Computational Simulation of Turbulent Natural Convection in a Volumetrically Heated Hemispherical Cavity

2014 ◽  
Author(s):  
Camila Braga Vieira ◽  
Bojan Ničeno ◽  
Jian Su
Keyword(s):  
Experimental Data ◽  
Natural Convection ◽  
Nuclear Power ◽  
Computational Simulation ◽  
Numerical Data ◽  
Local Heat Transfer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Turbulent Heat ◽  
Turbulent Natural Convection

This work presents an analysis of the turbulent natural convection in a volumetrically hemisphere cavity, representing a lower plenum of a Light Water Reactor (LWR) vessel. Considering isothermal top and bottom walls at temperature of 298.5 K and a fluid with Prandtl number (Pr) equal to 8.52, the experiment performed by Asfia et al. [1] was reproduced, so that their experimental data could be used for the validation of the numerical data provided by the present work. Ranging the internal Rayleigh numbers (Rai) from 108 to 3.03 × 1013, numerical simulations were performed using the open-source Computational Fluid Dynamics (CFD) code OpenFOAM (Open Field Operation and Manipulation), making use of the “code-friendly” modification of the turbulence model V2-f, a Reynolds Averaged Navier-Stokes (RANS) based equation model. The calculation of the turbulent heat fluxes was done by the Simple Gradient Diffusion Hypothesis (SGDH), which proved to be adequate for the case considered. Bearing in mind the complexity of the phenomena inside a molten core after a severe accident in a nuclear power plant, some assumptions were done such as the fluid was considered Newtonian, with no-phase change and homogeneous. Numerical correlations of Nusselt number (Nu¯) over the bottom and top walls as function of the Rai were obtained and a good agreement with the experimental data provided by Asfia et al. [1] was reached for the local heat transfer with respect to the angle for the cooled top wall.

Can Natural Convection on Smooth Vertical Plates in the Laminar Regime Be Improved by Adding Forward Facing Triangular Roughness Elements?

2019 ◽  
Author(s):  
Jakob Hærvig ◽  
Anna Lyhne Jensen ◽  
Henrik Sørensen
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Aspect Ratio ◽  
Dead Zone ◽  
Local Heat Transfer ◽  
Local Heat ◽  
Heat Fluxes ◽  
Smooth Surfaces ◽  
Roughness Elements ◽  
Triangular Elements

Abstract Vertical smooth surfaces are commonly used for transferring heat by natural convection. Many studies have tried altering smooth surfaces in various ways to increase heat transfer. Many of these studies fail to increase global heat transfer. The problem commonly reported is dead zones appearing just upstream and downstream obstructions that effectively decrease wall temperature gradients normal to the surface. In this study, we simulate how changes geometry of forward facing triangular roughness elements affect local and global heat transfer for isothermal plates. We change the aspect ratio of the triangular elements from L/h = 5 to L/h = 40 at Grashof numbers of GrL = 8.0 · 104 and GrL = 6.4 · 105. In all cases the flow remains laminar. Even when accounting for the increase in surface area, we keep observing a decrease in global heat transfer compared to the smooth vertical plate. However, the results show by carefully selecting the aspect ratio and pitch distance of the triangular elements based on the Grashof number, the dead zone behind the horizontal part can be eliminated thereby significantly increasing local heat transfer. This observation could help to improve cooling of electronics with high localised heat fluxes.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Effects of Various Parameters on Supercritical-Pressure Heat Transfer and Limits for Heat Transfer Correlations Obtained in Vertical Bare Tubes

2017 ◽  
Author(s):  
Hakim Maloufi ◽  
Hanqing Xie ◽  
Andrew Zopf ◽  
William Anderson ◽  
Christian Langevin ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Heat Fluxes ◽  
Operating Conditions ◽  
Transfer Coefficients ◽  
Wide Range ◽  
Heat Transfer Correlations ◽  
Bare Tube

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine. In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs. The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes. The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Computational Simulation of Natural Convection of a Molten Core in Lower Head of a PWR Pressure Vessel

2010 ◽  
Author(s):  
Camila B. Vieira ◽  
Jian Su
Keyword(s):  
Natural Convection ◽  
Rayleigh Number ◽  
Excellent Agreement ◽  
Pressure Vessel ◽  
Flat Surface ◽  
Computational Simulation ◽  
Internal Heat ◽  
Severe Accident ◽  
Bottom Surface ◽  
Lower Head

The paper reports computational simulation of natural convection of a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel. Numerical simulations were carried out for a two-dimensional semi-circular slice geometry which represents the lower head of a PWR pressure vessel. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate is simulated by using a commercial computational fluid dynamics (CFD) software ANSYS CFX 12.0. The bottom circular surface and top flat surface are both kept at a same constant temperature. The Boussinesq model is used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (shear stress transport) turbulent model is used to model the turbulent eddy viscosity. Two Prandtl numbers are considered, 6.13 and 7.0. Five Rayleigh numbers were simulated for each Prandtl number: 109, 1010, 1011, 1012, and 1013. The average Nusselt numbers on the bottom surface calculated by the computational simulations are in excellent agreement with Mayinger et al. (1976) correlation, with the average Nusselt number on the top flat surface calculated by the computational simulation is in excellent agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations only at Ra = 109. Numerical results of the velocity and temperature fields show the α-phenomena at lower Rayleigh number and ν-phenomena at lower Rayleigh number.

Effect of Surface Emissivity on Conjugate Turbulent Natural Convection in an Air-Filled Cavity with a Heat Source

2016 ◽  
Vol 685 ◽  
pp. 315-319 ◽  
Author(s):  
Igor V. Miroshnichenko ◽  
Mikhail A. Sheremet
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Source ◽  
Surface Radiation ◽  
Turbulent Heat Transfer ◽  
Turbulent Heat ◽  
Heat Conducting ◽  
Surface Emissivity ◽  
Turbulent Natural Convection ◽  
Effect Of Surface

The interaction of conjugate turbulent natural convection and surface thermal radiation in an air-filled square enclosure having heat-conducting solid walls of finite thickness and a heat source has been numerically studied. The primary focus was on the influence of surface emissivity on complex heat transfer. The mathematical model has been formulated in dimensionless variables such as stream function, vorticity and temperature using k-ε turbulent model. The effect of surface emissivity on the average total Nusselt number has been defined. The distributions of streamlines and temperature fields, describing characteristics of the analyzed fluid flow and heat transfer have been obtained. The results clearly show an essential effect of surface radiation on unsteady turbulent heat transfer.

The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies

2016 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Nuclear Power ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Light Water Reactors ◽  
Light Water ◽  
Accident Management ◽  
Water Reactors ◽  
The Impact

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

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