为了研究海洋核动力平台非能动余热排出换热器(PRHR HX)池沸腾换热特性，设计搭建功率比1∶50的实验装置，研究PRHR HX运行过程中池沸腾传热特性，评价传统经验关系式在预测PRHR HX池沸腾换热系数时的适用性。实验结果表明PRHR HX局部池沸腾换热不均匀，PRHR HX下部沸腾强度明显弱于上部；随着热负荷升高，池沸腾换热趋于均匀。实验数据拟合所得到的半经验换热关联式与实验结果符合良好，偏差在±9%以内。研究结果可为海洋核动力平台非能动安全系统设计提供参考。
In order to investigate the heat transfer characteristics of the marine nuclear power platform passive residual heat removal heat exchanger, an experimental platform with power ratio 1:50 was established to simulate marine nuclear power platform working conditions, and the heat transfer characteristics of the marine nuclear power platform PRHR HX was studied. The results demonstrate that the heat transfer characteristics of the PRHR HX is not uniform, the upper part of PRHR HX is more efficient in transfer heat than the lower part. With the increase of heat load, the PRHR HX heat transfer tends to be uniform. The experimental pooling boiling heat transfer coefficient was compared with calculated values of new correlations. The new correlation show a good agreement with experimental data and the relative deviation is less 9%. This paper can provide a reference for the design of floating nuclear power plant reactor safety system.
2022,44(6): 84-89 收稿日期：2022-01-23
DOI：10.3404/j.issn.1672-7649.2022.06.018
分类号：TL33
基金项目：国家重点研发计划项目（2017YEC0307800）
作者简介：李鹏拯(1995-)，男，硕士研究生，研究方向为海洋核动力装置
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