Development of sub-channel analysis tool for TRU Fuel Fabrication

Takashi Takata, Yuichiro Manabe, Akira Yamaguchi, Koichi Hishida, Kazuo Ikeda, Unihiro Itoh

Research output: Contribution to journalArticle

Abstract

The development of the fast reactor (FR) cycle is being advanced to utilize plutonium and transuranium (TRU) in Japan. In the fabrication process, it is considered that a fuel pin spirally wrapped with a thin wire is laid horizontally. Then, cooling air flows vertically from the bottom side into the gap of the pin bundle so as to suppress the temperature increase due to decay heat. From the viewpoint of safety assessment during the fabrication, a thermal hydraulic analysis method plays an important role in investigating the maximum temperature and the temperature distribution of the fuel pins. In the present paper, a subchannel analysis tool has been developed. Using the developed tool, the benchmark analysis of the mocked up experiment has been carried out, as well as the numerical investigation of a multidimensional effect of fuel cladding thermal conductivity on the maximum temperature. It is demonstrated that the multidimensional effect of the cladding thermal conductivity is not negligible in the analysis. A good agreement is achieved in the case of a comparatively large clearance size between the side wall and the pin bundle when one considers a natural convection heat transfer at the outermost boundary with a comparatively low computational cost.

Original languageEnglish
Pages (from-to)839-848
Number of pages10
JournalJournal of Nuclear Science and Technology
Volume47
Issue number9
DOIs
Publication statusPublished - 2010

Keywords

  • Cross flow
  • Fast reactor
  • Low decontaminated fuel
  • Subchannel analysis
  • Thermal hydraulics
  • Transuranium

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering

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  • Cite this

    Takata, T., Manabe, Y., Yamaguchi, A., Hishida, K., Ikeda, K., & Itoh, U. (2010). Development of sub-channel analysis tool for TRU Fuel Fabrication. Journal of Nuclear Science and Technology, 47(9), 839-848. https://doi.org/10.1080/18811248.2010.9711660