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SCWR类三角形通道超临界流动传热定位格架结构影响研究 被引量:1

Study on Influence of Supercritical Flow and Heat Transfer Grid Spacer Inner Structural of Supercritical Water Cooled Reactor Triangular Channel
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摘要 基于带定位格架类三角形子通道超临界水流动传热试验,数值研究了棒径为8 mm,栅距比为1.4的超临界水冷堆(Supercritical Water Cooled Reactor,SCWR)类三角形通道超临界流动传热定位格架结构影响,分析了同型定位格架典型结构参数和不同定位格架型式对堆芯通道超临界流动传热特性的影响规律。研究结果表明:定位格架可强化堆芯通道超临界水传热,同型格架本体厚度越大,压力损失越高,格架处壁面温度越低,局部换热能力越好,当增大格架本体厚度,弱化程度无明显差异;阻流片型定位格架下游局部换热能力提高显著,阻流片直径越大,上游压力越大,局部壁温越低,换热系数越高,增大阻流片直径可减小传热弱化区域大小,强化传热能力;不同定位格架型式对比研究发现交错叶片型弱化区域最大,阻流片型定位格架弱化区域最小,阻流片型定位格架具有最佳的传热强化效果。 Based on the supercritical water flow heat transfer test of a triangular coneshaped core channel with a grid spacer,The numerical study on the supercritical watercooled Reactor SCWR-type triangular channel supercritical flow heat grid spacer inner structural with a rod diameter of 8 mm and a pitch ratio of 1.4.The influence of the typical structural parameters of the same-type grid and the different grid space patterns on the supercritical flow heat transfer-characteristics of the core channel are studied.The results show that the grid space can enhance the supercritical water heat transfer of the core channel.The larger the thickness of the same type grid body,the higher the pressure loss,the lower the wall temperature at the grid,the better the local heat transfer capacity.When the thickness of the grid body is increased,there is no significant difference in the degree of weakening.The local heat transfer capacity of the choke-type grid spacer is improved,the larger the diameter of the baffle,the larger the upstream pressure,the lower the local wall temperature,the higher the heat transfer coefficient,and the larger the diameter of the baffle can reduce the heat transfer,weaken the size of the area and enhance heat transfer capacity.The comparison of different grid space types shows that the stray blade type weakening area is the largest,the choke-type grid spacer weakening area is the smallest,and the choke-type grid space has the best heat transfer enhancement.
作者 徐维晖 闫友志 王为术 崔强 XU Weihui;YAN Youzhi;WANG Weishu;CUI Qiang(Institute of Thermal Energy Engineering,North China University of Water Resources and Electric Power,Zhengzhou of Henan Prov.450011,China;Huadian Zhengzhou Mechanical Design Research Institute Co.Ltd,Zhengzhou of Henan Prov.450011,China)
出处 《核科学与工程》 CAS CSCD 北大核心 2019年第6期872-877,共6页 Nuclear Science and Engineering
基金 国家自然科学基金项目(51876024) 河南省高校科技创新团队支持计划资助项目(16IRTSYHN017) 河南省科技创新人才计划(154100510011)
关键词 超临界水冷堆(SCWR) 定位格架 类三角形 传热特性 数值研究 Supercritical water cooled reactor(SCWR) Grid spacer Quasi-triangle Heat transfer characteristics Numerical studies
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  • 1陈畏葓,张虹,朱力,熊万玉.CFD方法在棒束定位格架热工水力分析中的应用研究[J].核动力工程,2009,30(S1):34-38. 被引量:12
  • 2熊万玉,陈炳德,肖泽军.棒束定位格架内单相流体三维流场研究[J].原子能科学技术,2005,39(4):326-329. 被引量:13
  • 3Yamagata K, Nishikawa, K, Hasegawa S, et al. Forced convection heat transfer to supercritical water flowing in tubes[J]. International Journal of Heat and Mass Transfer, 1972, 15: 2575-2593.
  • 4Xiaojing Zhu, Qincheng Bi, Dong Yang, et al. An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube[J]. Nuclear Engineering and Design, 2009, 239(2): 381-388.
  • 5Wang J G, Li H X, Yu S Q, et al. Comparison of the heat transfer characteristics of supercritical pressure water to that of suberitical pressure water in vertically-upward tubes [J]. International Journal of Multiphase Flow, 2011, 37(7): 769-776.
  • 6Holloway M V, McClusky H L, Beasley D E. The effect of support grid features on local, single-phase heat transfer measurements in rod bundles[C]. Proc. of HT2003 ASME Summer Heat Transfer Conference, Las Vegas, Nevada, USA, July 21-23, 2003.
  • 7Kim S H,Kim Y I,BaeY Yet al.Numerical Simulation of the Vertical upward Flow of Water in a Heated Tube at Supercritical Pressure. Proc.of ICAPP04 . 2004
  • 8Cheng X,Kuang B,Yang Y H.Numerical Analysis of Heat Transfer in Supercritical Water Cooled Flow Chan-nels. Nuclear Engineer The . 2006
  • 9Yang J,Oka Y,Ishiwatari Yet al.Numerical Investi-gation of Heat Transfer in upward Flows of Supercritical Water in Circular Tubes and Tight Fuel Rod Bundles. Nuclear Engineer The . 2006
  • 10Silin V A,Voznesensky V A,Afrov A M.The Light Water Integral Reactor with Natural Circulation of the Coolant at SCP B-500SKDI. Nuclear Engineer The . 1993

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