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 second of two companion papers, we concentrate on the thermal-hydraulic behavior of the panel. Starting from the optimized design obtained from the hydraulic analysis - part I of this study - we start analyzing the finger geometry, to assess at which level the computed evolution of the temperature field in the structure is influenced by the details of the model of the cooling pipe and coolant. Based on the results of this first step, the thermal (-hydraulic) analysis of the entire panel is carried out, considering two full plasma cycles. The computed evolution of the temperature distribution in the structure during the transient is presented. The local values of the heat transfer coefficient (HTC) computed in the panel are also shown and compared to the correlations available from the literature.
First Wall panel study for the ITER blanket module #6. Part II: thermal analysis.
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 second of two companion papers, we concentrate on the thermal-hydraulic behavior of the panel. Starting from the optimized design obtained from the hydraulic analysis - part I of this study - we start analyzing the finger geometry, to assess at which level the computed evolution of the temperature field in the structure is influenced by the details of the model of the cooling pipe and coolant. Based on the results of this first step, the thermal (-hydraulic) analysis of the entire panel is carried out, considering two full plasma cycles. The computed evolution of the temperature distribution in the structure during the transient is presented. The local values of the heat transfer coefficient (HTC) computed in the panel are also shown and compared to the correlations available from the literature.
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