自然循环中朝下曲表面临界热流密度试验研究
发布时间：2018-08-01 13:27
【摘要】：日本福岛核事故成为继美国三里岛、前苏联切尔诺贝利核事故以来的最严重的核灾难，反思这次教训，单独依靠事故预防在现实条件下已经无法满足更高的核电站安全需求，反应堆严重事故缓解得到更大的重视。我国引进消化吸收AP1000第三代先进压水堆技术，该技术中一项重要的非能动安全策略为熔融物堆内滞留（In-VesselRetention,IVR)策略。而IVR策略中重要方案之一为压力容器外部冷却（External Reactor Vessel Cooling, ERVC）。在严重事故工况下，向压力容器外部的反应堆堆腔注入冷却水，如果冷却水能通过自然循环将堆内熔融物衰变热热量全部带出，就可以对反应堆进行充分有效的冷却，，将堆芯熔融物滞留在压力容器内，从而避免压力容器熔穿，确保压力容器完整性。保证冷却水充分带出衰变热的一项最关键的要求是：熔池通过下封头壁向外传出的热流不能超过外部冷却的限值，该限值即压力容器下封头外壁临界热流密度（Critical Heat Flux, CHF）。 为了更加细致的了解IVR-ERVC过程中涉及的两相自然循环流动过程特别是CHF值的影响因素，同时也为CAP1400的IVR-ERVC验证试验提供一些经验和启发，本文开展自然循环条件下试验的设计和研究。试验采用约2m半径90°圆弧、150mm×150mm的流道，在不同过冷度的去离子水中，以7.5°、37.5°、67.5°、82.5°加热铜块中心倾角和5.5m、6.5m冷凝器高度条件下，开展加热面朝下的自然循环临界热流密度试验，并使用高速摄影仪进行拍摄和分析，研究自然循环条件下朝下曲表面上沸腾换热以及CHF特性。试验研究表明：朝下曲表面上的CHF随着试验段入口过冷度减小而降低，随加热面角增加而增加，并且受到流动形式和自然循环流量的影响。
[Abstract]:The Fukushima nuclear accident in Japan has become the most serious nuclear disaster since the three Mile Island in the United States and the Chernobyl nuclear accident in the former Soviet Union. Reflecting on this lesson, relying on accident prevention alone can no longer meet higher nuclear power plant safety needs under actual conditions. More attention has been paid to the serious accident mitigation of the reactor. The third generation advanced pressurized water reactor (PWR) technology of digesting and absorbing AP1000 is introduced in China. One of the most important inactive security strategies is the In-Vessel retention (IVR) strategy. One of the most important schemes in IVR strategy is the external cooling (External Reactor Vessel Cooling, ERVC). Of pressure vessel. Under severe accident conditions, cooling water is injected into the reactor chamber outside the pressure vessel. If the cooling water can take out the decay heat and heat of the molten matter in the reactor through natural circulation, the reactor can be cooled fully and effectively. The core melt is trapped in the pressure vessel to avoid melting through the pressure vessel and ensure the integrity of the pressure vessel. One of the most important requirements to ensure that cooling water fully brings out decay heat is that the heat flux from the molten pool passing through the lower head wall cannot exceed the limit of external cooling, which is the critical heat flux (Critical Heat Flux, CHF). Of the outer wall of the head under the pressure vessel. In order to understand the influence factors of the two-phase natural circulation process, especially the CHF value, involved in the IVR-ERVC process in detail, it also provides some experience and inspiration for the IVR-ERVC verification test of CAP1400. In this paper, the design and research of experiments under natural circulation conditions are carried out. In the experiment, the critical heat flux density of the natural circulation was tested in deionized water with different undercooling degrees by using a flow channel with about 2m radius of 90 掳circular arc and 150 mm 脳 150mm, under the conditions of 7.5 掳~ 37.5 掳~ (37.5 掳) ~ 67.5 掳~ 82.5 掳heating copper block central dip angle and 5.5 m ~ 6.5m condenser height. The characteristics of boiling heat transfer and CHF on the downward curved surface under natural circulation were studied by means of high speed photography. The experimental results show that the CHF on the downward curved surface decreases with the decrease of the subcooling at the inlet of the test section and increases with the increase of the heating surface angle, and is affected by the flow form and natural circulation flow rate.
【学位授予单位】：上海交通大学
【学位级别】：硕士
【学位授予年份】：2013
【分类号】：TL364.4
本文编号：2157729
[Abstract]:The Fukushima nuclear accident in Japan has become the most serious nuclear disaster since the three Mile Island in the United States and the Chernobyl nuclear accident in the former Soviet Union. Reflecting on this lesson, relying on accident prevention alone can no longer meet higher nuclear power plant safety needs under actual conditions. More attention has been paid to the serious accident mitigation of the reactor. The third generation advanced pressurized water reactor (PWR) technology of digesting and absorbing AP1000 is introduced in China. One of the most important inactive security strategies is the In-Vessel retention (IVR) strategy. One of the most important schemes in IVR strategy is the external cooling (External Reactor Vessel Cooling, ERVC). Of pressure vessel. Under severe accident conditions, cooling water is injected into the reactor chamber outside the pressure vessel. If the cooling water can take out the decay heat and heat of the molten matter in the reactor through natural circulation, the reactor can be cooled fully and effectively. The core melt is trapped in the pressure vessel to avoid melting through the pressure vessel and ensure the integrity of the pressure vessel. One of the most important requirements to ensure that cooling water fully brings out decay heat is that the heat flux from the molten pool passing through the lower head wall cannot exceed the limit of external cooling, which is the critical heat flux (Critical Heat Flux, CHF). Of the outer wall of the head under the pressure vessel. In order to understand the influence factors of the two-phase natural circulation process, especially the CHF value, involved in the IVR-ERVC process in detail, it also provides some experience and inspiration for the IVR-ERVC verification test of CAP1400. In this paper, the design and research of experiments under natural circulation conditions are carried out. In the experiment, the critical heat flux density of the natural circulation was tested in deionized water with different undercooling degrees by using a flow channel with about 2m radius of 90 掳circular arc and 150 mm 脳 150mm, under the conditions of 7.5 掳~ 37.5 掳~ (37.5 掳) ~ 67.5 掳~ 82.5 掳heating copper block central dip angle and 5.5 m ~ 6.5m condenser height. The characteristics of boiling heat transfer and CHF on the downward curved surface under natural circulation were studied by means of high speed photography. The experimental results show that the CHF on the downward curved surface decreases with the decrease of the subcooling at the inlet of the test section and increases with the increase of the heating surface angle, and is affected by the flow form and natural circulation flow rate.
【学位授予单位】：上海交通大学
【学位级别】：硕士
【学位授予年份】：2013
【分类号】：TL364.4
【参考文献】
相关期刊论文 前2条
1 文青龙;陈军;卢冬华;赵华;;倾斜下朝向加热表面汽泡行为可视化实验研究[J];核动力工程;2012年03期
2 李飞;李永春;程旭;;针对REPEC加热实验的RELAP5程序模拟与分析[J];原子能科学技术;2012年07期
本文编号：2157729
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