Numerical prediction of cooling margins for a fluid with internal heat generation (original) (raw)
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A possible severe accident scenario is a general meltdown and relocation of the reactor core during which molten core material accumulates in the lower plenum of the reactor vessel. The decay heat generated in a radioactive material would have to be removed through the walls of the lower plenum in order to ensure the integrity of the reactor pressure vessel. Numerical simulations of turbulent natural convection in a geometry representing the lower plenum cavity of a reactor pressure vessel were conducted. A two-dimensional numerical code based on a finite-volume method was developed to simulate turbulent natural convection in a fluid with internal heat generation using large-eddy simulation. Simulations were performed at Rayleigh numbers 1e10 and 2e11 and Prandtl numbers 1.2, 7 and 8, which corresponds to conditions in the numerical investigations made by Nourgaliev et al. (1997) and in the experimental work done by Asfia and Dhir (1996). The results are shown to be in satisfactory agreement.
Nuclear Engineering and Technology, 2015
Background: A numerical simulation was carried out to investigate the difference between internal and external heat-flux distributions at the reactor vessel wall under in-vessel retention through external reactor vessel cooling (IVR-ERVC). Methods: Total loss of feed water, station blackout, and large break loss of coolant accidents were selected as the severe accident scenarios, and a transient analysis using the elementbirth-and-death technique was conducted to reflect the vessel erosion (vessel wall thickness change) effect. Results: It was found that the maximum heat flux at the focusing region was decreased at least 10% when considering the two-dimensional heat conduction at the reactor vessel wall. Conclusion: The results show that a higher thermal margin for the IVR-ERVC strategy can be achieved in the focusing region. In addition, sensitivity studies revealed that the heat flux and reactor vessel thickness are dominantly affected by the molten corium pool formation according to the accident scenario.
Modeling of natural convection phenomena in nuclear reactor core melt
A failure of reactor core cooling and major safety systems may cause melting of nuclear fuel and reactor vessel equipment. In the reactor vessel, the melt flows down into the lower plenum, where it is accumulated. In the past, the common opinion was that the melt would break through the reactor vessel and start to desintegrate the reactor concrete base. However, recent investigations revealed that the core melt can be safely retained in the reactor vessel lower plenum if it is properly cooled. The processes in the reactor core melt in the lower plenum are specific and not yet fully understood. As revealed by a comprehensive overview of the lower plenum cooling problem, natural convection is the most important phenomenon that controls heat transfer from the melt. In the case of natural convection, fluid motion is caused by volumetric forces and density gradients. If these are strong enough, thermal instabilities may result in hydrodynamic instabilities. It was discovered that transition from laminar to turbulent flow occurs at the value of Rayleigh number Ra=5e5 in the case of Rayleigh-Bénard convection and at Ra=1e6 in the case of fluid with internal heat generation. The main problem of turbulent phenomena modeling is the size of turbulent fluid flow structures, which are in general too small to be described accurately using a discrete numerical mesh. The base of the Smagorinsky model is the assumption that the smallest flow structures, which are separated and modeled as a subgrid term, are isotropic and homogeneous. Therefore, viscous dissipation is equal to the production of turbulent kinetic energy. As the Smagorinsky model is too dissipative in the vicinity of the walls, turbulent viscosity wall functions have to be implemented. Natural convection in the melt of nuclear reactor core was modelled as natural convection in a fluid with internal heat generation in a rectangular cavity. The value of Rayleigh number was Ra=1e10 and the value of Prandtl number was Pr=1.2. Numerical simulations were restricted to two-dimensional space, due to computer hardware limitations. The finite volume method was used for spatial discretisation and a combination of Adam-Bashford method and projection scheme was used for time integration. As the calculation of heat transfer in the form of dimensionless Nusselt number revealed, the most severe thermal loads occur on the side walls in the vicinity of the cavity upper boundary. Calculated values of heat transfer can be safely extrapolated to higher values of Rayleigh number.
Dynamics of heat transfer in the melt pool at nuclear severe accident conditions
Prediction of thermal loads on nuclear reactor vessel lower plenum after core melting and relocation during a severe accident requires knowledge about the core melt behavior, especially the circulation pattern. To analyze the heat transfer dynamics on the lower plenum walls, two-dimensional numerical simulations of a fluid flow with internal heat generation were performed for Rayleigh numbers 10^6, 10^7, 10^8, 10^9, 10^11 and 10^13 at Prandtl number 0.8. For subgrid motion modeling, a Large-Eddy Simulation Smagorinsky model was implemented. The minimum, time-average and maximum Nusselt numbers on the boundaries were calculated. The dynamics of fluid structures were analyzed to reveal the instability mechanisms and transition to turbulence. Results disclose Rayleigh-Taylor instabilities as a dominant mechanism for turbulence appearance, which occurs when the Rayleigh number is increased over 10^8. The structure dependence of fluid motion at high Rayleigh numbers makes the time-average of heat transfer hard to assess. The time-average values should be supplemented with probability distributions of related variables.
Preliminary Study of Turbulent Flow in the Lower Plenum of a Gas-Cooled Reactor
A preliminary study of the turbulent flow in a scaled model of a portion of the lower plenum of a gas-cooled advanced reactor concept has been conducted. The reactor is configured such that hot gases at various temperatures exit the coolant channels in the reactor core, where they empty into a lower plenum and mix together with a crossflow past vertical cylindrical support columns, then exit through an outlet duct. An accurate assessment of the flow behavior will be necessary prior to final design to ensure that material structural limits are not exceeded. In this work, an idealized model was created to mimic a region of the lower plenum for a simplified set of conditions that enabled the flow to be treated as an isothermal, incompressible fluid with constant properties. This is a first step towards assessing complex thermal fluid phenomena in advanced reactor designs. Once such flows can be computed with confidence, heated flows will be examined. Experimental data was obtained using three-dimensional Particle Image Velocimetry (PIV) to obtain non-intrusive flow measurements for an unheated geometry. Computational fluid dynamic (CFD) predictions of the flow were made using a commercial CFD code and compared to the experimental data. The work presented here is intended to be scoping in nature, since the purpose of this work is to identify improvements that can be made to subsequent computations and experiments. Rigorous validation of computational predictions will eventually be necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.
Annals of Nuclear Energy, 2014
The design of passive heat removal systems is one of the main characteristics of the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is a key heat removal system during normal and off normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive the postulated accidents. The commercial Computational Fluid Dynamics (CFD) STAR-CCM+/V5.02.009 code was used for three-dimensional system modeling and analysis of the RCCS. Different RCCS geometries and configurations were investigated to analyze heat exchange in the VHTR cavity. Sensitivity analyses over the RCCS cavity height and cooling panel location with respect to the reactor pressure vessel (RPV) wall were performed. The objective of the present work was to use CFD tools for addressing the behavior of the RCCS following accident conditions.
Computational Fluid Dynamics Applied to Study Coolant Loss Regimes in Very High Temperature Reactors
Brazilian Journal of Radiation Sciences
The nuclear energy is a good alternative to meet the continuous increase in world energy demand. In this pers-pective, VHTRs (Very High Temperature Reactors) are serious candidates for energy generation due to its inherently safe performance, low power density and high conversion efficiency. However, the viability of these reactors depends on an efficient safety system in the operation of nuclear plants. The HTR (High Temperature Reactor)-10 model, an experimental reactor of the pebble bed type, is used as a case study in this work to perform the thermohydraulic simulation. Due to the complex patterns flow that appear in the pebble bed reactor core CFD (Computational Fluid Dynamics) techniques are used to simulate these reactors. A realistic approach is adopted to simulate the central annular column of the reactor core. As geometrical model of the fuel elements was selected the BCC (Body Centered Cubic) arrangement. Parameters considered for reactor design are available in the tec...
Nuclear Engineering and Design, 2022
This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the coexistence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.