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Comparison of the Subchannel Code CTF to Steady-State and Transient Heat Transfer Experiments at Intermediate Pressures in Water

Pan Wu, David Novog

Nuclear Technology / Volume 205 / Number 1-2 / January-February 2019 / Pages 364-376

Technical Paper / dx.doi.org/10.1080/00295450.2018.1495000

Received:April 27, 2018
Accepted:June 26, 2018
Published:December 12, 2018

The CTF code is a subchannel thermal-hydraulic code developed based on the COBRA-TF code. In this work, the CTF code is used to predict the single- and two-phase heat transfer, pressure drop, onset of nucleate boiling, and dryout heat flux in water at several temperatures and pressures under steady-state and transient conditions. The conditions cover a range of pressures from 2 to 6 MPa, flows from 1000 to 2500 kg/(m2∙s), and inlet subcooling from 40°C to 70°C. Experimental heat balance tests show agreement between coolant enthalpy change and the electrical power with a difference of no more than 1.0%. Steady-state experiments were performed at constant inlet conditions in a cylindrical directly heated Inconel test section where the wall temperatures were measured at each power level. For each steady-state test, the experimental boiling curve is compared to CTF predictions. Transient experiments were performed by initiating a blowdown from the test section outlet plenum using a fast-acting valve with an open time of less than 100 ms. The time of dryout in these transient experiments is compared with the CTF results to clarify the pressure transient effect on the dryout prediction.

 
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