Researches of steady-state thermal hydraulic performance of a large power PWR had been made using the COBRA III C/MIT-2 code. The basic thermal hydraulic parameters and characteristics of the core channels were studied. The results showed that the hottest channel and the maximum enthalpy rise hot channel were not boiling. The result will lay the foundation for further studies in the thermal-hydraulic design of a large advanced PWR.
Subchannel Analysis of Large Power PWR
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Xie, F. "Subchannel Analysis of Large Power PWR." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2A: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02AT09A054. ASME. https://doi.org/10.1115/ICONE22-30428
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