本篇研究為評估壓水式反應器壓力槽外部冷卻裝置（External Reactor Vessel Cooling，ERVC）將熔融物滯留壓力槽內（In-Vessel Retention，IVR）之能力，本研究以系統熱水流分析程式RELAP5-3D為主要分析工具，嚴重事故分析程式MELCOR為輔助分析工具。建立RELAP5-3D輸入檔，模擬IVR-ERVC之自然對流與移熱能力，利用模擬結果之熱水流參數，帶入臨界熱通率(Critical Heat Flux，CHF)經驗公式，計算分析不同位置之臨界熱通率，了解壓水式反應器 IVR 相關設計的安全餘裕，並比較MELCOR與RELAP5-3D之模擬結果。
本論文分為兩部分，第一階段為以RELAP5-3D程式模擬AP-1000 IVR裝置，以兩種熔融爐渣池層次結構以固定功率帶入RELAP5-3D輸入檔計算，將RELAP5-3D模擬結果代入ULPU、SULTAN、SBLB和KAIST臨界熱通率關係式的相關參數中，以確定AP-1000 IVR設計的安全裕度。結果顯示臨界熱通率隨角度增加。但反應器壓力槽底部半球形90度角處的臨界熱通率與臨界熱通率比 (Critical Heat Flux Ratio，CHFR)最低。結果顯示熔融池層次結構分佈對IVR安全性能的影響非常顯著。
研究第二部分使用MELCOR模擬高功率壓水式反應器CAP1700嚴重事故的演變。將MELCOR計算得到之熔融爐渣池對壓力槽壁的瞬態熱負載輸入RELAPR-3D模型，模擬IVR流道內的自然對流，再計算所得之IVR流道熱水流參數代入ULPU，SULTAN，SBLB和MELCOR臨界熱通率關係式中，並評估IVR的有效性。 MELCOR的模擬結果表示，壓力槽槽壁熱負載取決於熔融爐渣池層次結構型態。在483分鐘時，爐心底部外壁的熱通率在傾角約為68°時達到最高值。隨著事故的發展，熱通率峰值從小傾角移動到更大的角度。 MELCOR和RELAP5-3D的模擬結果均顯示，當IVR流道中的水達到飽和狀態時，流量會出現大幅度波動。肇因為兩相流引起的不穩定性。如果IVR的入口水溫度可以保持足夠低以避免IVR通道沸騰，RELAP5-3D所預測的流道流量將接近大致穩定狀態。MELCOR 與RELAP5-3D IVR 流道的熱水流模擬結果顯示，MELCOR 程式的熱水流模擬部分尚有改進的空間。
各臨界熱通率關係式所預測臨界熱通率與時間的關係差異均非常不明顯，差異最大的為SULTAN預測值，差異小於1.7%。分析結果顯示，熱通率最大值 (483分鐘) 時，不同經驗公式所預測之臨界熱通率比（CHFR）的最小值在55°和75°之間。除了MELCOR關係式之外，所預測的臨界熱通率比(CHFR)皆大於1.2。CAP1700反應器發生冷卻水流失事故引發之嚴重事故中，IVR 可以發揮功效終止嚴重事故的進一步惡化。

In the advanced design of Pressurized Water Reactor (PWR), In-Vessel Retention (IVR) device is incorporated to enhance the capability of removing heat through the outer wall of the reactor vessel during a core meltdown. External reactor vessel cooling (ERVC) is established by flooding the reactor cavity during a severe accident. Previous studies have concluded that the amount of heat removed from the debris pool in the vessel lower plenum is limited by the critical heat flux (CHF) at the outer surface of the vessel wall. In the present study, the effectiveness of ERVC/IVR in terminating the progression of a core melt accident is investigated. The codes used in the assessment are RELAP5-3D and MELCOR. The RELAP5-3D thermal-hydraulic code is used to simulate the natural convection flow within the water channel of the IVR device. MELCOR is used in the integrated simulation of a core melt sequence to determine the heat load on the vessel wall of lower plenum. The results of RELAP5-3D simulation are substituted into selected Critical Heat Flux (CHF) correlation to determine the heat removal capability of IVR.
The study is separated into two parts. In the first part of study, a three-dimensional model of RELAP5-3D is constructed to micmic the IVR design of AP1000. The heat load to the lower head is based on two bounding melt configurations. Configuration I involves a stratified light metallic layer on top of a molten ceramic pool, and Configuration II represents the condition that an additional heavy metal layer forms below the ceramic pool. The results of RELAP5-3D IVR simulation demonstrated that a natural convection flow can be established smoothly after the heat load is imposed and the heat transfer mode at the vessel outer surface is nucleate boiling. The results of RELAP5-3D simulations are substituted into ULPU, SULTAN, SBLB, and KAIST critical heat flux correlations to determine the safety margin of the AP-1000 IVR design. Among these four correlations, the CHFs predicted by the ULPU and SBLB correlations are insensitive to the thermal-hydraulic conditions of the coolant channel. The CHF predicted by SULTAN is the lowest among the four correlations. The results demonstrate that the predicted CHF increases with the angle. Nevertheless, the ratio of CHF to heat flux is lowest around the upper quarter of the vessel. The impact of corium pool configuration on the safety performance of IVR is very significant. Under Configuration II, as predicted by the SULTAN and KAIST correlations, the performance of the current AP-1000 IVR design is marginal.
In the second part of the study, the transient heat load to the vessel wall of vessel lower plenum is obtained from an integrated analyses for a selected core melt sequence of CAP1700 PWR with thermal power around 5000 MW. The sequence selected is a large Loss of Coolant Accident (LOCA) in conjunction with station blackout. The transient load to the lower plenum vessel wall is transported to RELAP5-3D simulation of IVR. The MELCOR simulation demonstrated that the heat load at the vessel wall of the lower plenum depends on the configuration of the debris pool in the lower plenum. The heat flux to the vessel wall reached a maximum at 483 min, at an inclination angle of approximately 68°. The peak heat flux moved from a small inclination angle to a larger angle as the accident progressed. Both MELCOR and RELAP5-3D calculations predicted a gradual buildup of natural convection flow within the IVR channel following the application of a heat load to the vessel wall. Both codes predicted that the flow would experience large amplitude fluctuations as the water in the IVR flow channel reached saturation. These fluctuations were attributed to instability induced by two-phase flow. The flow as predicted by these two codes is significantly different.
If the inlet temperature can be kept sufficiently low to obviate boiling in the IVR channel, RELAP5-3D predicts that the channel flow will approach an approximately steady state. The selected CHF correlations predicted significantly different CHFs. Nevertheless, the predicted CHFs are relatively time independent. The minimum ratio of CHF to heat flux occurs at 483 minutes after accident initiation. The minimum margin was found between 55° and 75° in all correlations. Except the MELCOR correlation, the CHFR predicted by other three correlations are greater than 1.2. MELCOR correlation was developed based on the experimental data of pool boiling. It can be concluded that the IVR design of CAP1700 can effectively retain the core melt with the vessel lower plenum.

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