Post Dryout Research Articles

Experimental Investigation of Post-Dryout Heat Transfer with R-134a at High Pressures

An experimental study of post-dryout (PDO) heat transfer in the coolant R-134a was performed in a vertical round tube with upward flow. Experiments were conducted at high pressures from 28.4 bars up to 39.8 bars, corresponding to a reduced pressure of 0.7 to 0.98, respectively. Mass flux was varied in the interval of 300 to 1500 kg/m2∙s, and overall equilibrium vapor quality was between −1.88 and 4.89. Depending on the settings of the experimental parameters, the heat flux was varied from around 11 to 100 kW/m2. The uniformly heated tube had an inside diameter of 10 mm and a heated length of 3000 mm. In total, more than 10 000 PDO data points were obtained. In the PDO region, the wall temperature distributions had similar behavior across the pressure range. At the occurrence of dryout, the wall temperature suddenly increased until it reached a peak. For higher mass flux, the wall temperature decreased after reaching the peak, followed by a second temperature increase with a lower slope. For lower mass flux, the wall temperature kept increasing after the dryout point. The temperature peak after the dryout point was smaller at higher pressure, while this effect was even stronger near the critical pressure. Likewise, the vapor quality corresponding to the first peak shifted to even lower values with increasing pressure. Furthermore, it was found that at increasing pressure and at increasing mass flux, the dryout location and the total temperature distribution shifted toward lower vapor qualities. In addition, several PDO correlations were assessed, and a new correlation for high-pressure conditions was developed and compared with the results of the existing ones.

Flow boiling heat transfer of R245fa under hypergravity realized by rotating machine

To measure the influence of hypergravity on flow boiling heat transfer, an experiment with refrigerant R245fa flowing inside a horizontal 1.62 mm tube was conducted, and a turntable was used to simulate the hypergravity environment. Four experimental conditions consisted of the combinations of two mass fluxes (351, 705 kg/m2 s) and two heat fluxes (32, 64 kW/m2) were tested under nominal gravity levels from 1 g to 2.77 g and a pressure of 217.8 kPa. The heat transfer enhancement was observed in both sing-phase flow and two-phase flow when the turntable started to turn, which reveals the particularity of the hypergravity environment realized by rotating machine and the contributions of the Coriolis force due to rotation. As the hypergravity increased further, the heat transfer phenomena were different for different vapor qualities when performed in low mass flux and low heat flux conditions. At low vapor qualities, it was hard to judge whether the heat transfer was enhanced or deteriorated, but it was slightly deteriorated at high vapor qualities before dry-out and enhanced in the region of post dry-out. In addition, the gravitational effect diminished when increasing heat flux or mass flux, while the heat transfer enhancement caused by the tube rotation did not diminish. Besides, the data analysis was also performed to confirm the source of the heat transfer enhancement when the turntable started to turn.

A new correlation for post-dryout heat transfer in upward vertical flow

The present study established a new experimental databank of stable post-dryout (PDO) heat transfer which contains 31,603 data points in total. It is demonstrated by the assessment work with the new databank that previous PDO heat transfer prediction methods could not provide high prediction accuracy to various working conditions. Therefore, this study aims to develop a universal correlation for stable PDO heat transfer. With effect of convective and radiation heat transfer considered, a new prediction model for stable PDO heat transfer is developed, which is a dimensionless one and independent to upstream information. By identifying significant dimensionless parameters with Spearman’s rank correlation coefficient method and the group-wise statistic method, three dimensionless parameters (ReTP, PrW, and xe) are selected to establish the new PDO correlation. Though developed from water experiments for vertical flow in uniformly heated round tubes, the new correlation shows good predictive capability when validated against water experiments with non-uniform heating and R134a experiments with uniform heating, which means the new correlation can be applied to various working conditions. However, if applied to cryogenic pipeline chill-down transient or nuclear power plant accident scenarios, validation with transient experiments needs to be completed in the future.

Experimental investigation and modelling of entrainment rate during bottom reflooding in a narrow rectangular channel

Entrainment rate during bottom reflooding in a narrow rectangular channel is extremely important for heat transfer and reactor security when large break loss of coolant accident (LBLOCA) happens. In order to investigate the entrainment rate under different conditions, an experimental apparatus was established to simulate the process of bottom reflooding under different initial and boundary conditions, the volume of entrainment liquid was obtained by mass conservation according to the pressure drop of entrained water tank and the experimental results revealed that the entrainment rate increased with increasing inlet velocities, coolant temperature and the heating power of the experimental section, but decreased with the increase of system pressure.Furthermore, a new entrainment rate model was developed based on the experimental results combined with the liquid sheet breakup mechanism which was verified to occur in the post-CHF region. Experimental results were used to correct the model with the dimensionless number which was relative to the vaporization of the entrained liquid in the post dryout region. Comparison of the new model with experimental data in open literature and existing model was performed to validate the accuracy of the new developed model, which suggested that the new model could accurately predict the entrainment rate of bottom reflooding in a rectangular channel with the average relative error of 13.1%.

Configuration of Evaporators for Once-Through Heat Recovery Steam Generators of Combined-Cycle Units

Various designs are examined of evaporators for horizontal or vertical once-through heat recovery steam generators (HRSG) of combined-cycle units. Special attention is given to the Benson boilers manufactured by Siemens, which are most widely used now. Vertical HRSG with evaporators having horizontal heating surface sections are also analyzed. Their advantages and disadvantages are outlined. It is demonstrated that HRSGs with vertical heating coils provided with equalizing headers are more advantageous. The equalizing headers are connected with stubs to bottom U-bends of the coils. All heating surfaces are drained by gravity. The temperature conditions of this evaporator were investigated in a GTU load range of 25–100%. It has been found that partial load operation results in poorer heat transfer due to the boiling crisis. However, the maximum temperature jump in the post dryout region will be approximately 25–30°С, which is quite acceptable for reliable operation of low-alloy steel tubes. Assessment of the flow stability at the precrisis conditions demonstrates that the two-phase flow in the evaporator is stable over the applicable load range of HRSG. The calculations were performed for the evaporator without an orifice at the coil inlets. On the whole, the thermal-hydraulic operating conditions of these boilers are much better than those of Benson boilers. From the construction standpoint, a coil evaporator with equalizing headers is simpler and less metal intensive. A quite high liquid velocity is kept at all loads. Dynamic properties are also advantageous, since the material and the geometry of metal intensive elements of the evaporator, such as separators and water tanks, have been selected to improve performance of the overall HRSG.

Mechanistic prediction of post dryout heat transfer and rewetting

Abstract Post dryout heat transfer and rewetting are important processes determining the level and the time duration of high temperature phase and the integrity of the fuel cladding. In spite of extensive studies in the past decades, reliable prediction methods are still missing due to the complexity of processes involved, which consist mainly of interaction between solid wall, main gas flow and droplets. In the present study, a phenomenological model is proposed considering the three individual heat transfer processes between the three parts. Main new features of the present model compared to the models available in the open literature are the mechanistic modeling of (a) droplet concentration and droplet size, (b) turbulent fluctuation velocity of droplets and its critical value, (c) evaporation rate of droplets arriving the wall. Comparison of the new model with selected experimental data shows at least qualitatively good agreement. The experimental behavior of wall temperature can be well explained. According to the new model the Leidenfrost effect results in the hysteresis behavior of wall temperature.

A simplified energy dissipation based model of heat transfer for post-dryout flow boiling

A model for post dryout mist flow heat transfer is presented based on considerations of energy dissipation in the flow. The model is an extension of authors own model developed earlier for saturated and subcooled flow boiling. In the former version of the model the heat transfer coefficient for the liquid single-phase convection as a reference was used, due to the lack of the appropriate model for heat transfer coefficient for the mist flow boiling. That issue was a fundamental weakness of the former approach. The purpose of present investigation is to fulfil this drawback. Now the reference heat transfer coefficient for the saturated flow boiling is that corresponding to vapour flow the end of the mist flow. The wall heat flux is based on partitioning and constitutes of two principal components, namely the convective heat flux for vapour flowing close to the wall and two phase flow droplet–vapour in the core flowing. Both terms are accordingly modelled. The results of calculations have been compared with some experimental correlations from literature showing a good consistency.

Comparative experimental procedures for measuring the local heat transfer coefficient during flow boiling in a microchannel

Most of the microchannel flow boiling experiments are performed using electrical heaters to promote the boiling process. However, there are many heat exchangers applications (e.g. automotive, refrigeration, air-conditioning) in which the vaporization heat is provided by a secondary fluid (water, air…). The paper presents a comparative analysis between two experimental techniques that can be adopted when the heat flux is not imposed by Joule effect heating and hence its direct measurement from electrical parameters is not possible. The two methods are applied to the measurement of the local heat transfer coefficient during flow boiling inside a 0.96mm diameter microchannel. The first method is the one proposed by Del Col et al. (2013), the second one is presented here for the first time. A key advantage of the new technique is the fact that the heat transfer coefficients can be measured in the dryout zone as well. The two methods are used to reduce new experimental data of flow boiling of R1234ze(E), which is a fluorinated propene isomer and a possible low-GWP refrigerant to replace R134a in some air conditioning and refrigeration applications. The flow boiling heat transfer coefficient strongly depends on the heat flux, with little effect of vapor quality and mass velocity. Only the adoption of the new technique allows to investigate the post dryout region as well.

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Prediction of dryout and post-dryout heat transfer using a two-phase CFD model

This study focuses on development of an integrated CFD model for diabatic high quality two-phase flow including tarns-dryout regions from annular-mist regime to mist regime. One unified three-field CFD model accounting for droplets, gas, and liquid film was developed to simulate both pre and post dryout regions, with local models to determine the dryout occurrence. The thin liquid film model was coupled to the gas core flow model, which is described using the Eulerian–Eulerian approach. For the post-dryout region, the various heat and mass transfer mechanisms between the wall, the gas phase, and the droplets were identified, including the wall-gas convective heat transfer, the droplet evaporation, the droplet-wall direct contact heat transfer and the thermal radiation, to calculate the temperature of the wall and the fluid. Of the most interests, dryout location and wall temperature measurements from a post-dryout heat transfer experiment have been used for the validation. Simulation results show that the dramatic temperature excursion could be well captured using current models. Nevertheless, more work will be continued to improve the accuracy of the results.

Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

An open calculation of RD-14M small-break LOCA experiments using CATHENA code

The open calculations of two Small Break Loss of Coolant Accident (SBLOCA) experiments in the RD-14M integral test facility are performed as an International Atomic Energy Agency (IAEA) activity of International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate the thermal–hydraulic analysis code against qualified data for Heavy Water Reactor (HWR) systems. This ICSP started with the first meeting of the participants in Vienna in 2007 November. Two tests were selected for this activity, test B9006, a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection, performed in 1990, and test B9802, a 3-mm inlet header break experiment, performed in 1998, to provide data with full channel power to study boiling in the channels and condensation in the steam generators. The previously performed blind simulations demonstrated that the CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. However, several significant discrepancies between the code predictions and the measurement data were noted, and attributed to the input errors and the code models relevant to post-dryout (PDO) heat transfer phenomena in the heated channels of test B9802. Therefore, the open calculations are performed to allow correction of previous blind calculations in the present study. These open calculations include the Senaratne and Leung model with the optimized coefficients of its correlation for the test B9802, which can predict more accurately the onset of sustained dryout on the surface of the heater rods. The improvement of the open calculation results are shown by comparing them with the test data and also with the blind calculation results.

Prediction of dryout and post-dryout wall temperatures using film thickness model

In this paper, a model for calculating dryout and post-dryout wall temperature for two phase flow in a vertical tube using film thickness model is presented. The model is based on the mass and energy balance equations for vapor core, liquid film and corresponding closure laws for interface transfer processes. It is assumed that the dryout is reached when the local film thickness in the annular flow approaches zero. Thermal-hydraulic processes along the whole length of the boiling channel are simulated, from sub-cooled liquid flow at channel inlet, up to the liquid film dryout, gas entrained droplets, mist flow and post dryout region at tube exit. In the post dryout region, thermal non-equilibrium was assumed between vapor and entrained droplets and wall temperature was obtained iteratively by using a wall heat flux partitioning model. Also, the model is verified against several available experimental data and proved viable for prediction of the wall temperature and the dryout location.

Bubble growth, departure and the following flow pattern evolution during flow boiling in a mini-tube

In the present study, the bubble growth, departure and the following flow pattern evolution during flow boiling in the mini-tube were visualized and quantitatively investigated, along with the simultaneous measurement of the local heat transfer coefficient around a specified nucleation site. Liquid nitrogen was employed as the working fluid and the test section was a segment of vertically upward quartz glass tube with the inner diameter range of 1.3–1.5 mm, which was coated by a layer of transparent ITO film as the heater on the outer surface. The growth rates of bubbles had similar and constant growth rate in two periods of time, i.e., before and after the bubbles departing from the nucleation site, which indicated the bubble growth was primarily governed by the inertial force. The bubble departure diameter and bubble period were investigated and the corresponding correlation was obtained based on the experimental data, which showed that the tube size of the mini-tube had no notable effect on the bubble departure and the trend of the bubble departure was similar to that in macro-tubes. Whereas the following flow pattern evolution was apparently confined due to the size effect, which presented desirable heat transfer performance in mini-tubes. The heat transfer coefficients for different flow patterns along the mini-tube were obtained in terms of bubbly, slug, annular flow and the flow regimes of flow reversal and post dryout. It was found that the dominant heat transfer mechanism was the liquid film evaporation which offered desirable heat transfer capability. The heat transfer performance would be deteriorated in the post dryout regime, while flow reversal could somewhat enhance the heat transfer upstream of the nucleation site. Boiling curves around the specified nucleation site were recorded and analyzed based on the recorded flow patterns.

Prediction of amount of entrained droplets in vertical annular two-phase flow

Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air–water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2 mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam–water annular flow conditions. In each experiment, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using the liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant [Sawant, P.H., Ishii, M., Mori, M., 2008. Droplet entrainment correlation in vertical upward co-current annular two-phase flow. Nucl. Eng. Des. 238 (6), 1342–1352] for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air–water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air–water, helium–water and air–genklene experimental data measured by Willetts [Willetts, I.P., 1987. Non-aqueous annular two-phase flow. D.Phil. Thesis, University of Oxford]. However, comparison of the correlations with the steam–water data available in literature showed significant discrepancies. It is proposed that these discrepancies might have been caused due to the inadequacy of the liquid film extraction method used to measure the entrainment fraction or due to the change in mechanism of entrainment under high liquid flow conditions.

Analysis of the steam generator for the safe integral reactor concept: I. Steady state

A steady state model for the Safe Integral Reactor (SIR) once- through steam generator is developed. The aim is to predict the thermal performance. The model takes into account the special features and conditions of the steam generator. The appropriate heat transfer correlations are utilized in each heat transfer regime especially in the post dryout region where thermodynamic nonequilibrium is expected. Finally, results are compared to other work done independently.

Post dryout heat transfer prediction method

A method for post dryout mist flow forced convection heat transfer is presented and recommended. Comparison is made with other methods and with data for round tubes and rod bundles.