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Featured researches published by Hanyang Gu.


2017 25th International Conference on Nuclear Engineering | 2017

Effects of Heater Material and Surface Orientation on Heat Transfer Coefficient and Critical Heat Flux of Nucleate Boiling

Yong Mei; Yechen Zhu; Botao Zhang; Shengjie Gong; Hanyang Gu

External reactor vessel cooling (ERVC) is the key technology for In-Vessel Retention (IVR) to ensure the safety of a nuclear power plant (NPP) under severe accident conditions. The thermal margin of nucleate boiling heat transfer on the reactor pressure vessel (RPV) lower head is important for ERVC and of wide concern to researchers. In such boiling heat transfer processes, the reactor vessel wall inclination effect on the heat transfer coefficient (HTC) and critical heat flux (CHF) should be considered. In this study, experiments were performed to investigate the effects of heater material and surface orientation on the HTC and CHF of nucleate boiling. Copper and stainless steel (SS) surfaces were used to perform boiling tests under atmosphere pressure. The orientation angle of both boiling surfaces were varied between 0° (upward) and 180° (downward). The experimental results show that the surface orientation effects on the HTC is slight for both the copper surface and the SS surface. In addition, the relationship of measured CHF values with the inclination angles was obtained and it shows that the CHF value changes little as the inclination angle increases from 0° to 120° but it decreases rapidly as the orientation angle increases towards 180° for both boiling surfaces. The material effect on CHF is also observed and the copper surface has higher CHF value than the SS surface. Based on the experimental data, a correlation for CHF prediction is developed which includes both the surface orientation effect and the heater material effect.Copyright


2013 21st International Conference on Nuclear Engineering | 2013

Experimental Research of Supercritical Water Heat Transfer in Different Channels

Hongbo Li; Meng Zhao; Hanyang Gu; Fei Wang; Jianmin Zhang; Yong Zhang; Jue Yang

The experimental research of supercritical water heat transfer has been performed on the supercritical water multipurpose test loop (SWAMUP) with tube, annular channel, and bundles. The normal heat transfer, heat transfer deterioration (HTD) and heat transfer enhancement were observed; and the heat transfer experimental data were obtained. The experimental results show that: the first kind of HTD caused by buoyancy effect only occurs with low mass flow velocity and high heat flux when the fluid temperature is below pseudo-critical point in all the tested channels and the second kind of HTD caused by acceleration effect always occurs when the fluid temperature reaches pseudo-critical point in tube and annular channel; the heat transfer enhancement occurs when the fluid temperature reaches pseudo-critical point with low mass flow velocity in tube; and the heat transfer enhancement in bundles is caused by the space grids. It is concluded that the heat transfer in bundles is better than in other tested channels.Copyright


Nuclear Engineering and Design | 2012

Experimental and numerical investigation of turbulent convective heat transfer deterioration of supercritical water in vertical tube

Ge Zhang; Hao Zhang; Hanyang Gu; Yanhua Yang; Xu Cheng


Nuclear Engineering and Design | 2014

Experimental study on heat transfer of supercritical Freon flowing upward in a circular tube

Siyu Zhang; Hanyang Gu; Xu Cheng; Zhenqin Xiong


Nuclear Engineering and Design | 2008

CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

Hanyang Gu; Xu Cheng; Y.H. Yang


Nuclear Engineering and Design | 2011

Fluid-to-fluid scaling of heat transfer in circular tubes cooled with supercritical fluids

Xu Cheng; X.J. Liu; Hanyang Gu


Nuclear Engineering and Design | 2010

CFD analysis of thermal–hydraulic behavior of supercritical water in sub-channels

Hanyang Gu; Xu Cheng; Y.H. Yang


Progress in Nuclear Energy | 2015

Experimental study on the sub-atmospheric loop heat pipe passive cooling system for spent fuel pool

Zhenqin Xiong; Cheng Ye; Minglu Wang; Hanyang Gu


Annals of Nuclear Energy | 2015

Experimental study on heat pipe heat removal capacity for passive cooling of spent fuel pool

Zhenqin Xiong; Minglu Wang; Hanyang Gu; Cheng Ye


Nuclear Engineering and Design | 2014

The thermal performance of a loop-type heat pipe for passively removing residual heat from spent fuel pool

Zhenqin Xiong; Hanyang Gu; Minglu Wang; Ye Cheng

Collaboration


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Zhenqin Xiong

Shanghai Jiao Tong University

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Yao Xiao

Shanghai Jiao Tong University

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Xu Cheng

Shanghai Jiao Tong University

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Shengjie Gong

Shanghai Jiao Tong University

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Zhenxiao Hu

Shanghai Jiao Tong University

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Yong Mei

Shanghai Jiao Tong University

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Minglu Wang

Shanghai Jiao Tong University

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Shuo Chen

Shanghai Jiao Tong University

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Da Liu

Shanghai Jiao Tong University

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Cheng Ye

Shanghai Jiao Tong University

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