【摘要】：日本福島核事故成為繼美國三里島、前蘇聯(lián)切爾諾貝利核事故以來的最嚴(yán)重的核災(zāi)難，反思這次教訓(xùn)，單獨(dú)依靠事故預(yù)防在現(xiàn)實(shí)條件下已經(jīng)無法滿足更高的核電站安全需求，反應(yīng)堆嚴(yán)重事故緩解得到更大的重視。我國引進(jìn)消化吸收AP1000第三代先進(jìn)壓水堆技術(shù)，該技術(shù)中一項(xiàng)重要的非能動安全策略為熔融物堆內(nèi)滯留（In-VesselRetention,IVR)策略。而IVR策略中重要方案之一為壓力容器外部冷卻（External Reactor Vessel Cooling, ERVC）。在嚴(yán)重事故工況下，向壓力容器外部的反應(yīng)堆堆腔注入冷卻水，如果冷卻水能通過自然循環(huán)將堆內(nèi)熔融物衰變熱熱量全部帶出，就可以對反應(yīng)堆進(jìn)行充分有效的冷卻，，將堆芯熔融物滯留在壓力容器內(nèi)，從而避免壓力容器熔穿，確保壓力容器完整性。保證冷卻水充分帶出衰變熱的一項(xiàng)最關(guān)鍵的要求是：熔池通過下封頭壁向外傳出的熱流不能超過外部冷卻的限值，該限值即壓力容器下封頭外壁臨界熱流密度（Critical Heat Flux, CHF）。 為了更加細(xì)致的了解IVR-ERVC過程中涉及的兩相自然循環(huán)流動過程特別是CHF值的影響因素，同時也為CAP1400的IVR-ERVC驗(yàn)證試驗(yàn)提供一些經(jīng)驗(yàn)和啟發(fā)，本文開展自然循環(huán)條件下試驗(yàn)的設(shè)計(jì)和研究。試驗(yàn)采用約2m半徑90°圓弧、150mm×150mm的流道，在不同過冷度的去離子水中，以7.5°、37.5°、67.5°、82.5°加熱銅塊中心傾角和5.5m、6.5m冷凝器高度條件下，開展加熱面朝下的自然循環(huán)臨界熱流密度試驗(yàn)，并使用高速攝影儀進(jìn)行拍攝和分析，研究自然循環(huán)條件下朝下曲表面上沸騰換熱以及CHF特性。試驗(yàn)研究表明：朝下曲表面上的CHF隨著試驗(yàn)段入口過冷度減小而降低，隨加熱面角增加而增加，并且受到流動形式和自然循環(huán)流量的影響。
[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 擄circular arc and 150 mm 脳 150mm, under the conditions of 7.5 擄~ 37.5 擄~ (37.5 擄) ~ 67.5 擄~ 82.5 擄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.
【學(xué)位授予單位】：上海交通大學(xué)
【學(xué)位級別】：碩士
【學(xué)位授予年份】：2013
【分類號】：TL364.4

【參考文獻(xiàn)】

相關(guān)期刊論文 前2條

1 文青龍;陳軍;盧冬華;趙華;;傾斜下朝向加熱表面汽泡行為可視化實(shí)驗(yàn)研究[J];核動力工程;2012年03期

2 李飛;李永春;程旭;;針對REPEC加熱實(shí)驗(yàn)的RELAP5程序模擬與分析[J];原子能科學(xué)技術(shù);2012年07期

本文編號：2157729