• International Journal of Air-Conditioning and Refrigeration
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• v.13 no.4
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• pp.167-174
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• 2005

The evaporation heat transfer coefficient of $CO_2$ (R-744) in a horizontal tube was investigated experimentally. The experiments were conducted without oil in a closed refrigerant loop which was driven by a magnetic gear pump. The main components of the refrigerant loop are a receiver, a variable-speed pump, a mass flow meter, a pre-heater and evaporator (test section). The test section consists of a smooth horizontal stainless steel tube of 7.75 mm inner diameter. The experiments were conducted at mass flux of 200 to $500kg/m^2s$, saturation temperature of $-5^{\circ}C\;to\;5^{\circ}C$, and heat flux of 10 to $40kW/m^2$. The test results showed the evaporation heat transfer of $CO_2$ has greater effect on nucleate boiling than convective boiling. The evaporation heat transfer coefficient of $CO_2$ is highly dependent on the vapor quality, heat flux and saturation temperature. The evaporation heat transfer coefficient of $CO_2$ is very larger than that of R-22 and R-134a. In comparison with test results and existing correlations, the best fit of the present experimental data is obtained with the correlation of Jung et al. But the existing correlations failed to predict the evaporation heat transfer coefficient of $CO_2$. Therefore, it is necessary to develop reliable and accurate predictions determining the evaporation heat transfer coefficient of $CO_2$ in a horizontal tube.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.17 no.12
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• pp.1113-1122
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• 2005

The experimental heat transfer coefficients have been measured for two-phase convective boiling in two circular microtubes with inner diameters of $430{\mu}m\;and\;792{\mu}m$. While the heat transfer was greatly affected by the heat flux in the low quality region, the mass flux played a role in the high quality region. The smaller microtube had greater heat transfer coefficients. When the heat flux is varied from $20kW/m^2\;to\;30kW/m^2\;at\;G=240kg/m^2s$, the difference between the average heat transfer coefficients of the test tube $A(D_i=430{\mu}m)$ and the test tube $B(D_i=792{\mu}m)$ changes from $32.5\%\;to\;52.1\%$. At $G=370kg/m2^s$, the difference between the average heat transfer coefficients changes from $47.0\%\;to\;53.8\%$. A new correlation for the evaporative heat transfer coefficients in microtubes was developed by considering the following factors; the laminar flow heat transfer coefficient of liquid-phase flow, the enhancement factor of the convective heat transfer, and the nucleate boiling correction factor. The correlation developed in this study predicts the experimental heat transfer coefficients within an absolute average deviation of $8.4\%$.

• Proceedings of the SAREK Conference
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• 2006.06a
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• pp.68-73
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• 2006

The present paper deals with an experimental study of boiling heat transfer characteristics of R-290, and is focused on pressure gradient and heat transfer coefficient of the refrigerant flow inside horizontal smooth minichannel with inner diameter of 3.0 mm and length of 2000 mm. The direct heating method applied for supplying heat to the refrigerant where the test tube was uniformly heated by electric current which was applied to the tube wall. The experiments were conducted with R-290 with purity of 99.99% at saturation temperature of 0 to $10^{\circ}C$. The range of mass flux is $50{\sim}250kg/m^2s$ and heat flux is $5{\sim}20kW/m^2$. The heat transfer coefficients of R-290 increases with increasing mass flux and saturation temperature, wherein the effect of mass flux is higher than that of the saturation temperature, whereas the heat flux has a low effect on increasing heat transfer coefficient. The significant effect of mass flux on heat transfer coefficient is shown at high quality, the effect of heat flux on heat transfer coefficient at low quality shows a domination of nucleate boiling contribution. The heat transfer coefficient of the experimental result was compared with six existing heat transfer coefficient correlation. Zang et al.'s correlation(2004) gave the best prediction of heat transfer coefficient.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Metals and materials international
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• v.24 no.6
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• pp.1394-1402
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• 2018

This study investigates the mechanism of MnS precipitation on $Al_2O_3-SiO_2$ inclusions during the solidification of non-oriented silicon steel, especially the influence of the phase structures and sizes of the oxides on the MnS precipitation, by scanning electron microscopy and transmission electron microscopy coupled with energy dispersive spectrometry. The investigation results show that MnS tends to nucleate on submicron-sized $Al_2O_3-SiO_2$ inclusions formed by interdendritic segregation and that it covers the oxides completely. In addition, MnS can precipitate on micron-sized oxides and its precipitation behavior is governed by the phase structure of the oxides. The MnS embryo formed in a MnO-containing oxide can act as a substrate for MnS precipitation, thus permitting further growth via diffusion of solute atoms from the matrix. MnS also precipitates in a MnO-free oxide by the heterogeneous nucleation mechanism. Furthermore, MnS is less prone to precipitation in the $Al_2O_3$-rich regions of the $Al_2O_3-SiO_2$ inclusions; this can be explained by the high lattice disregistry between MnS and $Al_2O_3$.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.

• Journal of the Korean Society of Visualization
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• v.18 no.3
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• Journal of the Korean Society of Propulsion Engineers
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• v.24 no.3
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• pp.59-70
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• 2020

Obtaining external forced convection heat transfer from bubble boiling and validating it with experimental results using cryogenic liquids are suggested to derive total heat transfer coefficient with pool boiling condition in the case of coil type heat exchanger with a bundle of tubes and to overcome the limitation of using the empirical correlation. Experiment is conducted with pool boiling heat transfer of saturate liquid nitrogen with helical coil type heat exchanger using liquid oxygen as hot stream fluid. Experimentally measured heat transfer coefficient is well-agreed with the estimated curve considering nucleate boiling and forced convection induced by bubble rise.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4329-4337
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• 2022

Jet impinging device is designed for decay heat removal on horizontal fuel rods in a low temperature heating reactor. An experimental system with a fuel rod simulator is established and experiments are performed to evaluate water film covering capacity, within 0.0287-0.0444 kg/ms mass flow rate, 0-164.1 kW/m² heating flux and 13.8-91.4℃ feeding water temperature. An effective method to obtain the film coverage rate by infrared equipment is proposed. Water film flowing patterns are recoded and the film coverage rates at different circumference angles are measured. It is found the film coverage rate decreases with heating flux during single-phase convection, while increases after onset of nucleate boiling. Besides, film coverage rate is found affected by Marangoni effect and film accelerating effect, and surface wetting is significantly facilitated by bubble behavior. Based on the observed phenomenon and physical mechanism, dry-out depth and initial dry-out rate are proposed to evaluate film covering potential on a heating surface. A model to predict film coverage rate is proposed based on the data. The findings would have reliable guide and important implications for further evaluation and design of decay heat removal system of new reactors, and could be helpful for passive containment cooling research.