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Experimental Study on Critical Heat Flux in Three Pin Bundle With Wire Spacer for Boiling Water Reactor

[+] Author Affiliations
Sho Tanabe, Dan Tri Le, Masatoshi Kondo, Minoru Takahashi

Tokyo Institute of Technology, Tokyo, Japan

Paper No. ICONE24-60997, pp. V005T15A073; 5 pages
doi:10.1115/ICONE24-60997
From:
  • 2016 24th International Conference on Nuclear Engineering
  • Volume 5: Student Paper Competition
  • Charlotte, North Carolina, USA, June 26–30, 2016
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5005-3
  • Copyright © 2016 by ASME

abstract

Related to the high conversion type BWR, the experimental study on the CHF were performed with a forced convection type water flow loop. In the test section of the loop, three heater pins as a rod bundle are installed, and water flows around the heater pins in the triangular flow channel at atomospheric pressure. The ratio of distance between two heater pins and the diameter of the heater pin (i.e., P/D) was the constant as 1.18 in all the experiments. This experiment conducted at the constant mass flux of 435 kg/m2s. The wire spacers, which were made of stainless steel wire coated by Teflon tube, were wound on the heater pins. The pitches of the wire spacer wound on the surface of the heater pins were 50 mm and 100mm. The water temperature at the inlet of the test section was controlled in the range between 60 and 90 °C by the pre-heater. The parameters in the current work were the pitch of the wire spacer and the quality at the location of the burnout (i.e., local quality). The results indicated that the CHF obtained in the test section with the wire spacer was larger than that without the wire spacer. The wire spacer must promote the CHF value. The smaller pitch of the wire spacer resulted to the smaller CHF value.

Copyright © 2016 by ASME

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