• Title/Summary/Keyword: CHF

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Transient Critical Heat Flux Under Flow Coastdown in a Vertical Annulus With Non-Uniform Heat Flux Distribution

  • Moon, Sang-Ki;Chun, Se-Young;Park, Ki-Yong;Baek, Won-Pil
    • 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 .

Experimental Study of the Ultrasonic Vibration Effects on CHF Occurring on Inclined Flat Surfaces (초음파 진동이 경사진 평판에서의 CHF에 미치는 영향에 대한 실험연구)

  • 정지환;김대훈;권영철
    • Journal of Energy Engineering
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    • v.12 no.2
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    • pp.139-144
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    • 2003
  • Augmentation of CHF by ultrasonic vibration in water pool is experimentally investigated under pool boiling condition. The experiments are carried out using copper coated plates and distilled water. Measurements of CHF on flat plate heated surface were made with and without ultrasonic wave and with variations in inclined angle of the surface and water subcooling. Experimental apparatus consists of a bath, power supply, test section, ultrasonic generator, and data acquisition system. The measurements show that ultrasonic wave enhances CHF and its extent is dependent upon inclination angle as well as water subcooling. The rate of increase in CHF increases with an increase in water subcooling while it decreases with an increase in inclination angle. Visual observation shows that the cause of CHF augmentation is closely related with the dynamic behavior of bubble generation and departure in acoustic field.

2相 流動에서의 熱傳達(II) -Post-Dryout 영역-

  • 이영환
    • Journal of the KSME
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    • v.24 no.2
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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.

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A Heuristic Application of Critical Power Ratio to Pressurized Water Reactor Core Design

  • Ahn, Seung-Hoon;Jeun, Gyoo-Dong
    • 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.

Experimental Study and Correlation Development of Critical Heat Flux under Low Pressure and Low Flow Condition

  • Kim, Hong-Chae;Baek, Won-Pil;Kim, Han-Kon;Chang, Soon-Heung
    • Proceedings of the Korean Nuclear Society Conference
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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.

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Prediction of Critical Heat Flux in Fuel Assemblies Using a CHF Table Method

  • Chun, Tae-Hyun;Hwang, Dae-Hyun;Bang, Je-Geon;Baek, Won-Pil;Chang, Soon-Heung
    • Proceedings of the Korean Nuclear Society Conference
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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.

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A study on etching mechanism of SBT thin flim by using Ar/$CHF_3$plasma (Ar/$CHF_34$플라즈마를 이용한 SBT 박막에 대한 식각 메카니즘 연구)

  • 서정우;장의구;김창일;이원재;유병곤
    • 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>.

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Effect of Spacer Grids on CHF at PWR Operating Conditions

  • Ahn, Seung-Hoon;Jeun, Gyoo-Dong
    • Nuclear Engineering and Technology
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    • v.33 no.3
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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.

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환상유로에서의 고압, 저유량조건하의 임계열유속

  • 천세영;전형길;정흥준;문상기;정문기
    • Proceedings of the Korean Nuclear Society Conference
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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.

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반구형 간극에서 CHF 가 발생했을 때 히터표면의 온도변화와 CHF 측정

  • 정지환;박래준;조영로;김상백;김희동
    • Proceedings of the Korean Nuclear Society Conference
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    • pp.675-680
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    • 1997
  • 반구형 간극에서 히터 출력이 임계열유속(CHF)에 이르렀을 때 히터표면의 온도 변화와 CHF 값을 측정하였다. 반구형 히터의 직경은 498mm 이고 간극은 1, 2mm 에서 실험을 수행하였다. 히터표면의 온도는 항상 간극상단의 특정 부분에서부터 증가하기 시작하였다. 즉, 이곳에서 국부적인 dryout이 발생한 것으로 판단된다. 히터의 열속이 증가함에 따라 dryout 부분은 원주방향과 아래방향으로 확장되었다. 한편 임계열유속보다 작은 열속에서는 dryout 영역이 변하지 않는 정상상태가 존재하였으나 임계열유속에서는 열속이 고정되어 있어도 dryout 영역이 스스로 확장되어 나갔다. 이 실험은 계속 진행중이며 현재까지 측정된 CHF 값을 제시하였다. CHF 값은 간극을 대상으로 개발된 기존의 실험식보다 낮게 측정되었다.

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