Thermal-hydraulics analysis of the new Penn State Breazeale Reactor core design using ansys fluent code

D. Uçar, Kenan Unlu

Research output: Contribution to journalArticle

Original languageEnglish (US)
Pages (from-to)301-302
Number of pages2
JournalTransactions of the American Nuclear Society
Volume106
StatePublished - 2012

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Reactor cores
Hydraulics
Hot Temperature

All Science Journal Classification (ASJC) codes

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

Cite this

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