AP1000先進型核能電廠圍阻體熱水流安全分析方法論

核能電廠安全分析之設計基準事故中，有許多熱水流現象須予以考慮。如當失水事故(Loss Of Coolant Accident; LOCA)發生時，圍阻體內部溫度壓力劇烈上升，其峰值可能會破壞核能電廠圍阻體之完整性，並且導致圍阻體阻擋輻射物質外洩之屏蔽能力失效。當反應器爐心裸露後，鋯水金屬反應產生大量氫氣會散逸進入圍阻體，由於氫氣濃度之累積有產生氫氣劇燃，甚至氫爆之威脅。本研究論文主要針對圍阻體內常見之熱流現象與氫氣擴散等重要行為進行分析；濃度累積時變值，及會否發生氫氣劇燃或氫爆現象進行分析研究與討論；另本研究分析研究之對象中採用AP1000先進型輕水壓水式核能電廠與國內之台電核能二、三廠為分析研究之標的。本研究主要分為兩大方向進行1)、AP1000被動安全系統之分析能力建成，與2)、圍阻體內部因應福島事件後之氫氣擴散模式分析與應用能力建立。第一點關注于AP1000先進核能電廠之圍阻體安全性分析工作上。AP1000主要在設計上採用了被動安全移熱機制，來增強核電廠之安全特性；被動安全機制主要採用浮力與自然對流來進行移除爐心之衰變熱能，以保障其為自發機制且能長期有效並安全移除燃料衰變熱，保證燃料及其護套不會過熱破損，有效防制其內產生之放射性同位素不致外釋。本研究主要針對AP1000圍阻體在預想事故發生後短程時段與長程時段中之溫壓反應事故安全分析。短程分析中，本研究採用美國核能管制委員會(Nuclear Regulation Commission, NRC)所公開AP1000之設計控制文件(Design Control Documents ,DCD)，並且採用DCD中計算所得之LOCA短程沖放時變數據作為邊界條件，來模擬LOCA後短程時間圍阻體內之事故演變情況；於LOCA長程分析計算模式下，考慮到爐心衰變熱之影響，本研究模擬了反應器壓力容器(Reactor Pressure Vessel, RPV)與其飼水破口(Feedwater Line Break)之沖放，並且加入了被動安全系統之考量與作動，進行長程事故時段分析之模擬。第二項重點則著重於氫氣擴散之行為分析驗證與其電廠應用之範疇。本研究亦參考國內台電公司核三廠相關資料進行了SBO事故情況下圍阻體熱水流安全分析。更藉由不同圍阻體建築結構數據建立模型，進一步分析圍阻體熱水流時變參數狀態與氫氣擴散遷移機制。最終經長期研究經驗，歸納出一組簡單方法論，以引導後學者於建立相同類似之圍阻體結構輸入模式時，有一有效之參考流程步驟依循，以增進國內核工界於圍阻體安全分析時，圍阻體輸入模式之自建能力。本研究採用GOTHIC 8.0程式作為圍阻體安全分析模擬軟體，計算結果將與DCD所提供之圍阻體溫壓反應數據進行比對驗證；並且加入Chen之水膜熱傳關係式來計算一次圍阻體外壁次冷態水膜對流蒸發熱傳之現象驗證。

關鍵字

圍阻體；電廠分析； AP1000 ；氫氣擴散； GOTHIC

並列摘要

In the design basis accidents (DBA) of nuclear safety analysis, there are many thermal hydraulic phenomena must concern. When a loss-of-coolant accident (LOCA) occurs, the pressure inside the containment rises rapidly. The peak value of the pressure could undermine the containment integrity, causing containment shielding leakage and release of the radioactive material into the atmosphere. During a LOCA, the reactor core cladding zirconium reacts with water to produce a large amount of hydrogen gas, which is released into the containment. The accumulation of hydrogen gas concentrations may cause hydrogen combustion, even lead to the threat of a hydrogen explosion. This research focused on the common thermal hydraulic phenomena inside the reactor containment and analyzed the hydrogen diffusion behavior in the containment; and analyze whether hydrogen explosion or hydrogen combustion phenomenon. Another analyzed objects of this study the use of AP1000 advanced type light water pressure reactor and nuclear Tai-power Kuosheng, Maanshan nuclear power plants for the analysis of the subject. There are two issues in this study: 1) The GOTHIC code simulate ability of AP1000 passive system and, 2) The hydrogen diffusion model in containment. The first issue was focused on the containment safety of AP1000 advanced nuclear power plant. This study investigates the integrity of the protective mechanism of the AP1000 containment system during a LOCA was investigated. The performance of the passive containment cooling system (PCCS) and its ability to perform decay heat removal for long-term cooling were evaluated. The PCCS utilizes gravity-driven natural convection and naturally induced passive safety devices to release the decay heat of fuels into the atmosphere. In this study, two accidents were analyzed: a double-ended guillotine break in a hot leg and a double-ended break in a main steam line. The analytical results were compared with the corresponding results provided in the AP1000 ‘‘Design Control Document’’ (DCD). Short-term calculations and comparisons with the DCD suggested that GOTHIC 8.0 with a natural convection model may appropriately represent the phenomenon of passive, safe heat removal. In addition, the long-term calculations of spraying the outer primary containment with a water film and the steam condensation of the inner containment were also simulated. The second issue of the study also referred to containment thermal hydraulic safety analysis data for Taipower Maanshan nuclear power plant under a SBO accident. Based on the different containment structure data because of differences in the containment models, we can further analyze the state of the hydraulic behavior and hydrogen diffusion transfer mechanism. Based on the results, we can summarize a set of simple methodologies for the GOTHIC code, which can enhance the nuclear containment safety analysis ability in Taiwan. GOTHIC 8.0 code was used in the analysis and simulation, and the results, such as containment temperature pressure response, were verified through comparison with the corresponding results in DCD. The code also included Chen’s correlation, which gives the relationship of water film heat transfer. Chen’s model can calculate the heat transfer by liquid film evaporation convection at the secondary containment air flow path area.