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SINGLE-PHASE TURBULENT ROD BUNDLE HEAT TRANSFER

机译:单相湍流棒束传热

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摘要

The present study investigates single-phase heat transfer coefficients in a 5X5 horizontal rod bundle representing a water-cooled nuclear reactor. The rod bundle assembly consists of a square array of parallel rods held in place by support grids. How-enhancing features such as disc blockage or split-vane pairs are often added to the downstream edge of the support grids in an attempt to improve the heat transfer performance of the rod bundle assembly. The effects on the local heat transfer coefficients of several support grid designs are examined in the present study. The local, average heat transfer coefficients are measured using a heated copper sensor for standard, disc, and split-vane pair grid designs. In addition, lateral flow fields are obtained for the split-vane pair grid design using Particle Image Velocimetry (PIV) measurements. Results indicate that the local heat transfer coefficient is enhanced just downstream of the support grids. This enhancement decays to the fully-developed value of heat transfer by five hydraulic diameters downstream of the support grid for all of the grid designs. Heat transfer measurements of the split-vane pair grid design indicate a region of decreased heat transfer below the fully-developed value. Identification of swirl migration and weak exchange of fluid between subchannels in the lateral flow fields of the split-vane pair grid design provides insight into the decreased region of heat transfer.
机译:本研究调查了代表水冷核反应堆的5X5水平棒束中的单相传热系数。杆束组件包括由支撑格栅固定在适当位置的平行杆的方形阵列。通常将诸如圆盘阻塞或裂片对之类的如何增强特征添加到支撑格栅的下游边缘,以试图改善杆束组件的传热性能。在本研究中检查了几种支撑格栅设计对局部传热系数的影响。使用用于标准,圆盘和分流叶片对栅格设计的加热铜传感器测量局部平均传热系数。此外,使用粒子图像测速(PIV)测量获得的分流叶片对栅格设计的横向流场。结果表明,刚好在支撑格栅的下游提高了局部传热系数。对于所有的格栅设计,这种增强都会衰减到支撑格栅下游五个水力直径的传热值。分体叶片对栅格设计的传热测量表明,传热减少的区域低于完全展开的值。识别分流叶片对格栅设计的横向流场中子通道之间的旋流迁移和流体的弱交换,有助于深入了解传热的减小区域。

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