• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1373-1376
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• 2008

A supporting structure for a long tube bundle of a large diameter is considered in this paper. The primary purpose of the present study is to develop a spacer grid structure for a so-called "dual cooled nuclear fuel", which has been being studied for a nuclear power uprate. The outer diameter of the fuel rod increases considerably from the conventional one. So a completely new shape of the supporting structure (spacer grid) needs to be developed. One of the challenges is to insert a supporting tube into the cross points of the grid straps. To meet a supporting performance, the load vs. displacement characteristics should be obtained. So the present study focuses on the finite element analysis technology to evaluate the characteristics through a parametric study. As a result, major influencing parameters are investigated for an optimized spacer grid design.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.3
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• pp.171-181
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• 2017

Two-phase cross flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. The drag force acting on a tube bundle subjected to air/water flow is evaluated experimentally. The cylinders subjected to two-phase flow are arranged in a normal square array. The ratio of pitch to diameter is 1.35, and the diameter of the cylinder is 18 mm. The drag force along the flow direction on the tube bundles is measured to calculate the drag coefficient and the two-phase damping ratio. The two-phase damping ratios, given by the analytical model for a homogeneous two-phase flow, are compared with experimental results. The correlation factor between the frictional pressure drop and the hydraulic drag coefficient is determined from the experimental results. The factor is used to calculate the drag force analytically. It is found that with an increase in the mass flux, the drag force, and the drag coefficients are close to the results given by the homogeneous model. The result shows that the damping ratio can be calculated using the homogeneous model for bubbly flow of sufficiently large mass flux.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.440-452
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• 1994

Nuclear fuel bundles are designed such that the heat flux at a-fuel pin surface should not exceed the critical heat flux (CHF) during normal operation and anticipated transient. Therefore, evaluation of the CHF for fuel bundle is demanded in an exact and reliable manner. One of the major concerns with the current application of CHF correlations is that the CHF based on circular tubes is applied to the fuel bundle subchannel analysis, mainly in terms of the hydraulic diameter with correction factors which may result in a source of possibly large uncertainties in CHF prediction. The hydraulic diameter does not recognize the local properties of fluid nor such effect as the surface curvature; the turbulence action on the convex surface is much more pronounced than that on the concave surface. Even for the tube having concave curvature, the effect of tube diameter on CHF becomes important with decreasing diameter. These facts imply that the convex curvature effect is significant and crucial to the reliable CHF prediction. This paper reviews and discusses analytical and experimental aspects of effect of transverse convex curvature in incompressible turbulent flow and heat transfer, and on CHF. Flow models to quantify this effect are briefly mentioned and future works are recommended.