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The buoyant cylindrical annuli flow occurs in a number of engineering applications, one of which is the boiler penetration cavities of nuclear power stations. In these cavities, there are pipes that carry a fluid from and into the boiler. The penetration cavities contain a hot gas that penetrates them from the reactor because the cavities run horizontally through the pressure vessel of the reactor. The inner pipes are much colder than the outer cylindrical casing of the cavities as these pipes contain a fluid that is much cooler than the hot gas from the reactor.
The cylindrical annuli configuration examined in this study has one inner cylinder that is maintained at a temperature that is higher than the temperature of the outer cylinder. The buoyancy driven flow in the inner cylinder’s boundary layer forms a plume that rises and impinges on the outer cylinder. The Rayleigh and the Prandtl numbers were set equal to 1.18*〖10〗^9 and 0.688, respectively.
High Rayleigh number flows are expensive to compute using wall-resolved Large eddy simulation (LES) which requires a grid fine enough in the three directions to resolve the near-wall turbulence structures. An alternative to LES is hybrid RANS-LES methods. The dual-mesh hybrid RANS-LES method (see Xiao & Jenny (2012)) was used in the present study. In this method an unsteady RANS and a coarse LES are run simultaneously on two different grids. A criterion is used to determine the locations at which each simulation is expected to perform better than the other. Consequently, at every location the less accurate simulation is corrected towards the more accurate one.
The dual-mesh results were validated against the QDNS data of Addad et al. (2015). The dual-mesh approach was found to provide reasonably accurate predictions of different quantities relevant to both the flow and the thermal fields.
Correcting for Tube Curvature Effects During External Condensation in the Presence of a Noncondensable Gas
Paul Scherrer Institut
Vapor condensation in the presence of noncondensable gases (NCG) attracts continued interest due to its widespread range of applications. To remove steam from the air-rich containment, many advanced nuclear plants rely on Passive Containment Cooling Systems (PCCS) which are tube-and-shell heat exchangers. In studies of NCG effects on condensation over external surfaces, experimentalists and analysts alike usually employ correction factors to transfer tube condensation rates to flat plate equivalents or vice-versa. For laminar free convection over vertical cylinders, these factors are estimated from correlations which are a function of the Grashof number and tube curvature, but not the NCG fraction. We show that this procedure is not rigorous and leads to significant errors, especially at small tube radii. Indeed, the usual way of correcting for curvature is strictly valid only in the limit where the NCG fraction is so large that convection heat transfer dominates condensation. On the other hand, as the NCG fraction becomes minute, the vapor condensation rates should tend to values predicted by the Nusselt theory, and no enhancement due to tube curvature should be expected.
This paper focuses on vapor condensation as the gas mixture moves along a vertical tube in the laminar free convection regime. The gas and liquid film boundary layers are solved in a coupled way. A marching solution procedure using local non-similarity is adopted. The model is first verified for near pure free convection flows, then the predictions are validated against data on flat plate steam condensation in the presence of air. Finally, a non-dimensional correlation is deduced from the numerical results to allow for the correct prediction of curvature effects on steam condensation rates in the presence of air. The proposed correction spans the whole range of NCG concentrations, and displays the expected behavior at the extremes of this range.
CFD Simulation of Sub-Channel Void Fraction Distribution in a 5X5 Mock-Up of PWR Fuel Assembly under Sub-Cooled Flow Boiling
Jun-Yi Zhang, Xiao Yan
Nuclear Power Insititute of China
Subcooled boiling heat transfer, which is related to Departure of Nucleate Boiling(DNB), is highly concerned in fuel assembly of PWR. It is of importance and significance to predict void fraction and mass/energy transfer characteristics among sub-channels under subcooled boiling with 3-D CFD code(STAR-CCM+) to obtain more details for better understanding of two-phase flow process in rod bundle. To better understand and predict void distribution and its transportation characteristics in rod bundle, two-phase flow simulation with two-fluid model was carried out with Heat Partitioning Model. In this study, the 5×5 full length PSBT rod bundle with Uniform-Axial Power Distribution (U-APD) was used under prototype condition (test serial of B5 configuration) to test the physical method. The parameter distribution in axial and radial orientation among different sub-channels at the downstream of the last Mixing Vane Grid(MVG), where the DNB occurs at the end of the heated section, was studied. Different void fraction distribution was observed among central channels, side channels and corner channels. It is induced by different mass transfer process results from the layout of mixing vanes. Moreover, different mass transfer characteristics occurred between corner channels at the different location. Energy transfer is related to the mass transfer among sub-channels. The simple support grid (SSG) shows a suppression of radial transverse flow which is introduced by the mixing vanes for heat transfer enhancement.
Pressure fluctuation of the two-phase flow in a subchannel under Boiling Water Reactor thermal hydraulic conditions
1Hitachi-GE Nuclear Energy, Ltd.; 2Central Research Institute of Electric Power Industry
In Boiling Water Reactors (BWRs), in order to evaluate the behavior of the coolant inside the fuel bundle, the cross-flow phenomenon has been studied so far. In the previous study of the cross-flow phenomenon for two subchannels under room temperature and atmospheric pressure, the correlation between the void fraction and the magnitude of the differential pressure fluctuation was observed irrespective of the geometry of the subchannel and the mass flux. Here, before the study of the cross-flow phenomenon, we have focused on the fundamental phenomenon, namely, the behavior of the two-phase flow in a subchannel. In our study, the test subchannel simulating the structure in a fuel rod bundle was used under the BWR thermal hydraulic conditions. In the test facility, the time-averaged void fraction (TVF) is measured by the three-dimensional X-ray computed tomography system, and the pressure fluctuation is obtained by the piezoelectric pressure transducer for measuring the instantaneous pressure. With this test facility, we obtained the relationship between the TVF and the magnitude of the pressure fluctuation in a subchannel and confirmed that the magnitude of the pressure fluctuation depended on the TVF and the mass flux of the two-phase flow. Also, a similar tendency was observed in the relationship between the flow quality and the magnitude of the pressure fluctuation. This work was partly conducted as a government-commissioned project of Agency for Natural Resources and Energy, “Technology development for common ground maintenance to contribute to nuclear safety enhancement in 2018 (Research and development for understanding two-phase flow behavior inside a fuel bundle)”.