealed that the most limiting accident was the case with a break at the Main Steamline Outside Containment. Its PCT was 1289.15 ℉, which is much below the upper temperature limit of 2200℉ set by the nuclear regulatory authority. The most limiting accident for the minimum water level measured outside the core shroud is the case with a break at the High Pressure Core Flooder (HPCF) Line, the minimum water level drops to 18.4 ft. For this case, even though the water level has dropped to 11.6 ft below the top of the active fuel, the reactor core was still completely covered by the two-phase water.

參考文獻


1. “RELAP5YA-A Computer Program For Light Water Reactor System Thermal-Hydraulic Analysis User’s Manual, ” Yankee Atomic Electric Company (1982).
4. “MAAP 3.0 Modular Accident Analysis Program User’s Manual, ” Fauske & Associates Inc. (1987).
17. “Preliminary Safety Analysis Report for Lungmen Nuclear Power Station Units 1 & 2, ”Taiwan Power Company, 1997.
2. “RETRAN-02-A Program for Transient Thermal-Hydraulic analysis of Complex Flow Systems, ” EPRI Document No. P1850-CCMa (1984).
3. R. O. Wooton, P. Cybulskis, S. F. Quayle “MARCH-2 ( Meltdown Accident Response Characteristics ) Code Description and User’s Manual, ” Battelle Columbus Laboratories (1984).

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