• Transactions of the Korean Society of Mechanical Engineers B
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• v.25 no.12
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• pp.1831-1843
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• 2001

In this study, convective boiling tests were conducted for enhanced tube bundles. The surface geometry consists of pores and connecting gaps. Tubes with three different pore sizes (d$_{p}$ = 0.20, 0.23 and 0.27 mm) were tested using R-123 and R-l34a for the following range: 8 kg/m$^2$s G 26 kg/m$^2$s, 10 kW/m$^2$ q0 40 kW/m$^2$and 0.1 $\chi$ 0.9. The convective boiling heat transfer coefficients were strongly dependent on heat flux with negligible dependency on mass flux or quality. For the present enhanced geometry (pores and gaps), the convective effect was apparent. The gaps of the present tubes may have served routes for the passage of two-phase mixtures, and enhanced the boiling heat transfer. The convective effect was more pronounced at a higher saturation temperature. More bubbles will be generated at a higher saturation temperature, which will lead to enhanced convective contribution. The pore size where the maximum heat transfer coefficient was obtained was larger for R-l34a (d$_{p}$ = 0.27 mm) compared with that for R-123 (d$_{p}$ = 0.23 mm). This trend was consistent with the previous pool boiling results. For the enhanced tube bundles, the convective effect was more pronounced for R-134a than for R-123. This trend was reversed for the smooth tube bundle. Possible reasoning is provided based on the bubble behavior on the tube wall. Both the modified Chen and the asymptotic model predicted the present data reasonably well. The RMSEs were 14.3% for the modified Chen model and 12.7% for the asymptotic model.model.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.355-362
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• 2012

The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1994.11a
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• pp.135-140
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• 1994

An improved steam generator model has been developed for the DSNP simulation of normal operational transient behavior of CANDU nuclear power plant. For more realistic prediction of steam generator behavior during transient, tube bundle region is divided into two separate control volumes, subcooled region and saturated region, and the variation of thermal hydraulic properties in the control volume is accounted for more realistic estimates of outlet enthalpy of each control volume. Test results for typical CANDU operational transient case show reasonable transient behavior of steam generator with overall CANDU operation and improved operational characteristics of steam generator with power variation.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.241-242
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• 2004

• Journal of Energy Engineering
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• v.15 no.1 s.45
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• pp.45-51
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• 2006

The purpose of this study is to investigate experimentally the heat transfer characteristics of combustion gas or a mixture fuel of LFG and LNG as compared LFG, LNG A Pilot combustion system is constructed. Tube bundle type heat exchangers with vertical and horizontal baffles are used, and the experiment is carried out for different operating conditions, the heating value, the concentration of methane (44.5%, 54.5%). The results show that the Nusselt number of LNG is higher than that of LFG at the same Reynolds number, and in case LFG, the Nusselt number of the mixture of LFG and LNG is larger than that of LFG alone. Therefore, heat transfer is improved by using LFG that is added to LNG pertinently, if and instability of LFG supply will be relaxed.

• Journal of KSNVE
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• v.4 no.2
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• pp.199-208
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• 1994

Two-phase cross-flow exists in many shell-tube heat exchangers such as nuclear steam generators, condensers and reboilers. An understanding of damping and of flow-induced vibration excitation mechanisms in necessary to avoid problems due to excessive tube vibration. In this paper, we present the results of experiments on normal-triangular tube bundles of pitch to tube diameter ratio, p/d, 1.22, 1.32 and 1.47. The bundle were subjected to air-water mixtures to simulate realistic mass fluxes and vapour qualities corresponding to void fractions from 5 to 99%. Damping, fluidelastic instability and turbulence- induced excitation are discussed. The behaivior of damping and two vibration mechanisms are different for intermittent flows from for bubbly flows. The effect of pitch to tube diameter ratio and void fraction is dominant on damping and fluidelastic instability.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.85-91
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• 1997

Cross-flow existing in a shell-and-tube steam generator can cause a tube to vibrate. There are four regions subjected to cross-flow in Yonggwang units 3 and 4 (YGN 3 and 4) steam generators, which are of the same design as the steam generators for Palo Verde nuclear power plant Palo Verde units 1 and 2 steam generators have experienced localized oar at the comers of the cold side recirculating fluid inlet regions. A number of design modifications were made to preclude tube failure in specific regions of YGN 3 and 4 steam generators. Therefore, flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube free lane region. The objective of this experiment is to demonstrate that the tube displacement is less than 0.01 inch rms at 100% of full power flow and to quantify the remaining design margin at 120ft and 140% of full power flow.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.17 no.6 s.123
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• pp.498-506
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• 2007

Using wind tunnel, experimental approaches were employed to investigate fluidelastic instability of tube bundles, subjected to uniform cross flow. There are several flow-induced vibration excitation mechanisms, such as fluidelastic instability, periodic wake shedding resonance, turbulence-induced excitation and acoustic resonance, which could cause excessive vibration in shell-and tube heat exchanges. Fluidelastic is the most important vibration excitation mechanism for heat exchanger tube bundles subjected to cross flow. The system comprised of cantilevered flexible cylinder(s) and rigid cylinders of normal square array, In order to see the characteristics of flow in tube bundles, particle image velocimetry was used. From a practical design point of view, Fluidelastic instability may be expressed simply in terms of dimensionless flow velocity and dimensionless mass-damping. The threshold flow velocity for dynamic instability of cylinder rows is evaluated and the data for design guideline is proposed for the tube bundles of normal square array.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.127-141
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• 2004

Various types of integrated steam generators, which integrate IHTS and a steam generator into a single unit of equipment for an LMR, were analyzed using an analytic solution with some simplification. The analysis showed that the undesirable reversed heat transfer, of which occurrence was previously observed only in an integrated single-region bundle type, can also occur in an integrated double-region bundle type. The mechanism of the reversed heat transfer occurrence in the double-region type is explained and it is shown the mechanism in the double-region type is completely different from that in the single-region type. Based on this finding, a method for preventing the aforementioned heat transfer is suggested. The performance of the four types of the integrated steam generators is assessed. For this assessment, a SG is actually designed for each type and the optimization in the geometric parameters and flow rate are optimized.

• Journal of computational fluids engineering
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• v.12 no.3
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• pp.13-19
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• 2007

A numerical simulation on the heat transfer and flow field was carried out to improve the performance of the shell and tube heat exchanger. The steady incompressible 3-D Navier-Stokes solution is obtained with the actual operational condition and geometry of the heat exchanger. Based on this study, it is noted that the present geometry of the heat exchanger causes poor heat transfer since the air inside shell does not flow through the tube bundle, but around it. The enhancement of the heat transfer can be achieved by the variation of the design factor like the sealing strip located on the top/bottom and middle of the baffle, but it causes the increasement of the pressure drop. In this paper, the effects of the location and size of the sealing strips and flow rate through the heat exchanger on the heat transfer and pressure drop are studied.