• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1126-1135
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• 2022

The MARS code has been assessed for the prediction of onset of flow instability (OFI) in a vertical channel. For assessment, we built an experiment database that consists of experiments under various geometry and thermal-hydraulic condition. It covers pressure from 0.12 to 1.73 MPa; heat flux from 0.67 to 3.48 MW/m²; inlet sub-cooling from 39 to 166 ℃; hydraulic diameters between 2.37 and 6.45 mm of rectangular channels and pipes. It was shown that the MARS code can predict the OFI mass flux for pipes reasonably well. However, it could not predict the OFI in a rectangular channel well with a mean absolute percentage error of 8.77%. In the cases of rectangular channels, the error tends to depend on the hydraulic diameter. Because the OFI is directly related to the subcooled boiling in a flow channel, we suggest a modified subcooled boiling model for better prediction of OFI in a rectangular channel; the net vapor generation (NVG) model and the modified wall evaporation model were modified so that the effect of hydraulic diameter and heat flux can be accurately considered. The assessment of the modified model shows the prediction of OFI mass flux for rectangular channels is greatly improved.

• Journal of the Korean Society of Visualization
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• v.21 no.3
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• pp.33-38
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• 2023

Spray cooling is a method of cooling high-temperature heating elements by spraying droplets. Recently, spray cooling has been proposed for use in next-generation nuclear reactors. When droplets are sprayed onto the outer wall of a heat exchanger tube, a film boiling occurs on the outer wall. Over time, the outer wall temperature decreases, and a liquid film forms on the outer wall, and the heat exchanger outer wall is subsequently cooled by the liquid film. In this case, the liquid film thickness has a great influence on the heat removal performance. In this study, an experimental study was conducted to measure the liquid film thickness distribution in a droplet spray environment. For this purpose, a method using the electrical conductivity of the liquid was adopted.

• Journal of Mechanical Science and Technology
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• v.20 no.8
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• pp.1101-1117
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• 2006

In the present article, recent progress of spray-wall interaction research has been reviewed. Studies on the spray-wall interaction phenomena can be categorized mainly into three groups: experiments on single drop impact and spray (multiple-drop) impingement, and development of comprehensive models. The criteria of wall-impingement regimes (i.e., stick, rebound, spread, splash, boiling induced breakup, breakup, and rebound with breakup) and the post-impingement characteristics (mostly for splash and rebound) are the main subjects of the single-drop impingement studies. Experimental studies on spray-wall impingement phenomena cover examination of the outline shape and internal structure of a spray after the wall impact. Various prediction models for the spray-wall impingement phenomena have been developed based on the experiments on the single drop impact and the spray impingement. In the present article, details on the wall-impingement criteria and post-impingement characteristics of single drops, external and internal structures of the spray after the wall impact, and their prediction models are reviewed.

• Transactions of the Korean hydrogen and new energy society
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• v.29 no.1
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• pp.105-110
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• 2018

The flow boiling heat transfer of water was experimentally investigated on plain and sintered microporous surfaces in a mini-channel. The effects of microporous coating on flow boiling heat transfer of subcooled water were investigated in a 300 mm long mini-channel with a cross section of $20{\times}10mm^2$. The test section has sufficiently long entrance length of 300 mm which provides a fully-developed flow before the channel inlet. The bottom side of the channel was heated by a copper block assembled with a high-density cartridge heater and other sides of the channel were insulated. The microporous surface was fabricated by sintering copper particles with the average particle size of $50{\mu}m$ on the top side of the copper block. Heat transfer measurement was conducted at the mass flux of $208kg/m^2s$ and the heat flux up to $500kW/m^2$. Microporous coated surface showed an earlier boiling incipience compared with plain surface regardless of the mass flux. Microporous coating were significantly attributed to local wall temperature and local heat transfer coefficient for flow boiling.

• 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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.7
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• pp.457-470
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• 2016

Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• Journal of Advanced Marine Engineering and Technology
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• v.33 no.8
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• pp.1137-1143
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• 2009

Experiments were conducted to evaluate pool boiling heat transfer performance between plain and micro finned surfaces with FC-72, which is chemically and electrically stable. Three kinds of micro fins with the dimension of $100{\mu}m\;{\times}\;10{\mu}m$, $150{\mu}m\;{\times}\;10{\mu}m$ and $200{\mu}m\;{\times}\;10{\mu}m$ (width $\times$ height) were fabricated on the surface of a silicon chip. The experiments were carried out on the liquid subcooling of 5, 10 and 15 K under the atmospheric condition. The micro finned surface with a larger fin width of $200{\mu}m$ provided a better pool boiling heat transfer performance. Also, the micro finned surfaces showed a sharp increase in heat flux with increasing wall superheat and a larger heat transfer enhancement compared to a plain surface.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.13 no.4
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• pp.242-251
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• 2001

Flow boiling heat transfer coefficients (HTCs) of R22, R134a, R407C, and R410A were measured for a horizontal plain tube. The test section was made of a copper tube of 8.8mm inner diameter and 1000mm length respectively. The refrigerant was heated by passing hot water through an annulus surrounding the test section. All tests were performed at a fixed refrigerant saturation temperature of $5^{\circ}C$ with mass fluxes of 100~300 kg/$m^2$,/TEX>s. HTCs were measured by two methods: the direct wall temperature measurement method and the indirect Wilson plot method. Experimental results showed that the Wilson plot method was affected greatly by the external test conditions and yielded inconsistent results. For the mass flux of 100kg/$m^2$,/TEX>s, HTCs were almost constant regardless of the quality for a given refrigerant HTCs of R134a and R407C were similar to those of R22 while those of R410A were 60% higher than those of R22. For the mass fluxes of 200 and 300kg/$m^2$,/TEX>s, HTCs of R407C were almost the same as those of R22, while HTCs of R134a and R410A were 12-13% and 20~23% higher than those of R22 respectively. For pure refrigerant, Shah\`s correlation yielded a good agreement with the measured data both qualitatively and quantitatively.

• 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.