Flow Control in a Mixing-Vane Grid to Enhance Thermal Hydraulic Performance

2010 ◽  
Author(s):  
Tsutomu Ikeno ◽  
Tatsuya Sasakawa ◽  
Shumpei Kakinoki ◽  
Momonori Murase
Keyword(s):  
Heat Flux ◽  
Flow Control ◽  
Pressure Loss ◽  
Axial Direction ◽  
Nucleate Boiling ◽  
Swirl Flow ◽  
Projection Area ◽  
Main Stream ◽  
Two Phase ◽  
Rod Bundle

To investigate the effect of mixing-vane shape, heat flux at departure from nucleate boiling (DNB) and pressure loss were measured. Computational fluid dynamics (CFD) was utilized to discuss the flow control. The pressure loss and the DNB tests were performed in a water and a Freon loops, respectively. Two mixing-vanes were designed to have same projection area but different inclination. The rod-bundle was 5 by 5 and 17 by 17 respectively at the water and Freon tests. The experimental results showed that the slightly inclined mixing-vane produced the same DNB heat flux as the deeply inclined mixing-vane and did smaller pressure loss than it. Pressure loss of the two mixing-vane grids was different in spite of the same projection area. The result of CFD showed a swirl flow decaying along the main stream in the axial direction. The swirl was stronger in the deeply inclined mixing-vane, however it decayed faster whereas one maintained long in the slightly inclined mixing-vane. This result suggested that the deep inclination caused a steep change in axial momentum to induce strong turbulence diffusion. This flow structure did not change the DNB heat flux because the two-phase discontinuity dominated the phenomena. This study provided a successful example of flow control in a mixing-vane grid.

A Nondimensional Analysis of Boiling Dry a Vertical Channel with a Uniform Heat Flux

1981 ◽  
Vol 103 (4) ◽  
pp. 667-672 ◽  
Author(s):  
K. H. Sun ◽  
R. B. Duffey ◽  
C. Lin
Keyword(s):  
Heat Flux ◽  
Nuclear Power ◽  
Closed Form Solution ◽  
Vertical Channel ◽  
Form Solution ◽  
Uniform Heat Flux ◽  
Two Phase ◽  
Drift Flux ◽  
Rod Bundle ◽  
Uniform Heat

A thermal-hydraulic model has been developed for describing the phenomenon of hydrodynamically-controlled dryout, or the boil-off phenomenon, in a vertical channel with a spatially-averaged or uniform heat flux. The use of the drift flux correlation for the void fraction profile, along with mass and energy balances for the system, leads to a dimensionless closed-form solution for the predictions of two-phase mixture levels and collapsed liquid levels. The physical significance of the governing dimensionless parameters are discussed. Comparisons with data from single-tube experiments, a 3 × 3 rod bundle experiment, and the Three Mile Island nuclear power plant show good agreement.

scholarly journals Determination of Pool Boiling Critical Heat Flux Enhancement in Nanofluids

2007 ◽  
Author(s):  
Bao H. Truong
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Pool Boiling ◽  
Nucleate Boiling ◽  
Porous Coating ◽  
Transfer Coefficients ◽  
Two Phase ◽  
Sem Images ◽  
Heat Transfer Coefficients ◽  
Nucleate Boiling Heat Transfer

Nanofluids are engineered colloids composed of nano-size particles dispersed in common fluids such as water or refrigerants. Using an electrically controlled wire heater, pool boiling Critical Heat Flux (CHF) of Alumina and Silica water-based nanofluids of concentration less than or equal to 0.1 percent by volume were measured. Silica nanofluids showed a CHF enhancement up to 68% and there seems to be a monotonic relationship between the nanoparticle concentration and the magnitude of enhancement. Alumina nanofluids had a CHF enhancement up to 56% but the peak occurred at the intermediate concentration. The boiling curves in nanofluid were found to shift to the left of that of water and correspond to higher nucleate boiling heat transfer coefficients in the two-phase flow regime. Scanning Electron Microscopy (SEM) images show a porous coating layer of nanoparticles on wires subjected to nanofluid CHF tests. These coating layers change the morphology of the heater’s surface, and are responsible for the CHF enhancement. The thickness of the coating was estimated using SEM and was found ranging from 3.0 to 6.0 micrometers for Alumina, and 3.0 to 15.0 micrometers for Silica.

Numerical Prediction of Void Distribution in Two-Phase Turbulent Flow in a Sub-Channel

2009 ◽  
Author(s):  
Tsutomu Ikeno ◽  
Tatsuya Sasakawa ◽  
Isao Kataoka
Keyword(s):  
Turbulent Flow ◽  
Transport Equation ◽  
Nucleate Boiling ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Simulation Code ◽  
Pressurized Water ◽  
Void Fractions ◽  
Rod Bundle ◽  
Void Distribution

Numerical simulation code for predicting void distribution in two-phase turbulent flow in a sub-channel was developed. The purpose is to obtain a profile of void distribution in the sub-channel. The result will be used for predicting a heat flux at departure from nucleate boiling (DNB) in a rod bundle for the pressurized water reactor (PWR). The fundamental equations were represented by a generalized transport equation, and the transport equation was transformed onto the generalized coordinate system fitted to the rod surface and the symmetric lines in the sub-channel. Using the finite-volume method the transport equation was discretized for the SIMPLE algorism. The flow field and void fraction at the steady state were calculated for different average void fractions. The computational result for atmospheric pressure condition was successfully compared with experimental data. Sensitivity analysis for the PWR condition was performed, and the result showed that the secondary flow slightly contributed to homogenizing the void distribution.

Two-Phase Pipe Quenching Correlations for Liquid Nitrogen and Liquid Hydrogen

2019 ◽  
Vol 141 (4) ◽  
Author(s):  
S. R. Darr ◽  
J. W. Hartwig ◽  
J. Dong ◽  
H. Wang ◽  
A. K. Majumdar ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Liquid Nitrogen ◽  
Critical Heat Flux ◽  
Flow Boiling ◽  
Nucleate Boiling ◽  
Liquid Hydrogen ◽  
Two Phase ◽  
Cryogenic Flow ◽  
Two Phase Heat Transfer

Recently, two-phase cryogenic flow boiling data in liquid nitrogen (LN2) and liquid hydrogen (LH2) were compared to the most popular two-phase correlations, as well as correlations used in two of the most widely used commercially available thermal/fluid design codes in Hartwig et al. (2016, “Assessment of Existing Two Phase Heat Transfer Coefficient and Critical Heat Flux on Cryogenic Flow Boiling Quenching Experiments,” Int. J. Heat Mass Transfer, 93, pp. 441–463). Results uncovered that the correlations performed poorly, with predictions significantly higher than the data. Disparity is primarily due to the fact that most two-phase correlations are based on room temperature fluids, and for the heating configuration, not the quenching configuration. The penalty for such poor predictive tools is higher margin, safety factor, and cost. Before control algorithms for cryogenic transfer systems can be implemented, it is first required to develop a set of low-error, fundamental two-phase heat transfer correlations that match available cryogenic data. This paper presents the background for developing a new set of quenching/chilldown correlations for cryogenic pipe flow on thin, shorter lines, including the results of an exhaustive literature review of 61 sources. New correlations are presented which are based on the consolidated database of 79,915 quenching points for a 1.27 cm diameter line, covering a wide range of inlet subcooling, mass flux, pressure, equilibrium quality, flow direction, and even gravity level. Functional forms are presented for LN2 and LH2 chilldown correlations, including film, transition, and nucleate boiling, critical heat flux, and the Leidenfrost point.

Critical Heat Flux and Onset of Nucleate Boiling Tests in a Finned Rod Bundle

2004 ◽  
Vol 148 (3) ◽  
pp. 287-293
Author(s):  
Hee Taek Chae ◽  
Jong Hark Park ◽  
Heonil Kim ◽  
Soon Heung Chang
Keyword(s):  
Heat Flux ◽  
Critical Heat Flux ◽  
Nucleate Boiling ◽  
Rod Bundle

scholarly journals ELETRIC FIELD ENHANCEMENT AND DETERIORATION OF BOILING HEAT TRANSFER AND CRITICAL HEAT FLUX IN DIELECTRIC FLUIDS

2001 ◽  
Vol 1 (1) ◽  
pp. 32
Author(s):  
P. M. Carrica ◽  
V. Masson
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Critical Heat Flux ◽  
Electric Fields ◽  
Boiling Heat Transfer ◽  
Nucleate Boiling ◽  
Film Boiling ◽  
Two Phase ◽  
Boiling Regime ◽  
Dielectric Fluids

We present the results of an experimental study of the effects of externally imposed electric fields on boiling heat transfer and critical heat flux (CHF) in dielectric fluids. The study comprises the analysis of geometries that, under the effects of electric fields, cause the bubbles either to be pushed toward the heater or away from it. A local phase detection probe was used to measure the void fraction and the interfacial impact rate near the heater. It was found that the critical heat flux can be either augmented or reduced with the application of an electric field, depending on the direction of . In addition, the heat transfer can be slightly enhanced or degraded depending on the heat flux. The study of the two-phase flow in nucleate boiling, only for the case of favorable dielectrophoretic forces, reveals that the application of an electric field reduces the bubble detection time and increases the detachment frequency. It also shows that the two-phase flow characteristics of the second film boiling regime resemble more a nucleate boiling regime than a film boiling regime.

DNB Prediction Method Using Local Void Fraction Applied to a PWR Rod Bundle With Mixing Vane Grid

2012 ◽  
Author(s):  
Tatsuya Sasakawa ◽  
Tsutomu Ikeno ◽  
Isao Kataoka
Keyword(s):  
Void Fraction ◽  
New Technology ◽  
Prediction Method ◽  
Nucleate Boiling ◽  
Two Phase ◽  
Average Value ◽  
Rod Bundle ◽  
Wide Range ◽  
Channel Analysis ◽  
Reasonable Prediction

A new code coupled between a sub-channel analysis code and a computational multi-fluid dynamics (CMFD) code was applied to a PWR rod bundle with mixing vane grid. The code was developed to predict the departure from nucleate boiling (DNB). This is a new technology of CMFD based on abundant experience and models developed for sub-channel analysis: CMFD computed void distribution to fit the average value calculated by two-phase models in sub-channel analysis code. A new source term represented centripetal motion of small bubbles in the wake behind a rising vapor slug. In order to apply the code to a PWR rod bundle, effects of local mass flux, pressure and mixing vane grid was modeled. The present method was applied to the analysis for DNB tests of simulated PWR fuel assembly with mixing vane grids. The result showed that, using only a critical void fraction, reasonable prediction was achieved in a wide range of flow condition and a variety of flow regimes.

Two-Phase Flow Control of Electronics Cooling With Pseudo-CPUs in Parallel Flow Circuits: Dynamic Modeling and Experimental Evaluation

2013 ◽  
Vol 135 (3) ◽  
Author(s):  
Nicolas Lamaison ◽  
Jackson Braz Marcinichen ◽  
John Richard Thome
Keyword(s):  
Heat Flux ◽  
Steady State ◽  
Flow Control ◽  
Parallel Flow ◽  
Heat Fluxes ◽  
Vapor Quality ◽  
Two Phase ◽  
Cooling Cycle ◽  
Chip Temperature ◽  
Test Loop

On-chip two-phase cooling of parallel pseudo-CPUs integrated into a liquid pumped cooling cycle is modeled and experimentally verified versus a prototype test loop. The system's dynamic operation is studied since the heat dissipated by microprocessors is continuously changing during their operation and critical heat flux (CHF) conditions in the microevaporator must be avoided by flow control of the pump speed during heat load disturbances. The purpose here is to cool down multiple microprocessors in parallel and their auxiliary electronics (memories, dc/dc converters, etc.) to emulate datacenter servers with multiple CPUs. The dynamic simulation code was benchmarked using the test results obtained in an experimental facility consisting of a liquid pumped cooling cycle assembled in a test loop with two parallel microevaporators, which were evaluated under steady-state and transient conditions of balanced and unbalanced heat fluxes on the two pseudochips. The errors in the model's predictions of mean chip temperature and mixed exit vapor quality at steady state remained within ±10%. Transient comparisons showed that the trends and the time constants were satisfactorily respected. A case study considering four microprocessors cooled in parallel flow was then simulated for different levels of heat flux in the microprocessors (40, 30, 20, and 10 W cm−2), which showed the robustness of the predictive-corrective solver used. For a desired mixed vapor exit quality of 30%, at an inlet pressure and subcooling of 1600 kPa and 3 K, the resulting distribution of mass flow rate in the microevaporators was, respectively, 2.6, 2.9, 4.2, and 6.4 kg h−1 (mass fluxes of 47, 53, 76 and 116 kg m−2 s−1) and yielded approximately uniform chip temperatures (maximum variation of 2.6, 2, 1.7, and 0.7 K). The vapor quality and maximum chip temperature remained below the critical limits during both transient and steady-state regimes.

The effect of surfactant concentrations and surface material on heat transfer coefficient in nucleate boiling regime

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 365-374
Author(s):  
A. M. Refaey ◽  
S. Elnaggar ◽  
S. H. Abdel-Latif ◽  
A. Hamza
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Boiling Heat Transfer ◽  
Two Phase Flow ◽  
Surfactant Solution ◽  
Nucleate Boiling ◽  
Phase Flow ◽  
Two Phase

Abstract The nucleate boiling regime and two-phase flow are greater importance to the safety analysis of nuclear reactors. In this study, the boiling heat transfer in nuclear reactor is numerical investigated. The computational fluid dynamics (CFD) code, ANSYS Fluent 17.2 is used and the boiling model is employed. The numerical predictions obtained are compared with the experimental data reported by A. Hamza et al. [9]. An experimental test rig is designed and constructed to investigate the effect of cooling water chemistry control and the material of heater surface. CFD software, allows the detailed analysis of the two-phase flow and heat transfer. In this paper, we evaluate the accuracy of the boiling model implemented in the ANSYS Fluent code. This model is based on the heat flux partitioning approach and accommodates the heat flux due to single-phase convection, quenching and evaporation. The validation carried out of surfactant fluid/vapor two-phase flow inside the 2-D cylindrical boiling vessel. A heated horizontal pipe with stainless steel, Aluminum, and Zircalloy surface materials are used to numerically predict the field temperature and void fraction. Different surfactant concentrations ranging from 0, (pure water) to 1500 ppm, and heat fluxes ranging from 31 to 110 kW/m2 are used. The results of the predicted model depict that the addition of SDS Surfactant and increasing the heat flux improves the coefficient of boiling heat transfer for a given concentration. Also, it was found that the increasing of the concentration of aqueous surfactant solution increases the pool boiling heat transfer coefficient. The aqueous surfactant solution SDS improved the heat transfer coefficient of Aluminum, Zircalloy and stainless steel surface materials by 135%.138% and 120% respectively. The results of the numerical model are nearly in agreement with that measured in experimental.

Global Flow Regime Identification in a Rod Bundle Geometry

2008 ◽  
Author(s):  
Sidharth Paranjape ◽  
Damian Stefanczyk ◽  
Yong Liang ◽  
Takashi Hibiki ◽  
Mamoru Ishii
Keyword(s):  
Flow Regime ◽  
Flow Regimes ◽  
Axial Direction ◽  
Distribution Functions ◽  
Flow Channel ◽  
Cumulative Probability ◽  
Spacer Grid ◽  
Two Phase ◽  
Rod Bundle ◽  
Flow Regime Transition

Flow regime maps were obtained for an adiabatic air-water two phase flow through a flow channel with 8×8 rod bundle, which simulated a typical rod bundle in a BWR. Impedance void meters were used to measure the area averaged void fraction at various axial locations in the flow channel The Cumulative Probability Distribution Functions (CPDF) of the signals from the impedance meters were fed to a self organizing neural network to identify the flow regimes. The flow regimes were identified at seven axial locations in the channel in order to understand the development of the flow regimes in axial direction. The experimental flow regime transition boundaries agreed well with the theoretical ones obtained using Mishima and Ishii (1984) model. In addition, the two impedance void meters located across a spacer grid, which were used to study the change in the flow regime across the spacer grid.

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