Computational Thermal Fluid Dynamic Analysis of Cooling Systems for Fusion Reactor Components

In a fusion reactor during plasma operation, the heat loads on plasma facing components can be as high as 5 MW/m2 [1], which should be removed by a proper mechanism to prevent the damage of reactor components. In order to handle such high heat fluxes a suitable heat sink with proper thermal hydraulics is required. In the recent past several heat sinks have been proposed; among which the Hypervapotron heat sink, operating in the highly subcooled boiling regime, is considered as one of the potential candidates. In order to accurately predict the performance of the system, a thermal hydraulic analysis is required. This thesis employs a Computational Fluid Dynamic (CFD) approach to do the thermal hydraulic analysis of the subcooled flow boiling inside the Hypervapotron channel. For this purpose four boiling models are tested using two commercial CFD codes. The four boiling models tested are Rensselaer Polytechnic Institute (RPI) boiling model [2] available in ANSYS-FLUENT 13, Bergles-Rohsenow (BR) model [3] implemented as an external User Defined Function (UDF) in the FLUENT code, the Rohsenow boiling model [4] extended with the capability of transition to film boiling for high heat fluxes available in STAR-CCM+ 7.02, and finally Transition boiling model [4] available in STAR-CCM+ 7.02. These models are used to test the thermal performance of Hypervapotron using the experimental data (showing the variation of temperature with heat flux) obtained from the experiments conducted at Efremov Institute Russia and Joint European Torus United Kingdom. Simulations were conducted using the above mentioned boiling models, the obtained results were compared against the experimental data and also different boiling models are compared with each other whenever possible to test their applicability. From the simulations conducted on the Hypervapotron geometries it is found that the Transition boiling model can capture the thermal performance (in terms of tracing the experimental data) better than any other model both quantitatively and qualitatively, covering the different boiling regimes shown by the experiments ( that is no boiling, nucleate boiling and hard boiling regimes), than the other models

Computational Thermal Fluid Dynamic Analysis of Cooling Systems for Fusion Reactor Components

In a fusion reactor during plasma operation, the heat loads on plasma facing components can be as high as 5 MW/m2 [1], which should be removed by a proper mechanism to prevent the damage of reactor components. In order to handle such high heat fluxes a suitable heat sink with proper thermal hydraulics is required. In the recent past several heat sinks have been proposed; among which the Hypervapotron heat sink, operating in the highly subcooled boiling regime, is considered as one of the potential candidates. In order to accurately predict the performance of the system, a thermal hydraulic analysis is required. This thesis employs a Computational Fluid Dynamic (CFD) approach to do the thermal hydraulic analysis of the subcooled flow boiling inside the Hypervapotron channel. For this purpose four boiling models are tested using two commercial CFD codes. The four boiling models tested are Rensselaer Polytechnic Institute (RPI) boiling model [2] available in ANSYS-FLUENT 13, Bergles-Rohsenow (BR) model [3] implemented as an external User Defined Function (UDF) in the FLUENT code, the Rohsenow boiling model [4] extended with the capability of transition to film boiling for high heat fluxes available in STAR-CCM+ 7.02, and finally Transition boiling model [4] available in STAR-CCM+ 7.02. These models are used to test the thermal performance of Hypervapotron using the experimental data (showing the variation of temperature with heat flux) obtained from the experiments conducted at Efremov Institute Russia and Joint European Torus United Kingdom. Simulations were conducted using the above mentioned boiling models, the obtained results were compared against the experimental data and also different boiling models are compared with each other whenever possible to test their applicability. From the simulations conducted on the Hypervapotron geometries it is found that the Transition boiling model can capture the thermal performance (in terms of tracing the experimental data) better than any other model both quantitatively and qualitatively, covering the different boiling regimes shown by the experiments ( that is no boiling, nucleate boiling and hard boiling regimes), than the other model

Computational thermal fluid dynamic analysis of Hypervapotron heat sink for high heat flux devices application

n fusion devices, plasma is the environment in which light elements fuse producing energy. More than 20% of this power reaches the surface of plasma facing components (e.g. The divertor targets, first wall), where the heat flux local value can be several MW/m2. In order to handle such heat fluxes several coolants are proposed such as water, helium and liquid metals along with different heat sink devices, such as Swirl tubes, Hypervapotrons, Jet cooling, Pin-fins, etc. Among these, Hypervapotron concept, operating in highly subcooled boiling regime with water as a coolant is considered as one of the potential candidates. In this paper, a Computational Fluid Dynamic (CFD) approach is used to analyze the boiling flow inside Hypervapotron channel using two different boiling models: Rohsenow boiling model and Transition boiling model, these models are available in the commercial CFD code STARCCM+, and uses Volume of Fluid approach for the multiphase flow analysis. They are benchmarked using experimental data obtained from experiments conducted at Joint European Torus, UK. The simulated results are then compared with each other and also with other simulated data available to test the quantitative, qualitative features of boiling models in modeling nucleate as well as hard boiling regimes

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications

Computational thermal fluid dynamic analysis of Hypervapotron heat sink for high heat flux devices application

n fusion devices, plasma is the environment in which light elements fuse producing energy. More than 20% of this power reaches the surface of plasma facing components (e.g. The divertor targets, first wall), where the heat flux local value can be several MW/m2. In order to handle such heat fluxes several coolants are proposed such as water, helium and liquid metals along with different heat sink devices, such as Swirl tubes, Hypervapotrons, Jet cooling, Pin-fins, etc. Among these, Hypervapotron concept, operating in highly subcooled boiling regime with water as a coolant is considered as one of the potential candidates. In this paper, a Computational Fluid Dynamic (CFD) approach is used to analyze the boiling flow inside Hypervapotron channel using two different boiling models: Rohsenow boiling model and Transition boiling model, these models are available in the commercial CFD code STARCCM+, and uses Volume of Fluid approach for the multiphase flow analysis. They are benchmarked using experimental data obtained from experiments conducted at Joint European Torus, UK. The simulated results are then compared with each other and also with other simulated data available to test the quantitative, qualitative features of boiling models in modeling nucleate as well as hard boiling regime

Optimization of the pressure drops in the helium in-module manifolds of the EU DEMO “Optimized Conservative” HCLL Breeding Blanket

In Europe the technology to develop DEMO is being supported by HORIZON 2020 research programme under EUROfusion Consortium. One of the key components that DEMO relies on is a working Breeding Blanket (BB). There are currently four Breeding Blanket concepts being developed in Europe, whose primary purpose is to convert kinetic energy of the neutrons into heat and to transfer this converted heat to a coolant, breed tritium to make plant self-sufficient of fuel and to shield magnets from radiation. This paper describes helium distribution system in the “Optimized Conservative” concept of Helium Cooled Lithium Lead (HCLL) BB, according to the HCLL 2014 Design Description Document. This contains three manifolds to distribute the helium in to several parts of the BB namely the First Wall, Cooling Plates, horizontal Stiffening Plates and the vertical Stiffening Plates. The work mainly focuses on optimization study and reduction of the pressure drops in the helium manifolds. The same is achieved by changing several geometrical parameters to keep the pressure drops under the limit. To achieve the results ANSYS-FLUENT commercial Computational fluid dynamic code is used. The final results include the modified geometry which gives lowest pressure drops for the given mass flow rate

CFD analysis of flow boiling in the ITER first wall

This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles–Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost

CFD analysis of flow boiling in the ITER first wall

This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles-Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cos

Computation of JT-60SA TF coil temperature margin using the 4C code

The recently developed 4C thermal-hydraulic code, currently under validation, is used to compute the temperature margin of the superconducting NbTi toroidal field coil of the ITER satellite tokamak JT-60SA, for the nominal burn operation of the machine. Repetitive conditions in the simulation are reached after two plasma pulses. For given (computed) thermal load distribution on winding and coil case due to nuclear heating only, helium temperature ∼4.4K at the winding inlet, reference strand, and the most recent conductor and winding layout, the computed minimum margin occurs in pancake #7 at the peak field location and turns out to be ∼0.2-0.3K above the design value of 1.2