• Nuclear Engineering and Technology
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• 제56권2호
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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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• 제42권3호
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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.

• Nuclear Engineering and Technology
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• 제47권4호
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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.

• Nuclear Engineering and Technology
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• 제20권2호
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• pp.112-119
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• 1988

이 연구의 목적은 가혹한 사고후 손상된 원자로심을 모의한 열이 발생하는 성충 데브리층에서의 강제대류 드라이아웃 열유속 자료를 실험 적으로 얻고자 한 것이다. 여 기서는 일정한 층의 깊이와 냉각재 유입온도 조건하에서 선정된 몇 가지 크기의 입자로 성층을 형성한 데브리층에서 주로 냉각재질량유속이 드라이아웃 열유속에 미치는 영향을 관찰하였다. 이 실험적 연구로부터 얻은 주요 결과는 다음과 같다 (1) 성층 데브리층에서의 드라이아웃 열유속은 질량유속의 증가와 더불어 증가하며 그 증가의 경향은 크기가 균일한 입자층에 대한 것과 유사하다. (2) 이론치와 실험치와의 비교에서 입자직경으로는 표면적 평균 직경을 사용하는 것이 적합하다.

• Nuclear Engineering and Technology
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• 제18권4호
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• pp.273-280
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• 1986

이 연구의 목적은 가혹한 사고후 손상된 원자로심을 모의한 열을 발생하는 데브리층에서의 강제대류 드라이아웃 열유속을 실험적으로 얻고져 한 것이다. 이 연구에서 냉각재 순환장치를 사용하여 대기압하에서 냉각재가 상향 강제대류하는 유도 가열된 강구 입자층에서의 드라이아웃 열유속을 얻었다. 이 실험에서는 주로 강제대류 드라이아웃에 대한 질량유속.입자크기.입자층의 높이 및 냉각재의 서브클링의 영향이 관찰되었다. 실험은 입자직경이 1.5, 2.5, 3.0 및 4.0mm의 탄소강입자를 55mm 내경의 Pyrex 유리용기에 넣어 고주파유도 전류를 통해 가열하여 이루어졌다. 냉각재로서 증류수를 질량유속 0~3.5kg/$m^2$s로 변화시키어 공금하고 층의 높이는 55mm와 110mm, 냉각재 유입온도는 2$0^{\circ}C$와 8$0^{\circ}C$로 변화시켰다. 주요 실험결과는 다음과 같다. (1) 드라이아웃 열유속은 상향 강제 대류 질량유속과 입자크기내의 증가에 따라 증가한다. (2) 질량유속이 없는 경우 드라이아웃 열유속은 기존 연구결과와 같이 입자직경에 의존한다. (3) 얕은 입자층에서의 드라이아웃 열유속은 깊은 입자층의 것보다 얼마간 높다.

• Journal of Mechanical Science and Technology
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• 제16권11호
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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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• 제40권5호
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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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• 제37권6호
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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 $290^{\circ}C$).

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.528-533
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• 1997

A CANFLEX(CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5㎜) to reduce maximum element power rating and buttons to enhance the critical heat flux(CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant H-l34a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power(CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF's for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only lot high pressure conditions. It indicates that the button is a useful design lot CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions.

• Journal of Mechanical Science and Technology
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• 제15권7호
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• pp.921-930
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• 2001

A new theoretical critical heat flux (CHF) model was developed for the forced convective flow boiling at high pressure, high mass velocity, and low quality. The present model for an intermittent vapor blanket was basically derived from the sublayer dryout theory without including any empirical constant. The vapor blanket velocity was estimated by an axial force balance, and the thickness of vapor blanket was determined by a radial force balance for the Marangoni force and lift force. Based on the comparison of the predicted CHF with the experimental data taken from previous studies, the present CHF model showed satisfactory results with reasonable accuracy.