• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.407-415
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• 2015

A developing post-dryout region is characterized by significant heat transfer enhancements compared with the fully developed post-dryout region. The heat transfer enhancements are mainly due to upstream disturbance and entrained droplets in the region immediately downstream of the critical heat flux location. In this paper, an improved heat transfer correlation is developed for the developing post-dryout regions in vertical tubes over a wide range of flow conditions. The correlation represents a correction factor for the fully developed film-boiling look-up table to be applied to the developing post-dryout region. The new correlation significantly improves the heat transfer prediction in the developing post-dryout regions and provides very good agreement with the experimental data.

• Journal of Mechanical Science and Technology
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• v.16 no.11
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• pp.1540-1549
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• 2002

Experimental results are presented for the effects of coil diameter, system pressure and mass flux on dryout pattern of two-phase flow in helically coiled tubes. Two tubes with coil diameters of 215 and 485 mm are used in the present study, Inlet system pressures range from 0.3 to 0.7 MPa, mass flux from 300 to 500 kg/㎡s, and heat flux from 36 to 80 kw/㎡. A partial dryout region exists because of the geometrical characteristics of the helically coiled tube. The length of the partial dryout region increases with coil diameter and system pressure. On the other hand, it decreases with increasing mass flux. The critical quality at the tube top side increases with mass flux, but decreases with increasing system pressure. This tendency is more notable when the coil diameter is larger. When the centrifugal force effect becomes stronger, dryout starts at the top and bottom sides of the tube. However, when the gravity effect becomes stronger, dryout is delayed at the tube bottom side. In some cases when the mass flux is low, dryout occurs earlier at the outer side than at the inner side of the tube because of film inversion.

• Nuclear Engineering and Technology
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• v.40 no.5
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• pp.387-396
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• 2008

Many experimental analyses for annular film dryouts, which is one of the Critical Heat Flux (CHF) mechanisms, have been performed because of their importance. Numerical approaches must also be developed in order to assess the results from experiments and to perform pre-tests before experiments. Various thermal-hydraulic codes, such as RELAP, COBRATF, MARS, etc., have been used in the assessment of the results of dryout experiments and in experimental pre-tests. These thermal-hydraulic codes are general tools intended for the analysis of various phenomena that could appear in nuclear power plants, and many models applying these codes are unnecessarily complex for the focused analysis of dryout phenomena alone. In this study, a numerical model was developed for annular film dryout using the drift-flux model from uniform heated tube geometry. Several candidates of models that strongly affect dryout, such as the entrainment model, deposition model, and the criterion for the dryout point model, were tested as candidates for inclusion in an optimized annular film dryout model. The optimized model was developed by adopting the best combination of these candidate models, as determined through comparison with experimental data. This optimized model showed reasonable results, which were better than those of MARS code.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.565-574
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• 2005

The CANFLEX Mk-V bundle is designed to improve upon the critical heat flux (CHF) characteristics of the CANFLEX Mk-IV bundle. The main difference between these two bundles is an increase in bearing pad height of about 0.3 mm in the CANFLEX Mk-IV bundle. This change in bearing pad height leads to an increase in gap flow at the bottom of the bundle, primarily eliminating the localized narrow-gap effect that limits the CHF of the CANFLEX Mk-IV bundle. The objective of this paper is to examine the effects of bearing pad height and pressure tube creep on the sheath-temperature distribution, dryout power, and dryout location, as observed ken full-scale bundle tests, between CANFLEX Mk-IV and Mk-V bundles In uncrept and crept channels. A comparison of surface-temperature differences between the top and bottom elements of the bundles showed that increasing the bearing pad height has led to a more homogeneous enthalpy distribution in subchannels of the bundle. Initial dryout locations of the CANFLEX Mk-V bundle were mainly observed at the mid-spacer plane of either the $10^{th}$ (about $80\%$) or $11^{th}$ ($20\%$) bundle in the 12-bundle string, as compared to the mid-spacer and downstream-button planes for the CANFLEX Mk-IV bundle. Dryout power and boiling-length-average (BLA) CHF values exhibit consistent trends and little scatter with varying flow conditions for both types of CANFLEX bundles in uncrept and crept channels. An increase in pressure tube creep has led to a reduction in dryout power (about $20\%$ far the $3.3\%$ crept channel and $27\%$ for the $5.1\%$ crept channel as compared to dryout powers for the uncrept channel). Increasing the bearing pad height of the CANFLEX bundle has led to an increase in the dryout power. Overall, the dryout power of the CANFLEX Mk-V bundle is 7 to $10\%$ higher than that of the CANFLEX Mk-IV bundle at the inlet temperature range of interest (i.e., between 243 and $290^{\circ}C$).

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.636-643
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• 2024

The coolability of the debris bed with a simulant of solidified corium is experimentally studied, focusing on the effects of the structure of the axial stratified debris bed on the dryout heat flux (DHF). DHF was obtained for the four structures with different particle sizes for the axial stratified debris bed under top flooding. The experimental results show that the dryout position of the axial stratified debris bed is formed at the stratified interface indicated by the temperature rise, and the DHF of the axial stratified bed is much lower than that of the homogeneous bed packed with the upper small particles. To predict the dryout heat flux of the stratified debris beds, by considering the properties of the mixed area, a one-dimensional dryout heat flux model of the porous medium is derived from a water and vapor momentum equation for porous medium, two-phase permeability modifications, interfacial drag, and the correlation between capillary pressure and liquid saturation and verified with the experimental data. The modified model can give reasonable results under different structures.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.297-304
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• 2010

In the design of advanced light water reactors (ALWRs) and in the safety assessment of currently operating nuclear power plants, it is necessary to evaluate the possibility of experiencing a degraded core accident and to develop innovative safety technologies in order to assure long-term debris cooling. The objective of this experimental study is to investigate the enhancement factors of dryout heat flux in debris beds by coolant injection from below. The experimental facility consists mainly of an induction heater, a double-wall quartz-tube test section containing a steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of the particle bed was achieved in the radial direction and the axial variation was within 20%. This paper reports the experimental data for 3.2 mm and 4.8 mm particle beds with a 300 mm bed height. The dryout heat density data were obtained for both the top-flooding and the forced coolant injection from below with an injection mass flux of up to $1.5\;kg/m^2s$. The dryout heat density increased as the rate of coolant injection increased. At a coolant injection mass flux of $1.0\;kg/m^2s$, the dryout heat density was ${\sim}6.5\;MW/m^3$ for the 4.8 mm particle bed and ${\sim}5.6\;MW/m^3$ for the 3.2 mm particle bed. The enhancement factors of the dryout heat density were 1.6-1.8.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.537-542
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• 1996

In the present work the influence of various physical parameters on the two-phase flow behavior in a self-heated porous medium has been studied using a numerical model, that is, the effects of heat generation rate, of porosity, of particle size, and of system pressure on the dryout process. To analyze the effect of these parameters, the variation of both liquid volumetric fraction and liquid axial velocity is evaluated at the steady state or at the onset of a first boiled-out region. The analysis of computational results indicate that a qualitative tendency exists between the parameters such as heat generation rate, porosity, effective particle diameter and the temporal development of the liquid volumetric fraction field up to dryout. In addition to these parameters, a variation of fluid properties such as phase density, phase viscosity due to a change of system pressure can be used for gaining insight into the nature of two-phase flow behavior up to dryout.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.220-223
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• 2004

The initial conditions such as the film thickness and the void fraction at the onset of annular flow are required for the analytical dryout model. The Disturbance Wave Instability model(DWI model) is one of the model describing the Annular-to-Intermittent Flow regime Transition(AIFT). The experimental CHF conditions for the uniformly heated tube were compared with the predictions by the modified Levy model, for which the initial conditions at AIFT were estimated by the DWI model. For the flow through long tubes with small inlet subcooling, the effect of AIFT model on the dryout prediction was little. However, the use of DWI model gave better prediction of CHF in a short tube.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.586-591
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• 1996

Through the analysis of many experimental post-dryout data, it is shown that the most probable flow regime near dryout or quench front is not annular flow but churn-turbulent flow when the mass flux is low. A correlation describing the initial droplet size just after the CHF position at low mass flux is suggested through regression analysis. In the post-dryout region at low pressure and low flow, it is found that the suggested one-dimensional mechanistic model is not applicable when the vapor superficial velocity is very low, i.e., when the flow is bubbly or slug flow regime. This is explained by the change of main entrainment mechanism with the change of flow regime. Therefore, the suggested correlation is valid only in the churn-turbulent flow regime ( $j_{g}$ $^{*}$＝0.5~4.5).).