Record number :
1594793
Title of article :
An experimental study on local fuel–coolant interactions by delivering water into a simulated molten fuel pool
Author/Authors :
Cheng، نويسنده , , Songbai and Matsuba، نويسنده , , Ken-ichi and Isozaki، نويسنده , , Mikio and Kamiyama، نويسنده , , Kenji and Suzuki، نويسنده , , Tohru and Tobita، نويسنده , , Yoshiharu، نويسنده ,
Issue Information :
روزنامه با شماره پیاپی سال 2014
Pages :
9
From page :
133
To page :
141
Abstract :
Analyses of severe accidents for sodium-cooled fast reactors have indicated that there is the possibility that the accident could proceed into a transition phase where a large whole-core-scale pool containing sufficient fuel to exceed prompt criticality by fuel compaction might be formed. Local fuel–coolant interaction (FCI) in the pool is regarded as one of the probable initiators that could lead to such compactive fluid motions. To clarify the mechanisms underlying this interaction, in this study a series of experiments was conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Based on the experimental data obtained from a variety of conditions, interaction characteristics including the pressure-buildup as well as resultant mechanical energy release and its conversion efficiency, is checked and compared. From the analyses, it is confirmed that under our experimental conditions the water volume, melt temperature and water release position are observable to have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The analyses also suggest that the pressurization and resultant mechanical energy release during local FCIs should be intrinsically limited, due to an observed suppressing role caused by the increasing of coolant volume entrapped within the pool as well as the transition of boiling mode. The evidence and fundamental data from this work will be utilized for future analyses and improved verifications of computer models developed in advanced fast reactor safety analysis codes.
Journal title :
Nuclear Engineering and Design Eslah
Journal title :
Nuclear Engineering and Design Eslah
Serial Year :
2014
Link To Document :
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