• 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 .

• 에너지공학
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• 제12권2호
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• pp.139-144
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• 2003

본 연구는 초음파 진동의 영향을 받는 풀비등 조건에서의 CHF 열전달 촉진현상을 이해하기 위해 수행되었다. 평판에 얇은 구리박판이 덮여 있는 시료와 증류수를 냉각제로 사용하여 초음파 유무, 경사각도의 변화 및 물의 과냉도 변화에 대하여 평판 가열면에서의 CHF를 측정하였다. 실험장치는 수조, 전원공급장치, 시험부, 초음파 발생장치, 데이터 획득장치 등으로 구성되었다. 실험조건은 3가지 과냉도에 대한 실험과 6가지 시편의 경사각도를 변화시켜 수행되었다. 측정값을 통해 초음파 진동이 CHF를 증진시키며, 그 영향은 경사각의 변화뿐만 아니라 과냉도에 따라서 크기가 달라지는 것을 알 수 있다. 초음파 진동에 의한 CHF증가율은 과냉도가 커질수록, 그리고 시편의 경사각도가 수직에서 수평하향 방향으로 기울어질수록 증가하였다. 가시화 실험을 통해 CHF 증가의 원인이 음장에서 기포 생성과 이탈의 동적거동과 밀접하게 관련되어 있음을 확인하였다.

• 기계저널
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• 제24권2호
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• pp.92-98
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• 1984

열전달영역은 임계열류속점(CHF)을 기준으로 pre-CHF 영역과 post-CHF 영역의 두가지로 대 별된다. Post-CHF 영역에 해당하는 열전달에는 천이비등열전달과 막비등열전달이 있으며 천 이비등은 CHF점과 최소 막비등점 사이에서 일어나는 현상으로 핵비등과 막비등이 조합된 열 전달 기구에 해당하고 막비등은 가열표면이 안정된 증기막에 의해 덮여 있는 상태의 열전달 기 구에 속한다. 전고에서는 pre-CHF와 CHF 열전달 영역의 특성을 살펴보았고 본고에서는 천이 비등과 최소막비등온도 및 막비등에서의 열전달 상관식의 특성을 살펴 보고자 한다.

• Nuclear Engineering and Technology
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• 제34권1호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.356-361
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• 1997

To investigate parametric effect on CHF and to get CHF data, experimental study has been performed with vertical round tubes under the condition of low pressure and low flow (LPLF). Test sections are made of Inconel-625 tube and have the geometry of 8 and 10 mm in diameter, and 0.5 and 1.0 m in heated length. All experiments have been conducted at the pressure of under 9 bar, the mass flux of under 250 kg/$m^2$ and the inlet subcooling of 350 and 450 kJ/kg, for stable upward flow with water as a coolant. Flow regime analysis has been performed for obtained CHF data with Mishima's flow regime map, which reveals that most of the CHF occur in the annular-mist flow regime. General parametric trends of the collected CHF data are consistent with those of previous studies. However, for the pressure effect on CHF, two different are observed; For relatively high mass flux, CHF increases with pressure and far lower mass flux, CHF decrease with pressure. Using modern data regression tool, ACE algorithm, two new CHF correlations for LPLF condition are developed based on local condition and inlet condition, respectively. The developed CHF correlations show better prediction accuracy compared with existing CHF prediction methods.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• 제33권3호
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• pp.283-297
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• 2001

The CHF in PWR rod bundles is usually predicted by the local flow correlation approach based on subchannel analysis while difficulty exists due to the existence of spacer grids especially with mixing vanes. In order to evaluate the effect of spacer grids on CHF, the experimental rod bundle data with various types of spacer grids were analyzed using the subchannel code, COBRA-IV-i. For the Plain grid data, a CHF correlation was described as a function of local flow conditions and heated length, and then the residuals of the CHF in mixing vaned grids predicted by the correlation were examined in various kinds of grids. In order to compensate for the residual, three parameters, distances between grids and from the last grids to the CHF site, and equivalent hydraulic diameter were introduced into a grid parameter function representing the remaining effect of spacer grids predicted most of the CHF data points in plaing grids within $\pm$20 percent error band. Good agreement with the CHF data was also shown when the grid parameter function for mixing vaned grids of a specific design was used to compensate for the residuals of the CHF data predicted by the correlation.

• 한국전기전자재료학회논문지
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• 제13권3호
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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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• 제55권9호
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• pp.3313-3319
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• 2023

A phenomenological study on CHF in a bilaterally heated annulus with equal heat flux on both sides was experimentally performed. The working fluid of the present test was R-134a. Variation characteristics of CHF and transition of CHF occurrence location were investigated under different pressure, mass flux and quality conditions. With the increase of critical thermodynamic quality, it was found that CHF first occurred on the outer surface of the annulus, then simultaneously occurred on both sides, and finally occurred on the inner surface at relatively high critical quality. After the CHF location transitioned to the inner rod, the sharp fall of CHF in the limiting critical quality region was observed. The critical quality corresponding to the CHF location transition decreased with the increase of mass flux and pressure. Besides, CHF in tube, internally heated, externally heated and bilaterally heated annuli were compared under the same hydraulic diameter conditions. The present study is conducive to improving the understanding of complicated CHF mechanism in bilaterally heated annulus, enriching the experimental database, and providing evidence for developing accurate CHF mechanism model for annuli.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 추계학술발표회논문집(1)
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• pp.248-253
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• 1996

한국원자력연구소에서는 최고 16MPa애서 운전할 수 있는 RCS 열수력 Loop를 저작하고 CHF 측정실험을 수행하고 있다. 본 연구에서는 저유량조건에서 압력이 CHF에 미치는 영향을 조사하기 위해 압력 12MPa, 질량유속 300~550kg/$m^2$.s, Test Section입구 과냉도 210 kJ/kg의 범위에서 CHF 실험을 수행하였다. 본 실험조건에서 CHF는 환상류영역에서 발생하였으며 발생기구는 Entrainment-Limited CHF가 지배적이었다. Doerffer의 CHF 상관식은 저압에서 예측능력이 현저하게 떨어지나 고압조건에서는 실험자료를 잘 예측하였다. Bowring의 상관식은 고압 및 저압에서도 양호한 예측능력을 보여주었다.