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Experimental study on thermal stratification in a reactor vessel of innovative sodium cooled fast reactor; Mitigation approach of temperature gradient across stratification interface

ナトリウム冷却高速炉の原子炉容器内温度成層化現象に関する実験研究; 成層界面内温度勾配の緩和策の検討

木村 暢之; 宮越 博幸; 上出 英樹

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki

ナトリウム冷却高速炉のスクラム過渡時温度成層化現象に関して、1/10縮尺炉上部プレナム水流動試験装置を用い、成層界面挙動に対する燃料交換機貫通孔プラグ高さの影響を評価した。プラグを炉上部プレナム内に深く挿入することで、炉容器壁近傍の成層界面温度勾配は小さくできるとともに、界面の変動に伴う温度変動も低減できることがわかった。

An innovative sodium cooled fast reactor has been investigated as part of the fast reactor cycle technology development (FaCT) project. Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V). An upper inner structure (UIS) has a slit in radial direction for fuel handling. A water experiment using an 1/10 scale model was carried out. Steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at a dipped plate for a fuel handling machine, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 13% smaller than that in the case of the higher plug position.

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