• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.13 no.3
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• pp.183-187
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• 2000

In this study the SrBi$_2$Ta$_2$$O_{9}$ (SBT) thin films were etched by using magnetically enhanced inductively coupled Ar/CHF$_3$plasma as function of CHF$_3$/(Ar+CHF$_3$)gas mixing ratio. Maximum etch rate of SBT thin films was 1650 $\AA$/min and the selectivities of SBT to Pt and photoresist(PR) were 1.35 and 0.94 respectively under CHF$_3$/(Ar+CHF$_3$) of 0.1 For study on etching mechanism of SBT thin film X-ray photoelectron spectroscopy (XPS) surface analyses and secondary ion mass spectrometry (SIMS) mass analysis of etched SBT surfaces were performed. Among the elements of SBT thin film. M(Sr, Bi, Ta)-O bonds are broken by Ar ion bombardment and form SrF and TaF$_2$by chemical reaction with F. SrF and TaF$_2$are removed more easily by Ar ion bombardment. Scanning electron microscopy(SEM) was used for the profile examination of etched SBT film and the cross-sectional SEM profile of etched SBT film under CHF$_3$(Ar+CHF$_3$) of 0.1 was about 85$^{\circ}$X>.

• Nuclear Engineering and Technology
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• v.34 no.1
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• pp.68-79
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• 2002

The approach for evaluating the critical heat flux (CHF) margin using the departure from nucleate boiling ratio (DNBR) concept has been widely applied to PWR core design, while DNBR in this approach does not indicate appropriately the CHF margin in terms of the attainable power margin-to-CHF against a reactor core condition. The CHF power margin must be calculated by increasing power until the minimum DNBR reaches a DNBR limit. The Critical Power Ratio (CPR), defined as the ratio of the predicted CHF power to the operating power, is considered more reasonable for indicating the CHF margin and can be calculated by a CPR orrelation based on the heat balance of a test bundle. This approach yields directly the CHF power margin, but the calculated CPR must be corrected to compensate for many local effects of the actual core, which are not considered in the CHF test and analysis. In this paper, correction of the calculated CPR is made so that it may become equal to the DNB overpower margin. Exemplary calculations showed that the correction tends to be increased as power distribution is more distorted, but are not unduly large.

• Nuclear Engineering and Technology
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• v.34 no.4
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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 .

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.528-533
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• 1997

A CANFLEX(CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5㎜) to reduce maximum element power rating and buttons to enhance the critical heat flux(CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant H-l34a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power(CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF's for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only lot high pressure conditions. It indicates that the button is a useful design lot CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.14 no.2
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• pp.8-14
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• 2018

A boiling heat transfer is used in various industry such as power generation systems, heat exchangers, air-conditioning and refrigerations. In the boiling heat transfer system, the critical heat flux (CHF) is the important factor, and it indicated safety of the system. It has kept up studies on the CHF enhancement. Recently, it is reported the CHF enhancement, when working fluid used the nanofluid with excellent thermal properties. Therefore, in this study, we investigated the influence of nano particles coated surface for heat transfer enhancement in pure water, oxidized multi-wall carbon nanotube nanofluid (OMWCNT), and oxidized graphene nanofluid (OGraphene). Nanoparticles were coated for 120 sec on the surface, and we measured the CHF at the flow velocities of 0.5, 1.0, and 1.5 m/sec, respectively. As the results, both of the OMWCNT and OGraphene nanofluids increased up to about 34.0 and 40.0%.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.136-136
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.99-99
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• 2004

• Journal of the Korean Institute of Telematics and Electronics S
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• v.34S no.2
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• pp.94-103
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• 1997

The WCHF-FSJTC (wavelet circular harmonic function frequency selective joint transform correlator) using th wavelet transformed CHF as the reference image in FSJTC is proposed for rotation invariant pattern recognition. Since the wavelet transform has the property of feature extraction, the proposed system can have the better DC (discrimination cpability) and the higher SNR(signal to noise ratio) compared with the conventional CHF-CJTC(circular harmonic function conventional joint transform correlator). And since the structure of the proposed system is FSJTC which can eliminate auto-correlation and cross-correlation between input images, it can eliminate false alarm caused by the overlapping among correlation peaks. The used wavelet functio is the morlet function, which is proper for the reference image used in this paper. the optimal dialation parameter and oscillation frequency of the wavelet function are also achieved with varying the parameters of the wavelet function. The computer simulation shows that the proposed system has the best performance when the dilation parameter is 0.8 and the oscillation frequency is 0.48.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.532-537
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• 1998

Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-l34a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLRX bundle is found to be about 20 % higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within ${\pm}\;5\;%$ error.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.269-285
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• 2002

Water heat transfer experiments were carried out in a uniformly heated annulus with a wide range of pressure conditions. The local heat transfer coefficients for saturated water (low boiling have been measured just before the occurrence of the critical heat flux (CHF) along the length of the heated section. The trends of the measured heat transfer coefficients were quite different from the conventional understanding for the heat transfer of saturated flow boiling. This discrepancy was explained from the nucleate boiling in the liquid film of annular flow under high heat flux conditions. The well-known correlations were compared with the measured heat transfer coefficients. The Shah and Kandlikar correlations gave better prediction than the Chen correlation. However, the modified Chen correlation proposed in the present work showed the best agreement with the present data among correlations examined .