1306 核融合炉設計における熱流動に関連する異常事象解析(GS-5 管路)

概要

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Thermal-hydraulic characteristics in a fusion experimental reactor such as ITER (international Thermonuclear Experimental Reactor) under an ingress-of-coolant event (ICE) and a loss-of-vacuum event (LOVA) were analyzed numerically. Void fraction and velocity vector distributions in the fusion reactor during an ICE event were clarified quantitatively using the TRAC-PF1 code, which was originally developed for accident analyses in light water reactors, and pressure rise speeds inside a vacuum vessel (VV) were evaluated for various injected water conditions. On the other hand, activated dust mobilization behavior in the VV of the fusion reactor during a LOVA event were predicted with a newly developed code and released dust rates from the VV through a breach to the outside were estimated under roof-break and side-break cases. Predicted released dust rates were less than 15%.
2000-03-16