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.

使用的设施是R-113闭环。测试部分是厚的壁铜管，长度为240mm。将100mm的OD和34或36mm的ID。将玻璃管插入测试管中，从垂直 - 鼻 - narrular-nular流量通道为0.5,1,2或5mm间隙。键缩放后，从反应器压力容器和TMI-2事故中估计的受损核心的堆积碎片之间的间隙宽度确定了间隙大小。进行的实验是（1）底部关闭的反电流和（2）底部排水反向流动的临界热通量实验。流量状态记录在带有CCD摄像头和纤维范围的VTR上。传热模式是核沸腾的。随着热通量的增加，最终达到了临界热通量（CHF）条件。但是，那时未观察到突然的壁温升高。热通量增加后，壁温度的急剧升高。在（1）和（2）实验中，启动反电流限制（CCFL）以限制液体渗透到流量通道中后，在加热的表面上部分形成了不稳定的临时Dray斑块，并达到了CHF点。在此阶段，没有急剧的壁温升高。热通量（即向上蒸气流）升高后，液体渗透的测试量更高，并且液体穿透速率变得少于蒸发速率，就会产生大稳定的干点，并且开始了壁温温度。在CHF热通量Q_ <CHF>中的（1）和（2）实验与温度偏移引发热通量Q_ <eck>之间观察到了差异。与现有相关性预测的值相比，在本实验中q_ <chf>的值要低得多，大约一个顺序。