The ITER blanket-shield system is the innermost part of the reactor directly exposed to the plasma. Its basic function is to provide the main thermal and nuclear shielding to the vacuum vessel and external reactor components. Its concept is a modular configuration: the different modules consist of a water-cooled stainless-steel shield block, on which a separable first wall (FW) panel is mounted. The FW panels, having typically the dimensions of 1 m x 1.5 m, consist of a complex structure, where the plasma-facing beryllium tiles are cooled by water at the pressure of 40 bar and inlet temperature of 70 °C, which flows in parallel ducts, called fingers, on the back of the tiles. The duct configuration can span from the rectangular hypervapotron geometry to the regular circular tube, according to the level of power deposition from the plasma. We concentrate on the panel of the FW blanket module #6, for which the maximum power density foreseen during operation is ~ 2 MW/m2 allowing the use of the circular tube. Here, in the first of two companion papers, we concentrate on the hydraulic behavior of the coolant in the panel. Starting from the present design status, the detailed hydraulic analysis of the coolant flow inside the panel, including the inlet/outlet pipes, the beam and the manifolds distributing the flow to the 48 fingers is performed using the commercial CFD code ANSYS Fluent. The aim of this analysis is to localize the stagnant regions inside the module and to compute the pressure drop across each finger and across the whole panel. Based on the computed results, an optimization of the panel geometry is proposed, that should achieve a more homogeneous distribution of the flow among the fingers, minimizing at the same time the stagnant regions and the pressure drop.
First Wall panel study for the ITER blanket module #6. Part I: hydraulic optimization.
CAU, FRANCESCA;
2012-01-01
Abstract
The ITER blanket-shield system is the innermost part of the reactor directly exposed to the plasma. Its basic function is to provide the main thermal and nuclear shielding to the vacuum vessel and external reactor components. Its concept is a modular configuration: the different modules consist of a water-cooled stainless-steel shield block, on which a separable first wall (FW) panel is mounted. The FW panels, having typically the dimensions of 1 m x 1.5 m, consist of a complex structure, where the plasma-facing beryllium tiles are cooled by water at the pressure of 40 bar and inlet temperature of 70 °C, which flows in parallel ducts, called fingers, on the back of the tiles. The duct configuration can span from the rectangular hypervapotron geometry to the regular circular tube, according to the level of power deposition from the plasma. We concentrate on the panel of the FW blanket module #6, for which the maximum power density foreseen during operation is ~ 2 MW/m2 allowing the use of the circular tube. Here, in the first of two companion papers, we concentrate on the hydraulic behavior of the coolant in the panel. Starting from the present design status, the detailed hydraulic analysis of the coolant flow inside the panel, including the inlet/outlet pipes, the beam and the manifolds distributing the flow to the 48 fingers is performed using the commercial CFD code ANSYS Fluent. The aim of this analysis is to localize the stagnant regions inside the module and to compute the pressure drop across each finger and across the whole panel. Based on the computed results, an optimization of the panel geometry is proposed, that should achieve a more homogeneous distribution of the flow among the fingers, minimizing at the same time the stagnant regions and the pressure drop.
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