共查询到18条相似文献,搜索用时 234 毫秒
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针对高压工况下偏离泡核沸腾(DNB)型临界热流密度(CHF)的特点,重新构建了Weisman & Pei模型的本构关系式;针对高压工况下干涸(Dry-out)型CHF,比较分析了Kataoka、Celata以及Hewitt?3个沉积率和夹带率计算关系式的结果。基于以上两类改进的CHF模型,建立了一个适用于高压工况的、结合DNB型和Dry-out型沸腾临界机理的CHF模型。采用高压工况下管内垂直向上流动沸腾CHF实验数据对建立的CHF机理模型进行了验证,分析了热工参数和几何参数的趋势效应。 相似文献
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水平圆管临界热流密度实验研究 总被引:1,自引:1,他引:0
对水平圆管内低质量流速临界热流密度(CHF)进行了实验研究和分析。实验研究发现,水平流动圆管沸腾临界发生在圆管加热壁面顶部。通过对沸腾临界发生时圆管出口的质量含汽率和流型进行分析发现,本文研究的参数范围内沸腾临界时的出口含汽率高,流型为环状流,沸腾临界类型为干涸型(Dryout)。将经验公式预测值与实验结果进行比较发现,Bowring公式和Lookup table的预测值远大于CHF的实验值。导致此现象出现的主要原因为:Bowring公式和Lookup table是基于竖直流动CHF实验数据开发的模型,水平流动时在重力的作用下环状流液膜呈非均匀分布,顶部液膜干涸提前触发沸腾临界造成CHF值降低。 相似文献
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Tomio Okawa Akio Kotani Isao Kataoka Masanori Naitoh 《Nuclear Engineering and Design》2004,229(2-3):223-236
The axial variation of film flow rate in annular regime was analyzed to predict the critical heat flux in water saturated flow boiling in various vertical channels. In the analyses, several quantities including the deposition and entrainment rates of droplets were evaluated with the correlations that were based on the separate experimental data of film flow rate and droplet flow rate in annular flow; model closure was achieved without introducing arbitrary constants. The predicted results were compared with extensive data of critical heat flux in flow boiling. Fairly good agreements were observed for the critical heat fluxes in round tubes with uniform and axially non-uniform heating and in non-circular shaped channels without sharp corners. It is hence expected that the important processes for the onset of critical heat flux condition in annular regime were satisfactorily described in the model. However, the critical heat fluxes were generally overestimated in thin rectangular channels with sharp corners. The non-uniform distribution of liquid film in channel cross-section would be the primary reason of the disagreement. 相似文献
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An investigation of flow characteristics and critical heat flux in vertical upward round tube 总被引:1,自引:0,他引:1
FAN Pu QIU Sui-Zheng JIA Dou-Nan 《核技术(英文版)》2006,17(3):170-176
Prediction of critical heat flux (CHF) in annular flow is important for the safety of once - through steam generator and the reactor core under accident conditions. The dryout in annular flow occurs at the point where the film is depleted due to entrainment, deposition, and evaporation. The film thickness, film mass flow rate along axial distribution, and CHF are calculated in vertical upward round tube on the basis of a separated flow modcl of annular flow. The theoretical CHF values are higher than those derived from experimental data, with error being within 30%. 相似文献
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Critical heat flux (CHF) is one of the major parameters for calculating safety margin of nuclear power plant systems based on US-Nuclear Regulatory Commission's safety regulations (defined under 10CFR50-46). Exceeding the CHF point suddenly increases the fuel clad temperature so that it may result in damage to the nuclear fuel pins. Dryout is a special CHF physical phenomenon occurred in Boiling Water Reactors (BWRs). Firstly, the present work identifies several existing dryout correlations and notes the major assumptions for the derivation of them. It follows with the derivation of a film splitting model (originally proposed by Chun et al., in 2003) for annular geometry where the liquid film is broken into an inner and outer film. Film splitting can result in better prediction of the critical film thickness where dryout can occur due to local variations especially due to turbulence and other effects. Secondly, an alternate approach to solving the film splitting model of Chun et al. is employed to evaluate the model against the same experimental data. It was observed that liquid entrainment and deposition into and out of the film layer were the missing components in the model and the model predictions shown in Chun et al. (2003) so that the model predictions with alternative solution methods were compared against experimental data by excluding entrainment/de-entrainment phenomena. Thirdly, the film split model is updated by adding entrainment/de-entrainment models to make the film splitting model more realistic. Fourthly, new film splitting model is compared against both tubular and pipe data. It was observed that entrainment and deposition are significant contributors in the new mechanistic dryout model. The mass flux of the inner film is mostly overestimated while the mass flux of the outer film is overestimated for some of the experimental cases when compared to the original correlation. 相似文献
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Critical Heat Flux (CHF) is an important parameter for the thermal design of any heat generating system, most importantly, nuclear reactors. Owing to the complex mechanisms of CHF there has been a large proliferation of the correlations, each having narrow range of validity, which shows that the empirical correlation is not an appropriate approach for the CHF prediction for a wide range of validity. This limitation has led to the development of the phenomenological approach of the CHF prediction. The film dryout mechanism is applicable to the high quality CHF corresponding to the annular flow pattern in which the progressive depletion of the liquid film leads to dryout. The basic concern in the prediction of dryout is the accuracy in the evaluation of the droplet deposition and entrainment. There are various models for the estimation of the entrainment and deposition of droplets. However, most of these models are based on the air-water data at the atmospheric conditions and hence their applicability to the BWR conditions needs to be confirmed. Some of the models are based on the steam-water data which needs to be validated for the dryout prediction under BWR conditions. In this paper, the film dryout modelling has been carried out for the prediction of CHF using appropriate models for entrainment fraction and deposition coefficient. The results have been compared with the CHF data generated to substantiate the appropriateness of the selected models under BWR conditions. 相似文献
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Sang-Ki Moon Se-Young Chun Seok Cho Won-Pil Baek 《Nuclear Engineering and Design》2005,235(21):643-2309
An experimental study of the critical heat flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3 × 3 rod bundle under low flow and a wide range of pressure conditions. The experiment was especially focused on the parametric trends of the CHF and the applicability of the conventional CHF correlations to a return-to-power conditions of a main steam line break accident whose conditions might be a low mass flux, intermediate pressure, and a high inlet subcooling. The effects of the mass flux and pressure on the CHF are relatively large and complicated in the low pressure conditions. At a high mass flux or a low critical quality, the local heat flux at the CHF location sharply decreases with an increasing local critical quality. However, at a low mass flux or a high critical quality, the local heat flux at the CHF location shows a nearly constant value regardless of the increase of the critical quality. The CHF data at the very low mass flux conditions are correlated well by the churn-to-annular flow transition criterion or the flow reversal phenomena. Several conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux of below about 100 kg/(m2 s). 相似文献
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A previously developed semi-empirical model for adiabatic two-phase annular flow is extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients. 相似文献
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It is generally assumed in the mechanistic film dryout model that the critical heat flux (CHF) arises when liquid film calculated from evaporation, droplet entrainment and deposition gets dryout. The dryout of film is usually assumed when film thickness becomes zero. However, it was indicated that the complete dryout assumption can estimate CHF well for uniform heating case but cannot simulate accurately for non-uniform heating case. The critical film thickness concept may be an appropriate approach physically because there is a possibility of instantaneous disappearance of liquid film when it gets very thin. Therefore, a critical dryout film thickness correlation was developed to properly model dryout phenomenon together with MARS code based on experimental data. The modified version of MARS implementing a newly developed critical dryout film thickness correlation was assessed using various dryout data including those of non-uniform heating case and flow reduction transient test. The prediction results showed improved agreement with the experimental data. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(9):643-653
A method was developed based on the conservation lows to predict critical heat flux (CHF) causing liquid film dryout in two-phase annular-mist flow in a uniformly heated narrow tube under BWR conditions. The applicable range of the method is within the pressure of 3–9 MPa, mass flux of 500–2,000 kg/m2·s, heat flux of 0.33–2.0 MW/m2 and boiling length-to-tube diameter ratio of 200–800. The two-phase annular-mist flow was modeled with the three-fluid streams with liquid film, entrained droplets and gas flow. Governing equations of the method are mass continuity and energy conservation on the three-fluid streams. Constitutive equations on the mass transfer which consist of the entrainment fraction at equilibrium and the mass transfer coefficient were newly proposed in this study. Confirmation of the present method were performed in comparison with the available film flow measurements and various CHF data from experiments in uniformly heated narrow tubes under high pressure steam- water conditions. In the heat flux range (q“<2MW/m2) practical for a BWR, agreement of the present method with CHF data was obtained as, (Averaged ratio)±(Standard deviation)=0.984±0.077, which was shown to be the same or better agreement than the widely-used CHF correlations. 相似文献