Single-Phase Turbulent Rod Bundle Heat Transfer

Abstract

Abstract The present study investigates single-phase heat transfer coefficients in a 5 × 5 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. Flow-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.