• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3473-3478
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• 2007

A dual side cooling annular fuel having internal and external coolant channels has many advantages basically due to low fuel temperature and high DNBR margin, which can make a significant increase of core power density possible. So recently a 12x12 square annular fuel array was proposed for the fuel assembly to be reloaded without structural interference with operating reactors of OPR-1000s. Even through the inherent potential of the annular fuel on the high power density, it may be seriously eroded in the case of a severe unbalanced mass flux split to the internal and external channels in standpoint of DNB. Mass flux split is determined pressure drop characteristics between inner and outer channels. The spacer grids binding fuel array influence greatly the pressure drop in outer channels and the mass flux split. As an important factor of DNB behavior, the enthalpy differences at both channel exits were evaluated using the mass flux splits.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2236-2241
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• 2008

The surface heat flux of nuclear fuel rod is the most important factor which can affect safety of reactor and fuel. If fuel rod surface heat flux exceeds the CHF(${\underline{C}}ritical$ ${\underline{H}}eat$ ${\underline{F}}lux$), fuel can be damaged. In case of double cooled annular fuel, which is under developing, contains two coolant channels. Therefore, a generated heat in the fuel pellet can move to inner or outer channel and heat flow direction is decided by both sides heat resistance which varied by dimension and material property change which caused by temperature and irradiation. The new program(called DUO) was developed. For the calculation of surface heat flux, a both sides convection by inner/outer coolant, s gap temperature jump and conduction in the fuel are modeled. Especially, temperature and time dependent fuel dimension and material property change are considered during the iteration. A sample calculation result shows that the DUO program has sufficient performance for annular fuel thermal hydraulics design.

• 설비공학논문집
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• 제12권1호
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• pp.67-74
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• 2000

One of the important application of a contact melting process is a latent thermal energy storage owing to its high heat flux. In some previous works, the split fins have been employed in order to enhance the melting speed. In the present work, the close contact melting was experimentally investigated using an ice as specimen for both split and non-split fins. It was shown that the contact melting by split fins increases the melting rate compared to that of non-split ones.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.382-395
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• 2002

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• International Journal of Air-Conditioning and Refrigeration
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• 제9권2호
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• pp.1-7
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• 2001

One of the important applications of a contact melting process is a latent thermal energy storage system owing to its high heat flux and small temperature variation. In some previous works, the split fins have been employed in order to enhance the melting rate. In the present work, the direct contact melting was experimentally investigated using an ice as specimen for both split and non-split fins. It was shown that the contact melting by split fins increases the melting rate compared to that of non-split ones.

• 설비공학논문집
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• 제12권11호
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• pp.1004-1011
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• 2000

Experimental and computational studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Axial development of mean and turbulent velocities in the subchannels were measured by the 2-color LDV system. Eddy diffusivity heat flux model and $k-varepsilon$ model were employed to analyze the turbulent heat and fluid flows in the subchannel. The turbulence generated by split mixing vanes has small length scales so that they maintain only about $10 D_H$ after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up to $25 D_H$after the spacer gird.

• 대한기계학회논문집B
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• 제29권1호
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• pp.151-158
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• 2005

A characteristic of discharge for a foam spray nozzle with various parameters was investigated. The discharge patterns from a fire foam spray nozzle are important to evenly spray over a maximum possible floor area. Two parameters of a foam spray nozzle were chosen, and compared with those from the standard one. Also, in order to evaluate the performance of discharged foam agents used to protect structures from heat and fire damages, the thermal characteristics of fire-protection foams were experimentally investigated. A simple repeatable test for fire-protection foams subjected to fire radiation was developed. This test involves foam generation equipment, a fire source for heat generation, and data acquisition techniques. Results show that the bubble size of foam is increased by large inside diameter of orifice or closed air hole, but phenomenon of discharge angle and expansion ratio is opposite. For the case of the open air hole, liquid film of a circular cone discharges with formation, growth, split and fine grain. In case of the closed air hole, a pillar of foam solution discharges with that. Though the temperature gradient in the foam increases with increased foam expansion ratio. it is not change with increased intensity of heat flux.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1143-1156
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• 2009

The SPACE code that is based on a multi-dimensional two-fluid, three-field model is under development for licensing purposes of pressurized water reactors in Korea. Among the participating research and industrial organizations, KAERI is in charge of developing the physical models and correlation packages for the constitutive equations. This paper introduces a developed wall-to-fluid heat transfer package for the SPACE code. The wall-to-fluid heat transfer package consists of twelve heat transfer subregions. For each sub-region, the models in the existing safety analysis codes and the leading models in literature have been peer reviewed in order to determine the best models which can easily be applicable to the SPACE code. Hence a wall-to-fluid heat transfer region selection map has been developed according to the non-condensable gas quality, void fraction, degree of subcooling, and wall temperature. Furthermore, a partitioning methodology which can take into account the split heat flux to the continuous liquid, entrained droplet, and vapor fields is proposed to comply fully with the three-field formulation of the SPACE code. The developed wall-to-fluid heat transfer package has been pre-tested by varying the independent parameters within the application range of the selected correlations. The smoothness between two adjacent heat transfer regimes has also been investigated. More detailed verification work on the developed wall-to-fluid heat transfer package will be carried out when the coupling of a hydraulic solver with the constitutive equations is brought to completion.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.331-336
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• 1996

The effects of a mixing vane on air-water mixed flow have been experimentally studied in this work, to investigate the basic mechanisms that the mixing vane affects critical heat flux (CHF). Experiment was performed for various flow rates focusing on bubbly flow and annular flow patterns. Acrylic tube (1.7m long, 11 mm I.D.) and the split vane type mixing vane were used, and ring-type conductance probes were used to measure the liquid film thickness in annular flow. Experimental results show that, (a) bubbly-to slug flow transition and churn-to-annular flow transition occur respectively near the mixing vane compared to the tests without mixing vane, (b) in bubbly flow region, the mixing vane breaks the bubbles into smaller ones and forwards bubbles to the center region of the tube by the centrifugal force, (c) the liquid film thickness in annular flow is decreased near the mixing vane for mass fluxes.