• Proceedings of the KSME Conference
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• 2001.11b
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• pp.191-198
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• 2001

A numerical analysis of the heat and mass transfer and pressure drop characteristics in modular shell and tube bundle heat exchanger was carried out. Finite Concept Method based on FVM and $k-\varepsilon$ turbulent model were used for this analysis. Condensation heat transfer enhanced total heat transfer rate $4\sim8%$ higher than that of dry heat exchanger. With increasing humid air inlet velocity, temperature and relative humidity, and with decreasing heat exchanger aspect ratio and cooling water velocity, total heat and mass transfer rate could be increased. Cooling water inlet velocity had little effect on total heat transfer.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4561-4569
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• 2023

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.

• Proceedings of the SAREK Conference
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• 2009.06a
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• pp.266-271
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• 2009

The present study numerically investigated the effect of the geometry of the flattened tube on the thermal performance of a high temperature generator (HTG) of a double effect LiBr-water absorption system. The heat transfer tubes of the HTG were arranged behind a metal fiber burner. The heat transfer of the tubes of HTG were consisted with a set of circular and flattened tubes in series. FLUENT, as a commercial code, was applied for estimating the thermal performance of the HTG. Key parameters were the tube arrangement in the HTG. Temperature and velocity profiles in the HTG were calculated to estimate the thermal performance of the HTG. The heat transfer rate of a HTG tube was increased, and the gas temperature around the flattened tube was decreased as the pitch ratio was increased. The heat transfer rate for the circular tube bundle with the pitch ratio of 2.48 were larger by 10% respectively than that of 2.10 and the heat transfer rate for the flattened tube bundle with the pitch ratio of 1.88 were larger by 36% respectively than that of 1.63.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.22 no.10
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• pp.704-709
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• 2010

The effect of oil on flow boiling of R-123 in a plain tube bundles was experimentally investigated for a range of quality and heat flux. It is shown that the heat transfer coefficient decreased as the oil concentration increased. Comparison with the previous pool boiling data reveals that the reduction of heat transfer coefficient by oil is more pronounced in pool boiling, and the difference increased with the increase of oil concentration and heat flux. Within the experimental range, the variation of mass flux or quality has negligible effect on the heat transfer coefficient.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.10
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• pp.961-969
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• 2013

In this study, the numerical analysis of a turbulent flow around both a staggered and an inline tube bundle was conducted using ANSYS CFX V.13, a commercial CFD software. The flow was assumed to be steady, incompressible, and isothermal. According to the CFD Best Practice Guideline, the sensitivity study for grid size, accuracy of the discretization scheme for convection term, and turbulence model was conducted, and its result was compared with the experimental data to estimate the applicability of the CFD Best Practice Guideline. It was concluded that the CFD Best Practice Guideline did not always guarantee an improvement in the prediction performance of the commercial CFD software in the field of tube bundle flow.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.3-8
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• 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.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.17 no.11
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• pp.1014-1021
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• 2005

Flow channels with non-circular cross-sections are encountered in a wide variety of heat exchangers. Accurate friction factor and Colburn j factor data are essential for the design and viable applications of such heat exchangers. In this study, an analysis is con ducted on heat transfer and pressure drop characteristics for tube-bundle heat exchanger with various arrangements of tubes, of which their geometry could easily be modified from a circular one in a harsh environment. The parameters investigated are aspect ratio, pitch, and inclined angle of tubes. The results obtained are: (1) Aspect ratio has larger influence on the j and f factor than pitch; (2) As aspect ratio increases, both j and f factors decrease; (3) The high performance is achieved when the pitch and aspect ratio are in the range of 1.5${\~}$2.5 and 1.25${\~}$2.0, respectively; and (4) the inclined arrangements of tubes show unfavorable results for both heat transfer and pressure drop characteristics in spite of the positive possibility of condensate removals in a latent heat recovery system.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.4
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Transactions of the Korean Society of Mechanical Engineers
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• v.17 no.1
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• pp.200-214
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• 1993

This paper presents the results of fluidelastic instability analysis performed for the U-tube bundle of a Westinghouse model 51 steam generator, one of the recirculating types designed at an early stage, in which the principal region of external cross-flow is associated with the U-bend portion of tube. The prerequisites for this analysis are detailed informations of the secondary side flow conditions in the steam generator and the free vibration behaviours of the U-tubes. In this study, the three-dimensional two-phase flow field in the steam generator has been calculated employing the ATHOS3 steam generator two-phase flow code and the ANSYS engineering analysis code has been used to calculate the free vibration responses of specific U tubes under consideration. The assessment of the potential instability for the suspect U-tubes, which is the final analysis process of the present work, has been accomplished by combining the secondary side velocity and density distributions obtained from the ATHOS3 prediction with the relative modal displacement and natural frequency data calculated using the ANSYS code. The damping of tubes in two-phase flow has been deduced from the existing experimental data by taking into account the secondary side void fraction effect. In operation of the steam generator, the tube support conditions at the tube-to-tube support plate intersections due to either tube denting degradation or deposition of tube support plate corrosion products or ingression of dregs. Thus, various hypothetical cases regarding the tube support conditions at the tube-to-tube support plate intersections have been considered to investigate the clamped support effects on the forced vibration response of the tube. Also, the effect of anti-vibration bars support in the curved portion of tube has been examined.