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Japan Atomic Energy Agency has been performing the research, development and design of a test blanket module with a water-cooled solid breeder for ITER. For our design, the TBM is mainly composed of a first wall, two side walls, a back wall and membrane panels of bulkhead sections for pebbles. The temperature of a coolant pressurized up to 15 MPa is designed as 553 K and 598 K in an inlet and an outlet of the test blanket module, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. In this study, the Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the flow rates in the channels of the side wall. It results in the inhomogeneous flow rates at each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mockup.

Numerical Simulation of Turbulent Flow of Coolant in a Test Blanket Module of Nuclear Fusion Reactor