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Report No.

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

Kimura, Nobuyuki; Miyakoshi, Hiroyuki; Kamide, Hideki 

Thermal stratification after a scram is one of main thermal loads of the reactor vessel (R/V) of sodium cooled fast reactor. 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 smaller than that in the case of the higher plug position.



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Category:Nuclear Science & Technology



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