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Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

[+] Author Affiliations
Alexander D. Vasiliev

Nuclear Safety Institute (IBRAE), Moscow, Russia

Paper No. IMECE2009-10960, pp. 1583-1593; 11 pages
  • ASME 2009 International Mechanical Engineering Congress and Exposition
  • Volume 9: Heat Transfer, Fluid Flows, and Thermal Systems, Parts A, B and C
  • Lake Buena Vista, Florida, USA, November 13–19, 2009
  • Conference Sponsors: ASME
  • ISBN: 978-0-7918-4382-6 | eISBN: 978-0-7918-3863-1
  • Copyright © 2009 by ASME


The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Copyright © 2009 by ASME



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