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Study on Accident Mitigation and Cooling of Damaged Core during Severe Accidents of Right water Reactor (Possibility of Cooling by Penetrated Water into a Narrow Gap between a Reactor Pressure Vessel and Piled-up Debris of Damaged Core)

右水反应堆严重事故期间受损堆芯的事故缓解和冷却研究（反应堆压力容器与堆积的受损堆芯碎片之间的狭窄间隙渗入水进行冷却的可能性）

基本信息

批准号：
08650272
负责人：
KOIZUMI Yasuo
金额：
$ 1.22万
依托单位：
Kogakuin University
依托单位国家：
日本
项目类别：
Grant-in-Aid for Scientific Research (C)
财政年份：
1996
资助国家：
日本
起止时间：
1996 至 1997
项目状态：
已结题

项目摘要

The facility used was a R-113 closed-loop. The test section was a thick wall copper pipe of 240mm in length. 100mm in OD and 34 or 36mm in ID.A glass pipe was inserted into the test pipe to from a vertical-narrow-annular flow passage of 0.5,1,2 or 5mm gap. Following the Bond number scaling, the gap size was determined from the gap width between the reactor pressure vessel and the piled-up debris of the damaged core estimated in the TMI-2 accident. Experiments performed were the critical heat flux experiments of (1) bottom-closed counter-current flow and (2) bottom drainage counter-current flow. The flow state was recorded on a VTR with a CCD camera and a fiber scope.The mode of heat transfer was nucleate boiling. As the heat flux was increased, the critical heat flux (CHF) condition was eventually reached. However, no abrupt wall temperature increase was observed then. After the heat flux was increased more, the sharp increase in the wall temperature was perceived. In both (1) and (2) experiments, after the counter-current flow limiting (CCFL) was initiated to restrict the liquid penetration into the flow passage, unstable-intermittent dray patches were formed partially on the heated surface and the CHF point was reached. At this stage, there is no sharp wall temperature increase. After the heat flux (i.e.the upward vapor flow) was elevated more, that liquid penetration was testricted more and the liquid penetrating rate becomes less than the evaporation rate, large-stable dry-patches were created and the wall temperature excursion started. No.difference was observed between the (1) and (2) experiments in the CHF heat flux q_<CHF> and the temperature excursion initiation heat flux q_<EXC>. The values of q_<CHF> in the present experiments were much lower, approximately one order, than values predicted with existing correlations.

所使用的设施是R-113闭环系统。试验段为一段长240 mm的厚壁铜管。外径100 mm，内径34 mm或36 mm。将玻璃管插入试管，形成0.5 mm、1 mm、2 mm或5 mm间隙的垂直窄环流道。在Bond Number标度之后，根据反应堆压力容器和TMI-2事故中估计的损坏堆芯堆积碎片之间的间隙宽度确定了间隙大小。进行了(1)底部封闭逆流和(2)底部排水逆流的临界热流实验。流动状态用摄像机和光纤显微镜记录在录像机上，换热方式为核态沸腾。随着热流密度的增加，最终达到临界热流密度条件。然而，当时并没有观察到壁温的突然升高。在热流密度进一步增加后，壁温急剧升高。在实验(1)和(2)中，在启动逆流限流(CCFL)以限制液体穿透流道后，加热表面部分形成不稳定-间歇性的拖带斑块，达到CHF点。在这个阶段，没有出现墙体温度急剧上升的情况。当热流(即向上蒸汽流)增大时，液体穿透受到更大的限制，穿透率小于蒸发率，形成较大的稳定干斑，壁温开始漂移。在(1)和(2)实验中，CHF热流密度Q_&lt；CHF&gt；和温度漂移起始热流Q_&lt；exc&gt；本实验中的Q&lt；CHF&gt；远低于现有关联式的预测值，约为一个数量级。