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

D. Uçar, Kenan Unlu

Research output: Contribution to journalConference article

Original languageEnglish (US)
Pages (from-to)301-302
Number of pages2
JournalTransactions of the American Nuclear Society
Volume106
StatePublished - Dec 1 2012
Event2012 ANS Annual Meeting and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors, NFSM 2012 - Chicago, IL, United States
Duration: Jun 24 2012Jun 28 2012

All Science Journal Classification (ASJC) codes

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

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