Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Flow Distribution and Pressure Loss in Subchannels of a Wire-Wrapped 37-pin Rod Bundle for a Sodium-Cooled Fast Reactor

  • Chang, Seok-Kyu (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute) ;
  • Euh, Dong-Jin (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute) ;
  • Choi, Hae Seob (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute) ;
  • Kim, Hyungmo (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute) ;
  • Choi, Sun Rock (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute) ;
  • Lee, Hyeong-Yeon (Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute)
  • Received : 2015.07.29
  • Accepted : 2015.12.30
  • Published : 2016.04.25
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Abstract

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

Keywords

References

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