CFD investigation of wire-wrapped fuel rod bundles and flow sensitivity to bundle size

L. M. Brockmeyer, F. S. Sarikurt, Y. A. Hassan, Elia Merzari

Research output: Chapter in Book/Report/Conference proceedingConference contribution

3 Scopus citations

Abstract

The development of a meshing and simulation procedure for the modeling of a 19 pin wire-wrapped fuel rod bundle is detailed. Parameters varied include base size, prism layer thickness, inlet condition and turbulence model. Care was taken to ensure the flow was highly resolved around walls, and the use of wall functions was avoided. The model was validated against established pressure drop correlations. The meshing and simulation procedure was then used to model 19, 37, and 61 pin fuel rod bundles. Again the pressure drops were compared against and agreed favorably with established pressure drop correlations. The velocity distributions of each bundle were found and analyzed qualitatively and quantitatively to determine at what bundle size interior subchannel velocity distribution can be considered independent of bundle size. A qualitative look at the velocity profiles indicates that the interior subchannels of each bundle were affected by the walls. The quantitative analysis reveals that for interior subchannels, the difference is negligible for certain subchannels. Results indicate that the velocity of the 19 pin bundles centermost subchannels agree within 1% with the corresponding subchannels of the 61 pin bundle, and are thus considered independent of bundle size. The two innermost layers of subchannels in the 37 pin bundle are similarly independent of bundle size. A single layer of subchannels acting as a buffer to wall effects was sufficient to isolate interior subchannels from wall effects.

Original languageEnglish (US)
Title of host publicationInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
PublisherAmerican Nuclear Society
Pages6678-6691
Number of pages14
ISBN (Electronic)9781510811843
StatePublished - Jan 1 2015
Event16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015 - Chicago, United States
Duration: Aug 30 2015Sep 4 2015

Publication series

NameInternational Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015
Volume8

Other

Other16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2015
CountryUnited States
CityChicago
Period8/30/159/4/15

All Science Journal Classification (ASJC) codes

  • Instrumentation
  • Nuclear Energy and Engineering

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