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Computation of Thermal Striping in the Upper Plenum of PGSFR

[+] Author Affiliations
Dong-Eun Kim, Sung-Ho Ko

Chungnam National University, Daejeon, Republic of Korea

Seok-Ki Choi, Tae-Ho Lee

Korea Atomic Energy Research Institute, Daejeon, Republic of Korea

Paper No. ICONE22-30666, pp. V004T10A027; 8 pages
  • 2014 22nd International Conference on Nuclear Engineering
  • Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory
  • Prague, Czech Republic, July 7–11, 2014
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4594-3
  • Copyright © 2014 by ASME


A computational study of a thermal striping in the upper plenum of the PGSFR (Prototype Generation-IV Sodium-cooled Fast Reactor) being developed at KAERI (Korea Atomic Energy Research Institute) is presented. First, previous experimental and numerical studies on the thermal striping are briefly discussed. Both RANS (Reynolds-Averaged Navier-Stokes) and LES (Large Eddy Simulation) approaches are employed for the simulation of thermal striping in the upper plenum of the PGSFR. For the RANS approach, the conventional kε turbulence model is employed and the LES is performed using the WALE (Wall-Adapting Local Eddy-viscosity) model. More than 11.8 million unstructured elements are generated in the upper plenum region of the PGSFR using the ICEM commercial code. From the RANS results, the time-averaged velocity components and temperature field in the complicated upper plenum of PGSFR are calculated. In the LES results, the time history of temperature fluctuation at the several locations of solid walls of UIS (Upper Internal Structure) and IHX (Intermediate Heat Exchanger) are additionally stored. Comparisons of the predicted time-averaged velocity components and temperature between the two methods are also presented. From the temporal variation of temperature at the solid walls, one can find the locations where the thermal stress is large and assess whether the solid structures can endure the thermal stress during the reactor life time.

Copyright © 2014 by ASME



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