Proving test on thermal-hydraulic performance of bwr fuel assemblies

Yoshiaki Tsukuda, Hiroshi Hayashi, Katsuichiro Kamimura, Toshiitsu Hattori, Hirohisa Kaneko, Shinichi Morooka, Torn Mitsutake, Miyuki Akiba, Nobuaki Abe, Masahiko Warashina, Yasuhiro Masuhara, Jiro Kimura, Akira Tanabe, Yuji Nishino, Koujun Isaka, Riichiro Suzuki

Research output: Contribution to journalArticlepeer-review

1 Citation (Scopus)


Nuclear Power Engineering Corporation (NUPEC) has conducted a proving test for thermal-hydraulic performance of BWR fuel (high-burnup 8×8, 9×9) assemblies entrusted by the Ministry of Economy, Trade and Industry (NUPEC-TH-B Project). The high-burnup 8×8 fuel (average fuel assembly discharge burnup: about 39.5GWd/t), has been utilized from 1991. And the 9 × 9 fuel (average fuel assembly discharge burnup: about 45GWd/t). has started to be used since 1999. There are two types (A-type and B-type) of fuel design in 9×9 fuel assembly. Using an electrically heated test assembly which simulated a BWR fuel bundle on full scale, flow induced vibration, pressure drop, critical power under steady state condition and post-boiling transition (post-BT) tests were carried out in an out-of pile test facility that can simulate the high pressure and high temperature conditions of BWRs. This paper completed the results of 9 × 9 fuel combined with the previously reported results of high-burnup 8×8 fuel. As a result of NUPEC-TH-B Project, the validity of the current BWR thermal-hydraulic design method was confirmed and the reliability of BWR thermo-hydraulic fuel performance was demonstrated. Based on the test data, a new correlation of the estimation of fuel rod vibration amplitude, new post-BT heat transfer and rewet correlations for the estimation of fuel rod surface temperature were developed.

Original languageEnglish
Pages (from-to)384-403
Number of pages20
JournalTransactions of the Atomic Energy Society of Japan
Issue number4
Publication statusPublished - 2002
Externally publishedYes


  • BWR type reactors
  • correlation
  • critical power
  • flow induced vibration
  • heat transfer
  • post-BT
  • pressure drop
  • reliability
  • rewet
  • thermal-hydraulic performance
  • vibration amplitude

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Safety, Risk, Reliability and Quality


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