An experimental and analytical study of the effect of axial power profile on CHF

The effect of axial heat flux distribution (AFD) on the critical heat flux (CHF) was investigated. CHF measurements were obtained with HFC-134a cooled vertical tubes having four non-uniform and one uniform AFD profiles. The HFC-134a test conditions covered a pressure range from 1.6 to 2.4 MPa, a mass-flux range from 2.8 to 4.7 Mg m⁻² s⁻¹, and an inlet-quality range from −0.9 to 0. The water-equivalent pressure and mass-flux ranges are 10–14 MPa and 4–6.5 Mg m⁻² s⁻¹, respectively.In general, the observed AFD effect on critical power is small at high inlet subcoolings. At low inlet subcoolings, the critical power for the inlet-peak profile is up to 15% higher than that for the outlet-peak profile. A local conditions analysis showed that the AFD has the strongest effect on CHF at high dryout qualities. CHF values for non-uniform AFDs could be 50% lower than those for the uniform AFD. The AFD effect on CHF becomes diminished with decreasing dryout quality.Four different approaches to account for the effect of AFD on CHF were assessed against the experimental values from the current experiment. The boiling-length-average heat-flux approach with the boiling-length starting point at the onset of annular flow (OAF) provided the best prediction of the critical power and the CHF location.     

Dryout Heat Flux for Core Debris Bed, (I)

For gaining basic data on decay heat coolability of debris bed in the post-accident heat removal, measurement of dryout heat flux was made, with stagnant water as coolant, in a 50mm I. D. pyrex glass cylinder vessel. The fuel debris bed subjected to decay heat was simulated by steel ball particles which were inductively heated with a power supply of 20 kHz and 30 kW. The bed was made of homogeneous size particles. An emphasis was placed on the influence of system pressure and particle size. The experiment covered the ranges over the steel ball diameters of 0.3–4.0 mm and the system pressure of 0.02–0.5 MPa.

The experimental results, as a whole, agreed fairly well with the prediction based on Lipinski's 0-D model with respect to the dependence of dryout heat flux both on pressure and on particle size. In detail, however, the dryout heat flux deviates toward a lower value at a higher pressure while to a higher value for a smaller size particle bed. Comparison of the results between the free and fixed beds suggests that the deviation to the higher side will be attributed to the channeling and/or levitation.     

Basic investigations on debris cooling

An experimental investigation of boiling phenomena in inductively heated particle beds has been performed. The major aim of these experiments is to provide data for validating numerical codes used in reactor safety. The experiments can be divided in three parts: boiling experiments, dryout experiments and quenching experiments. In boiling experiments, the pressure gradients have been measured along the bed height for different flow modes, different heat inputs and different system pressures. In dryout experiments, the minimum heat input has been determined for which the particle bed starts to superheat significantly above the saturation temperature. The final test series deals with the cool down behaviour of strongly superheated particles by flooding them with cold water. The initial temperatures ranged from 200 up to 900 °C in top-quenching experiments and from 230 up to 450 °C in bottom-quenching experiments. All experiments were performed with pre-oxidised stainless steel balls of 6 and 3 mm diameter in a cylindrical crucible. The bed height was 640 mm and the bed diameter was 125 mm for boiling and dryout experiments, respectively 150 mm for quenching experiments. The experimental results are compared with various available dryout models.     

An experimental study on pressure drop and dryout heat flux of two-phase flow in packed beds of multi-sized and irregular particles

This paper is concerned with debris bed coolability in a postulated severe accident of light water reactors, where the debris particles are irregular and multi-sized. To obtain and verify the friction laws predicting the hydrodynamics of the debris beds, the drag characteristics of air/water single- and two-phase flow in a particulate bed packed with multi-sized spheres or irregular sand particles were investigated on the POMECO-FL test facility. The same types of particles were then loaded in the test section of the POMECO-HT facility to obtain the dryout heat fluxes of the particulate beds heated volumetrically. The effective (mean) particle diameter is 2.25 mm for the multi-sized spheres and 1.75 mm for the sand particles, determined from the Ergun equation and the measured pressure drop of single-phase flow through the packed bed. Given the effective particle diameter, both the pressure drop and the dryout heat flux of two-phase flow through the bed can be predicted by the Reed model. The experiment also shows that the bottom injection of coolant improves the dryout heat flux significantly and the first dryout position is moving upward with increasing bottom injection flowrate. Compared with top-flooding case, the dryout heat flux of the bed can be doubled if the superficial velocity of coolant injection is 0.21–0.27 mm/s. The experimental data provides insights for interpretation of debris bed coolability (how to deal with the multi-sized irregular particles), as well as high-quality data for validation of the coolability analysis models and codes.     

CHF and dryout point in vertical narrow annuli

Experimental study associated with CHF and dryout point in narrow annuli is conducted with 1.5 mm and 1.0 mm gap, respectively. Distilled water is used as work fluid. The parameters examined were: pressure from 2.0 MPa to 4.0 MPa; mass flux from 26.0 kg/(m² s) to 69.0 kg/(m² s); heat flux from 10 kW/m² to 70 kW/m²; exit equilibrium mass quality from 0.52 to 1.08.It is found that CHF monotonously increases with mass flux in internally heated annuli and bilaterally heated annuli. However, the observed trends are not similar to that in externally heated annuli. The CHF is not affected significantly by mass flux.Critical qualities of dryout point (X_DO) decreases with mass flux and increases with inlet qualities. Under the same conditions X_DO in outer tube are always larger than that in inner tube. According to experimental data, a criterion for the appearance of dryout point for bilaterally heated has been presented.The comparison with the correlations [КУТАТЕЛАДЗЕ, C.C., 1979. Тедплоэнергетика, No. 6] and experimental data indicates that the existing correlations applied to tube cannot predict X_DO in narrow annuli well. Based on experimental data, a new correlation is developed.     

Void fraction and heat transport in two-dimensional mixed size particle beds with internal heat sources

The paper discusses the boiling heat transfer from a porous bed with internal heat sources and refers to the configuration in a nuclear reactor after a partial core melt. The flow of coolant, the temperature and the local liquid/vapor distribution were investigated in a two-dimensional configuration. Experiments were conducted using monodisperse beds as well as a mixture of two different particle sizes with a total porosity below 20%. In some tests the bed was supported by a shell of porous material to create a gap along the bottom of the test container. Water was used for tests up to 9% of the critical pressure, while other tests were made with R134a up to 44% of the critical pressure. The maximum heating rate realized inductively was 730 kW/m². The experiments have been compared to analytical results with a one-dimensional approach.It is shown that in contrary to the situation in small cylindrical configurations the heat transfer was increased by large buoyancy driven convective flows. If there was a gap along the container bottom an additional flow of liquid improved the coolability of the bottom region even if the upper part of the particle bed was already overheated. In case of high density ratios (water at low pressure), the measurements indicated a strong enhancement of the coolant flow above a certain minimum heating rate resulting in decreasing vapor fraction values which were nearly independent of the system pressure. This was assumed to be caused by the appearance of vertical channels through which the vapor could flow through the particle bed.     

Heat transport and void fraction in granulated debris

The paper discusses the boiling heat transfer from a porous bed with internal heat sources and refers to the configuration in a nuclear reactor after a partial core melt. The flow of coolant, the temperature and the local liquid/vapor distribution were investigated in a two-dimensional configuration. Experiments were conducted using monodisperse beds as well as a mixture of two different particle sizes with a total porosity below 20%. In some tests the bed was supported by a shell of porous material to create a gap along the bottom of the test container. Water was used for tests up to 9% of the critical pressure, while other tests were made with R134a up to 44% of the critical pressure. The maximum heating rate realized inductively was 730 kW/m². The experiments have been compared to analytical results with a one-dimensional approach.It is shown that in contrary to the situation in small cylindrical configurations the heat transfer was increased by large buoyancy driven convective flows. If there was a gap along the container bottom an additional flow of liquid improved the coolability of the bottom region even if the upper part of the particle bed was already overheated. In case of high density ratios (water at low pressure), the measurements indicated a strong enhancement of the coolant flow above a certain minimum heating rate resulting in decreasing vapor fraction values which were nearly independent of the system pressure. This was assumed to be caused by the appearance of vertical channels through which the vapor could flow through the particle bed.     

The effect of lateral flooding on the coolability of irregular core debris beds

The coolability of ex-vessel core debris is an important issue in the severe accident management strategy of, e.g. the Nordic boiling water reactors. In a core melt accident, the molten core material is expected to discharge into the containment and form a porous debris bed on the pedestal floor of a flooded lower drywell. The debris bed generates decay heat which must be removed by boiling in order to stabilize the debris bed and to prevent local dryout and possible re-melting of the material. The STYX test facility which consists of a cylindrical bed of irregular alumina particles has been used to investigate the effect of lateral coolant inflow on the dryout heat flux of the particle bed. The lateral flow was achieved by downcomers attached on the sides of the test rig. The downcomers provide coolant into the lower region of the bed by natural circulation. Both homogenous and stratified bed configurations have been examined. It was observed that the dryout heat flux is increased by 22-25% for the homogenous test bed compared to the case with no lateral flooding. For the stratified configuration with a fine particle layer on top of the bed, no significant increase in the dryout heat flux was observed. The experiments have been analyzed by using the MEWA-2D code. Models which include explicit consideration of gas-liquid friction were used in the calculations in order to realistically capture the lateral flow configuration.     

Critical heat flux of water in vertical round tubes at low pressure and low flow conditions

An experimental study on critical heat flux (CHF) has been performed for water flow in vertical round tubes under low pressure and low flow (LPLF) conditions to provide a systematic data base and to investigate parametric trends. Totally 513 experimental data have been obtained with Inconel-625 tube test sections in the following conditions: diameter of 6, 8, 10 and 12 mm; heated length of 0.31.77 m; pressure of 106951 kPa; mass flux of 20277 kg m⁻² s⁻¹; and inlet subcooling of 50654 kJ kg⁻¹, thermodynamic equilibrium critical quality of 0.3231.251 and CHF of 1081598 kW m⁻². Flow regime analysis based on Mishima & Ishii’s flow regime map indicates that most of the CHF occurred due to liquid film dryout in annular-mist and annular flow regimes. Parametric trends are examined from two different points of view: fixed inlet conditions and fixed exit conditions. The parametric trends are generally consistent with previous understandings except for the complex effects of system pressure and tube diameter. Finally, several prediction models are assessed with the measured data; the typical mechanistic liquid film dryout model and empirical correlations of (Shah, M.M., 1987. Heat Fluid Flow 8 (4), 326–335; Baek, W.P., Kim, H.G., Chang, S.H., 1997. KAIST critical heat flux correlation for water flow in vertical round tubes, NUTHOS-5, Paper No. AA5) show good predictions. The measured CHF data are listed in Appendix B for future reference.     

Boiling experiments for the validation of dryout models used in reactor safety

The coolability of fragmented corium is a major issue in reactor safety. Since the long-term coolability of such particle beds is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the pressure field inside the debris has a strong effect on the cooling potential in multi-dimensional cases as expected in severe accidents in light water reactors (LWR). Therefore, the determination of the pressure field for two-phase flows in porous media is one central point of interest.In this context simulation models and in particular dryout models were developed for reactor safety analyses which have to be validated by reliable experimental data. Therefore, basic experimental investigations have been carried out with inductively heated steel balls of 6 or 3 mm diameter to provide a database for the validation and modification of the friction laws included in these dryout models.The performed boiling and dryout experiments show clearly that models without the explicit consideration of the interfacial drag cannot predict the pressure distribution inside a boiling particle bed, not even qualitatively. Against it, models with an explicit consideration of the interfacial drag can describe the distribution of pressure inside a boiling particle bed.     

Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

A study of bubble behaviors in the volumetrically heated packed bed

The volumetrically heated packed bed has been widely utilized in modern industry, however, no research on the bubble behaviors in forced convection subcooled boiling was studied. To study the bubble behaviors in the volumetrically heated packed bed, here electromagnetic induction heating method was used to heat oxidized carbon steel balls adopted to stack packed bed, while water was utilized as the refrigerant in the experiment. Bubble behaviors were observed by a high speed camera for particle diameter varying from 8 mm to 12 mm, mass flux varying from 29.3 kg m⁻² s⁻¹ to 84.2 kg m⁻² s⁻¹, heat flux varying from 14.5 kW m⁻² to 50 kW m⁻², inlet pressure varying from 0.116 MPa to 0.125 MPa, inlet subcooling varying from 7 k to 9.2 k and porosity = 0.39. Obtained flow visualization images were analyzed. The experimental results indicated that the bubbles were blocked by steel balls and easily attached to the surface of balls, then slipped along the surface of steel balls. There was “regrowth phenomenon” in the packed bed and generated bubbles repeated growth several times in the lifetime. The nucleate boiling was firstly observed in the contact surface. Structures of contact surface had great impacts on the bubble shapes, departure diameter and frequency.     

Experimental results on the coolability of a debris bed with multidimensional cooling effects

Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations ( [Rashid et al., 2008], [Groll et al., 2008] and [0055]). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration.A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).     

Investigation of forced convection heat transfer of supercritical pressure water in a vertically upward internally ribbed tube

In the present paper, the forced convection heat transfer characteristics of water in a vertically upward internally ribbed tube at supercritical pressures were investigated experimentally. The six-head internally ribbed tube is made of SA-213T12 steel with an outer diameter of 31.8 mm and a wall thickness of 6 mm and the mean inside diameter of the tube is measured to be 17.6 mm. The experimental parameters were as follows. The pressure at the inlet of the test section varied from 25.0 to 29.0 MPa, and the mass flux was from 800 to 1200 kg/(m² s), and the inside wall heat flux ranged from 260 to 660 kW/m². According to experimental data, the effects of heat flux and pressure on heat transfer of supercritical pressure water in the vertically upward internally ribbed tube were analyzed, and the characteristics and mechanisms of heat transfer enhancement, and also that of heat transfer deterioration, were also discussed in the so-called large specific heat region. The drastic changes in thermophysical properties near the pseudocritical points, especially the sudden rise in the specific heat of water at supercritical pressures, may result in the occurrence of the heat transfer enhancement, while the covering of the heat transfer surface by fluids lighter and hotter than the bulk fluid makes the heat transfer deteriorated eventually and explains how this lighter fluid layer forms. It was found that the heat transfer characteristics of water at supercritical pressures were greatly different from the single-phase convection heat transfer at subcritical pressures. There are three heat transfer modes of water at supercritical pressures: (1) normal heat transfer, (2) deteriorated heat transfer with low HTC but high wall temperatures in comparison to the normal heat transfer, and (3) enhanced heat transfer with high HTC and low wall temperatures in comparison to the normal heat transfer. It was also found that the heat transfer deterioration at supercritical pressures was similar to the DNB at subcritical pressures.     

Two-phase flow and boiling heat transfer in two vertical narrow annuli

Experimental study associated with two-phase flow and heat transfer during flow boiling in two vertical narrow annuli has been conducted. The parameters examined were: mass flux from 38.8 to 163.1 kg/m² s; heat flux from 4.9 to 50.7 kW/m² for inside tube and from 4.2 to 78.8 kW/m² for outside tube; equilibrium mass quality from 0.02 to 0.88; system pressure from 1.5 to 6.0 MPa. It was found that the boiling heat transfer was strongly influenced by heat flux, while the effect of mass velocity and mass quality were not very significant. This suggested that the boiling heat transfer was mainly via nucleate boiling. The data were used to develop a new correlation for boiling heat transfer in the narrow annuli. In the two-phase flow study, the comparison with the correlation of Chisholm [Chisholm, D., 1967. A theoretical basis for the Lockhart–Martinelli correlation for two-phase flow. Int. J. Heat Mass Transfer 10, 1767–1778] and Mishima and Hibiki [Mishima, K., Hibiki, T., 1996. Some characteristics of air–water two-phase flow in small diameter vertical tubes. Int. J. Multiphase Flow 22, 703–712] indicated that the existing correlations could not predict the two-phase multiplier in the narrow annuli well. Based on the experimental data, a new correlation was developed.     

Film flow measurements for high-pressure diabatic annular flow in tubes with various axial power distributions

Measurements of film flow rates in diabatic annular flow in tubes with various axial power distributions were carried out in the high-pressure two-phase flow loop at the Royal Institute of Technology (KTH), Sweden. The measurements were performed at conditions typical for boiling water reactors, i.e. 7 MPa pressure and total mass flux in a range from 750 to 1750 kg/m²s. Four different axial power distributions were used and the film mass flow was measured at 7 axial locations for each set of boundary conditions. The results show that the outlet peaked distribution gives less film than the inlet peaked one. This result is consistent with well known trends from measurements of dryout power. The measurements also show that the film flow at the onset of dryout is very small at investigated conditions in agreement with earlier studies. Finally it is shown that the present data is well predicted by two selected phenomenological models of annular flow.     

Study on dryout point in vertical narrow annuli

Prediction of dryout point is experimentally investigated with deionized water upflowing through narrow annular channel with 1.0 mm and 1.5 mm gap respectively. The annulus with narrow gap is bilaterally heated by AC current power supply. The experimental conditions covered a range of pressure from 0.8 to 3.5 MPa, mass flux of 26.6 to 68.8 kg?m-2?s-1 and wall heat flux of 5 to 50 kW?m-2. The location of dryout is obtained by observing a sudden rise in surface temperature. Kutateladze cor…     

The influence of subcooling on dryout inception in sodium-saturated fuel particle beds with top cooling and adiabatic bottom

In the frame of a one-dimensional theory for the heat transfer behaviour of a top-cooled sodium-saturated particle bed it is possible to describe the onset of packed and/or channeled dryout as a function of the subcooling of the overlaying sodium plenum. The onset of bed channeling is modeled with the aid of capillary pressure. A critical value for the subcooling is derived for the change from packed to channeled dryout for decreasing subcooling. This model is applied for the interpretation of the results of the D4-inpile particle bed experiment performed at SANDIA Laboratories.     

An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube

Within the range of pressure from 9 to 30 MPa, mass velocity from 600 to 1200 kg/(m² s), and heat flux at inner wall from 200 to 600 kW/m², experiments have been performed to investigate the heat transfer characteristics of steam-water two-phase flow in vertical upward tube. The outer diameter of the tube is 32 mm, and the wall thickness is 3 mm. Based on results, it was found that Dryout is the main mechanism of the heat transfer deterioration in the sub-critical pressure region. Near the critical pressure, when the heat transfer deterioration occurs, the steam quality of water is lower than that in the sub-critical pressure region, so that DNB is the main mechanism in this pressure region. At supercritical pressure, the heat transfer performance in circular channel is improved and enhanced. Heat transfer deterioration phenomenon is observed when the fluid bulk temperature approaches to the pseudo-critical value. Nusselt correlation of the forced-convection heat transfer in supercritical pressure region has been provided, which can be used to predict heat transfer coefficient of the vertical upward flow in tube.