• Nuclear Engineering and Technology
• /
• 제56권2호
• /
• pp.410-418
• /
• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2001년도 추계학술대회논문집B
• /
• pp.3-8
• /
• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• 대한기계학회논문집B
• /
• 제30권1호
• /
• pp.24-31
• /
• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In thi present study, the large scale vortex flow(LSVF) is generated by rearranging the mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane. Heat transfer in the rod bundle occurs greatly at the same direction to cross flow, and maximum temperature at the surface of bundle drops about 1.5K

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2008년도 추계학술대회A
• /
• pp.1367-1372
• /
• 2008

A plate type supporting structure of a tube bundle in axial flow generates a certain band of a high frequency periodic excitation of a vortex shedding and/or a flow separation due to sharp edge of the plate thickness and a severe pressure drop due to a cross-sectional area of the supports. With a design consideration of the low vibration and a small flow resistance, the analysis method is uniquely confined to an experimental approach because a complex geometry of a cylindrical tube bundle and/or physical phenomena related to the fluid-structure interaction of tube bundle in a flow impede a theoretical or a numerical approach. A 5x5 cylindrical tube bundle with 5 supports which were discretely located along the bundle's axis was tested in the FIVPET hydraulic test loop for a design evaluation and an analysis perspectives. A high frequency flow-induced vibration of the supporting structures of the cylindrical tube bundle was measured at a outer surface of a supporting structure through a transparent flow housing by the laser dopper vibrometer. Pressure drop in-between three measurement distances was measured by the differential pressure transmitter. High frequency vibration and pressure drop fairly depends on the geometric design of supporting structure. So, these two parameters would be used as a qualitative design variables for design evaluation and analysis.

• Nuclear Engineering and Technology
• /
• 제36권2호
• /
• pp.153-164
• /
• 2004

A computational fluid dynamics (CFD) analysis has been performed to investigate the turbulent flow and heat transfer in a triangular rod bundle with pitch-to-diameter ratios (P/D) of 1.06 and 1.12. Anisotropic turbulence models predicted the turbulence-driven secondary flow in a triangular subchannel and the distributions of the time mean velocity and temperature, showing a significantly improved agreement with the measurements from the linear standard $k-{\epsilon}$ model. The anisotropic turbulence models predicted the turbulence structure for a rod bundle with a large P/D fairly well, but could not predict the very high turbulent intensity of the azimuthal velocity observed in the narrow flow region (gap) for a rod bundle with a small P/D.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2005년도 추계학술대회논문집
• /
• pp.503-506
• /
• 2005

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle (5$\times$5) is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid like the coolant mixing performance and to obtain the Flow-Induced Vibration (FIV) characteristics of the rod bundle over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the bundle prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the small scaled test bundle. For the damping factor of the rod bundle and the grid case at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the rod bundle is about 0.7% with reasonable error of 2% for the previous results. Nonlinear behavior of the rod bundle might be stem mainly Iron the rod-grid support configuration.

• 한국전산유체공학회지
• /
• 제10권1호
• /
• pp.63-66
• /
• 2005

A computational fluid dynamics(CFD) analysis has been made for fully developed turbulent flow in a triangular bare rod bundle with a pitch to diameter ratio (P/D) of 1.123. The nonlinear turbulence models predicted the turbulence-driven secondary flow in the triangular subchannel. The nonlinear quadratic κ-ε models by Speziale[1] and Myong-Kasagi[2] predicted turbulence structure in the rod bundle fairly well. The nonlinear quadratic and cubic k-ε models by Shih et al.[3] and Craft et al.[4] showed somewhat weaker anisotropic turbulence. The differential Reynolds stress model by Launder et al.[5} appeared to over predict the turbulence anisotropy in the rod bundle.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
• /
• pp.532-537
• /
• 1998

Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-l34a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLRX bundle is found to be about 20 % higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within ${\pm}\;5\;%$ error.

• 설비공학논문집
• /
• 제13권7호
• /
• pp.595-601
• /
• 2001

We applied PIV method to obtain instantaneous and ensemble averaged velocity fields from the first row to the fifth row of a staggered tube bundle. The Reynolds number based on the tube diameter and the maximum velocity was set to be 4,000. Remarkably different natures are observed in the developing bundle flow. Such differences are depicted in the mean recirculating bubble length and the vorticity distributions. The jet-like flow seems to be a dominant feature after the second row and usually skew. However, the ensemble averaged fields show symmetric profiles and the flow characteristics between the third and fourth measuring planes are not so different. comparison between the PIV data and the RANS simulation yields severe disagreement in spite of the same Reynolds number. It can be explained that the distinct jet-like unsteady motions are not to be accounted in th steady numerical analysis.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2001년도 춘계학술대회논문집D
• /
• pp.195-200
• /
• 2001

An experimental study on critical heat flux (CHF) has been performed for water flow in a uniformly heated vertical 3 by 3 rod bundle under low flow and a wide range of pressure conditions. The objective of this study is to investigate the parametric trends of CHF with 3 by 3 rod bundle test section where three unheated rods exist. The general trends of the CHF are coincident with previous understandings. At low flow and system pressure above 3 MPa, some critical qualities are larger than 1.0 due to counter-current flow in test sections. Since there is a supply of water to the heated section from unheated section, the maximum CHFs at system pressure between 2 and 4 MPa are not shown.