Predictions of critical heat flux using the ASSERT-PV subchannel code for a CANFLEX variant bundle /

समान संसाधन

• Nupec BFBT subchannel void distribution analysis using the matra and mars codes /
  द्वारा: Hwang, Dae-Hyun.
• Comparison of dryout power data between canflex MK-V and canflex MK-IV bundle strings in uncrept and crept channels /
  द्वारा: Jun, Ji Su.
• An experimental study on post-chf heat transfer for low flow of water in a 3x3 rod bundle /
• Assertiveness : practical approach /
  द्वारा: Holland, Stephanie.
  प्रकाशित: (1990)
• Thermalhydraulic evaluations for a canflex bundle with natural or recycled uranium fuel in the uncrept and crept channels of a candu-6 reactor /
  द्वारा: Jun, Ji-Su.