Critical heat flux under zero flow conditions in vertical annulus with uniformly and non-uniformly heated sections

Recent work has demonstrated that as the bulk convective velocity in subcooled nucleate flow boiling increases, the heat transfer tends to become independent of flow orientation with respect to gravity. There is significant interest in developing heat exchangers for next generation spacecraft that operate in the gravity-independent flow boiling regime. In order to develop such heat exchangers it is important to understand the effect of gravity on the critical heat flux and to determine whether a gravity on the critical heat flux and to determine whether a gravity-independent flow boiling critical heat flux regime exists. This work describes subcooled flow boiling experiments where the critical heat flux is measured over a range of flow orientations with respect to gravity: 0°, 45°, 90°, 135°, 180°, 225°, 270°, and 315°. It has been found that at low bulk flow velocities there is a large variation of critical heat flux with different flow orientations. At large convective velocities, the variation of critical heat flux with different flow orientations is significantly diminished. It appears that with further increases in bulk flow velocity, a gravity-independent critical heat flux regime exists, although the current experimental facility was not capable of operating at those flow conditions.

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Critical heat flux under choking flow conditions

Nuclear Engineering and Design ◽

10.1016/s0029-5493(00)00347-2 ◽

2001 ◽

Vol 205 (1-2) ◽

pp. 175-190 ◽

Cited By ~ 2

Author(s):

A. Olekhnovitch ◽

A. Teyssedou ◽

P. Tye

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Flow Conditions

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CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code

10th International Conference on Nuclear Engineering, Volume 3 ◽

10.1115/icone10-22419 ◽

2002 ◽

Cited By ~ 6

Author(s):

Jaakko Miettinen ◽

Holger Schmidt

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Void Fraction ◽

Bubble Diameter ◽

Velocity Profiles ◽

Global Model ◽

Model Parameters ◽

Fluid Viscosity ◽

Flow Conditions ◽

Two Phase

Framatome ANP develops a new boiling water reactor called SWR 1000. For the case of a hypothetical core melt accident it is designed in such a way that the core melt is retained in the Reactor Pressure Vessel (RPV) at low pressure owing to cooling of the RPV exterior and high reliable depressurization devices. Framatome ANP performs — in co-operation with VTT — tests to quantify the safety margins of the exterior cooling concept for the SWR 1000, for determining the limits to avoid the critical heat fluxes (CHFs). The three step procedure has been set up to investigate the phenomenon: 1. Water-air study for a 1:10 scaled global model, with the aim to investigate the global flow conditions. 2. Water-air study for a 1:10 scaled, 10% sector model, with the aim to find a flow sector with almost similar flow conditions as in the global model. 3. Final CHF experiments for a 1:1-scaled, 10% sector., the boarders of this model have been selected based on the first two steps. The instrumentation for the water/air experiments included velocity profiles, the vertically averaged average void fraction and void fraction profiles in selected positions. The experimental results from the air-water experiments have been analyzed at VTT using the Fluent-4.5.2 code with its Eulerian multiphase flow modeling capability. The aim of the calculations was to learn how to model complex two-phase flow conditions. The structural mesh required by Fluent-4 is a strong limitation in the complex geometry, but modeling of the 1/4 sector from the facility was possible, when the GAMBIT pre-processor was used for the mesh generation. The experiments were analyzed with the 150 × 150 × 18 grid for the geometry. In the analysis the fluid viscosity was the main dials for adjusting the vertical liquid velocity profiles and the bubble diameter for adjusting the phase separation. The viscosity ranged between 1 to 10000 times the molecular viscosity, and bubble diameter between 3 to 100 mm, when the calculation results were adjusted for a good agreement with the experimental data. The analysis results were very valuable for designing the final water/steam facility for final CHF tests. The validation against data from the air-water experiments proved that the present CFD codes approach to the state where they can be used for simulating such two-phase experiments, where the fraction of both phases is essential and the flow is strongly affected by the density differences. It is still too early to predict, if the CFD calculation of the 1:1 scale critical heat flux experiments is successful, could the result be used for formulating a new type of a critical heat flux correlation, where the effects of CRD’s on the flow patterns and gap dimensions are model parameters.

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Predictions of Critical Heat Flux in Annular Pipes with TRACE V4.160 Code

Volume 2: Thermal Hydraulics ◽

10.1115/icone14-89253 ◽

2006 ◽

Author(s):

Audrius Jasiulevicius ◽

Rafael Macian-Juan

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Power Distribution ◽

Low Flow ◽

Flow Conditions ◽

Experimental Database ◽

Mass Fluxes ◽

Power Profile ◽

Institute Of Technology ◽

Experimental Values

This paper presents the assessment of TRACE (version v4.160) against the Critical Heat Flux (CHF) experiments in annular tubes performed at the Royal Institute of Technology (KTH) in Stockholm, Sweden. The experimental database includes data for coolant mass fluxes between 250 and 2500 kg/m2s and inlet subcoolings of 10 and 40 K at a pressure of 70 bar. The work presented in this paper supplements the calculations of single round tube experiments carried out earlier and provides a broader scope of validated geometries. In addition to the Biasi and CISE-GE CHF correlations available in the code, a number of experimental points at low flow conditions are available for the annular geometry experiments, which also permitted the assessment of the Biasi/Zuber CHF correlation used in TRACE v4.160 for low flow conditions. Experiments with different axial power distribution were simulated and the effects of the axial power profile and the coolant inlet subcooling on the TRACE predictions were investigated. The results of this work show that the Biasi/Zuber correlation provides good estimation of the CHF at 70 bar, and, for the same conditions, the simulation of the annular experiments resulted in the calculation of lower CHF values compared to single-tube experiments. The analysis of the performance of the standard TRACE CHF correlations shows that the CISE-GE correlation yields critical qualities (quality at CHF) closer to the experimental values at 70 bar than the Biasi correlation for annular flow conditions. Regarding the power profile, the results of the TRACE calculations seem to be very sensitive to its shape, since, depending on the profile, different accuracies in the predictions were noted while other system conditions remained constant. The inlet coolant subcooling was also an important factor in the accuracy of TRACE CHF predictions. Thus, an increase in the inlet subcooling led to a clear improvement in the estimation of the critical quality with both Biasi and CISE-GE correlations. To complement the work, three additional CHF correlations were implemented in TRACE v4.160, namely the Bowring, Tong W-3 and Levitan-Lantsman CHF models, in order to assess the applicability of these correlations to simulate the CHF in annular tubes. The improvement of CHF predictions for low coolant mass flows (up to 1500 kg/m2s) is noted when applying Bowring CHF correlation. However, the increase in the inlet subcooling increases the error in predicted critical quality with the Bowring correlation. The Levitan-Lantsman and Tong-W-3 correlations provide results similar to the Biasi model. Therefore, the most correct CHF predictions among the investigated correlations were obtained using CISE-GE model in the standard TRAC v4.160 code.

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Flow Instability and Critical Heat Flux for Flow Boiling of Water in a Vertical Annulus at Low Pressure

ASME/JSME 2011 8th Thermal Engineering Joint Conference ◽

10.1115/ajtec2011-44292 ◽

2011 ◽

Cited By ~ 4

Author(s):

Christoph Haas ◽

Leonhard Meyer ◽

Thomas Schulenberg

Keyword(s):

Heat Flux ◽

Critical Heat Flux ◽

Mass Flux ◽

Flow Instability ◽

Flow Boiling ◽

Dynamic Contact Angle ◽

Dynamic Contact ◽

Low Pressure ◽

Vertical Annulus ◽

Zircaloy Tube

We investigated the critical heat flux (CHF) for flow boiling of water in a vertical annulus. The coaxial annulus has a diameter ratio of 1.37 and the inner zircaloy tube is heated directly over a length of 325 mm. CHF can occur prematurely due to flow instabilities. Therefore, we analyzed the flow stability at different heat input conditions using two types of pumps, a rotary and a gear type pump. The unstable CHF occurred at 61% and 90% of the stable value for the rotary and the gear type pump, respectively. Consequently, the following CHF experiments were conducted at stable flow conditions. The outlet pressure was constant at 120 kPa, the mass flux varied from 250 to 1000 kg/(m2s) and the inlet subcooling was at 102, 167, and 250 kJ/kg. The CHF results increase with mass flux from 0.67 to 2.62 MW/m2 and show similar trends compared to literature data. However, the experimental data for flow boiling in annuli at low pressure are limited. Additionally, we measured the dynamic contact angle between the zircaloy tube surface and water using the Wilhelmy method.

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