摘要
为研究一体化布置的核供热堆在发生破口失水事故中破口大小和从中间回路排出热量减少对断流过程的影响，选用不同的破口尺寸和不同的二回路工作状态，在５ＭＷ核供热堆热工水力模拟回路ＨＲＴＬ－５上进行了实验研究。稳态运行工况的系统压力为１．５ＭＰａ，在发生小破口失水事故后，加热功率维持为额定功率的５％以模拟剩余发热情况。实验研究并比较了不同条件下压力、温度、循环流量、液位和失水量等重要参数的变化。这些实验数据为核供热堆的安全分析提供了实验依据。
The main loops of 5 MW nuclear heating reactor and 200 MW nuclear heating reactor are integral natural circulation systems. In case of loss of coolant accident (LOCA), while the water level reaches the upper edge of the inlet of the heat exchanger, the flow break occurs. It has influence on cooling core and stability of the main loop. Some tests for different break area and different state of the secondary loop are described and their results are discussed in the paper. Tests are carried out on the thermohydraulics test system HRTL 5 of the nuclear heating reactor. The system pressure is 1 5 MPa, the heating power under the accident is kept 5% of the steady state power. Results show the influences of the break area and the state of the secondary loop on the natural circulation break process, and that the temperature of the heated element isn't too high and the element couldn't be burn out.
出处
《原子能科学技术》
EI
CAS
CSCD
北大核心
1997年第5期413-417,共5页
Atomic Energy Science and Technology
基金
国家"八五"科技攻关项目
关键词
核供热堆
失水事故
自然循环
断流
安全
破口
Nuclear heating reactor LOCA Natural circulation Flow break Safety
