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为探究影响核反应堆运行安全的非能动余热排出热交换器(PRHR HX)管外过冷沸腾传热特性，搭建单根 C 型管装置，结合可视化技术开展实验研究。结果表明：随着成核数量的增加及汽泡尺寸的增大，水平及竖直观测点对应水温分别由 95.1 和 84.7 ℃上升至饱和温度，传热系数分别在 7 930.9～14 545.9 W·m-2·℃-1及 2 876.9～8 742.2 W·m-2·℃-1持续增大，过冷度分别由 4.9 及 15.3 ℃减小至接近 0 ℃，壁温及过热度的变化均在 2.0 ℃内。热流密度实验值与q ～Δθnsat模型和叠加模型对应经验公式预测值偏差较大，基于削弱系数模型分别对水平及竖直段拟合了过冷沸腾经验公式，平均偏差为 1.9%和 6.4%。此外，对水平段拟合了饱和沸腾经验公式，偏差在±5% 以内。
Abstract：In order to explore heat transfer characteristics out of passive residual heat removal heat exchanger(PRHR HX) tubes, which affects operation safety of nuclear reactors, a single C-shaped tube setup was built and experimental study was conducted with visualization technology. The results show that with the increase of nucleation number and bubble size, the water temperatures at horizontal and vertical observation points increase from 95.1 and 84.7 ℃ to saturation temperature, respectively. Correspondingly, the heat transfer coefficients increase in the range of 7 930.9-14 545.9 W·m-2·℃-1 and 2 876.9-8 742.2 W·m-2·℃-1, and the subcooling degrees decrease from 4.9 and 15.3 ℃ to nearly 0 ℃. The wall temperature and superheat degree vary within2.0 ℃. The differences between experimental heat fluxes and predicted values with empirical correlations based on theq - Δθnsa t and superposition model were obvious. Empirical correlations were fitted for horizontal and vertical sections based on the weakening coefficient model under subcooled boiling conditions, and the average prediction errors were 1.9% and 6.4%, respectively. Horizontal section under saturation boiling conditions was also fitted and the error was within ±5%.
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基本信息:
中图分类号:TK124
引用信息:
[1]刘延斌,王学生,王浩,等.单根C型管外过冷沸腾传热特性实验及分析[J],2022,36(03):346-353.