• International Journal of Air-Conditioning and Refrigeration
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• 제6권
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• pp.93-103
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• 1998

The nucleate boiling heat transfer experiments are performed using a ternary refrigerant R407C which is a candidate of alternatives of HCFC 22. The boiling phenomena of R-32, R-125 and R-134a which are the constituent refrigerants of R407C are also investigated. The nucleate boiling heat transfer coefficients of R407C are less than those of HCFC 22 which have the similar physical and transport properties. In our experimental pressure range, which is similar to the operational pressure of air conditioning system, the deterioration of boiling heat transfer coefficients of mixture refrigerant R407C does not appear for moderate wall superheat region. Since nucleate boiling heat transfer coefficients cannot be obtained from ideal mixing law of mixture, Thome's method was used to predict. To account for the heat flux effect and system pressure in Thome's method, the correcting factor, a(P.L1T), was introduced and obtained from experiments for ternary refrigerant R407C.

• 설비공학논문집
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• 제20권9호
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• pp.571-580
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• 2008

In this work, pool boiling heat transfer coefficients(HTCs) of 5 refrigerants of differing vapor pressure are measured on horizontal smooth square surface of 9.52 mm length. Tested refrigerants are R123, R152a, R134a, R22, and R32 and HTCs are taken from $10\;kW/m^2$ to critical heat flux of each refrigerant. Wall and fluid temperatures are measured directly by thermocouples located underneath the test surface and by thermocouples in the liquid pool. Test results show that pool boiling HTCs of refrigerants increase as the heat flux and vapor pressure increase. This typical trend is maintained even at high heat fluxes above $200\;kW/m^2$. Zuber's prediction equation for critical heat flux is quite accurate showing a maximum deviation of 21% for all refrigerants tested. For all refrigerant data up to the critical heat flux, Stephan and Abdelsalam's well known correlation underpredicted the data with an average deviation of 21.3% while Cooper's correlation overpredicted the data with an average deviation of 14.2%. On the other hand, Gorenflo's and lung et al.'s correlation showed only 5.8% and 6.4% deviations respectively in the entire nucleate boiling range.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.138.2-138.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.546-551
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• 1997

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling.

• 에너지공학
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• 제24권1호
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• pp.149-163
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• 2015

본 연구에서는 전산유체역학(Computational Fluid Dynamics, CFD) 해석코드를 사용하여 마스트집합체의 열수력적 안전성에 대한 연구를 수행하였다. 이를 위해 자연대류 벤치마크 문제를 선정하여 CFD 코드의 물리모델을 선정 및 해석 능력을 검증하고 이를 이용하여 마스트집합체에 대한 자연대류 열전달 해석을 수행하였다. 본 연구에서는 Betts et al.의 사각 수직공동에서 난류 자연대류 실험결과를 대상으로 CFD 해석을 수행하여 자연대류 조건에 적용하기 위한 난류 모델로 표준 $k-{\omega}$ 모델을 선정하였다. 이렇게 도출된 난류모델을 CFD코드에 적용하여 Bates et al.에 의해 수행된 PNL(Pacific Northwest Laboratory)의 $2{\times}6$ 번들 실험과 이에 대한 Kwon et al.의 MATRA, Fluent 코드의 해석과 비교 계산을 수행하여 CFD코드의 부수로조건 자연대류 열전달 해석 능력을 검증하였다. 최종적으로 도출된 $k-{\omega}$ 난류 모델을 사용하여 마스트집합체 및 핵연료 집합체에 대한 자연대류 해석을 수행하였다. 해석 결과 수조 내부 및 부수로 내에서 안정적인 자연대류 유동이 발생함을 확인하였으며, 본 유동 조건에서 핵비등이탈비를 계산함으로써 열수력적 안전성을 정량적으로 평가하였다.