The coolability characteristics of a eutectic metal debris bed, which has a low melting point, have been assessed from the viewpoint of in-vessel retention. Ag-50wt%Zircaloy eutectic alloy, constituents of which are principal metals constituting a reactor core, was chosen as the particulate core-metal debris in the present experimental study. Ag particulates and shortly chopped Zircaloy tubes were melted by induction heating, and then the molten Ag-50wt%Zircaloy was dropped into a water pool with 80cm in depth, resulting in debris particulates. The upper interface temperature of the particulate metal debris, which was electrically heated to simulate decay heat, ranged from 500°C to 900°C, and the temperature of a water layer at the bottom side was kept at 100°C. The heat flux and the temperature at the upper interface were measured for 30 minutes. Under the wet condition where heat conducted from debris particulates to a water layer produces steam, it is confirmed that the particulate eutectic-alloy debris bed is oxidized and the perfectly oxidized parts with thin cross-section are cracked into pieces. The mass median diameter measured after each run clearly decreases compared with that measured before the run. Sieving after each run shows that an amount of small particulates less than 1mm, which is expected to produce a high capillary force, drastically increases due to oxidation. The present experimental results therefore show that the particulate eutectic-alloy debris bed exposed to a vapor atmosphere is oxidized in a short time period and consequently could be cooled because of a capillary force of small particulates produced by oxidization.
- Nuclear Engineering Division
An Experimental Study on Coolability of Particulate Core-Metal Debris Bed With Oxidization
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Aoki, H, Sugiyama, K, Su, GH, Sakashita, H, & Kojima, Y. "An Experimental Study on Coolability of Particulate Core-Metal Debris Bed With Oxidization." Proceedings of the 12th International Conference on Nuclear Engineering. 12th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 25–29, 2004. pp. 847-852. ASME. https://doi.org/10.1115/ICONE12-49557
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