0
Research Papers

COBRA-TF Simulation of DNB Response During Reactivity-Initiated Accidents Using the NSRR Pulse Irradiation Experiments

[+] Author and Article Information
Vefa N. Kucukboyaci

Westinghouse Electric Company LLC,
1000 Westinghouse Drive, Suite 233, Cranberry Township, PA 16066
e-mail: kucukbvn@westinghouse.com

Liping Cao

Westinghouse Electric Company LLC,
1000 Westinghouse Drive, Suite 427, Cranberry Township, PA 16066
e-mail: caol@westinghouse.com

Yixing Sung

Westinghouse Electric Company LLC,
1000 Westinghouse Drive, Suite 233, Cranberry Township, PA 16066
e-mail: sungy@westinghouse.com

1Corresponding author.

Manuscript received February 12, 2015; final manuscript received January 12, 2016; published online June 17, 2016. Assoc. Editor: Jovica R. Riznic.

ASME J of Nuclear Rad Sci 2(3), 031002 (Jun 17, 2016) (5 pages) Paper No: NERS-15-1018; doi: 10.1115/1.4032594 History: Received February 12, 2015; Accepted January 21, 2016

COBRA-TF (Coolant Boiling in Rod Arrays–Two Fluid), or CTF, is a transient subchannel code selected to be the reactor core thermal-hydraulic simulation tool in the multiphysics code-development project of the Consortium for Advanced Simulation of Light Water Reactor (CASL) sponsored by the US Department of Energy (DOE). In this paper, CTF’s capability for departure from nucleate boiling (DNB) prediction is evaluated by modeling and simulating power-burst experiments with high-burnup pressurized water reactor (PWR) fuel rods, conducted at the Nuclear Safety Research Reactor (NSRR) in Japan. Experiments using reactor fuel segments have been modeled and simulated to evaluate CTF’s prediction capability for onset of DNB and heat transfer from single-phase to post-critical heat flux (CHF) during fast reactivity-initiated accident (RIA) transients. The calculations demonstrated that CTF is able to simulate a fast transient with a large power pulse. CTF predicted DNB occurrence in all of the cases, after the power pulse, consistent with experimental observations. In the experiments, all cases experienced DNB, as predicted by the correlations in CTF; however, fuel failure occurred in only two of the cases: one at the peak power and the other after the peak power. The remaining cases survived with enthalpies significantly higher than those that failed while experiencing DNB occurrences.

FIGURES IN THIS ARTICLE
<>
Copyright © 2016 by ASME
Your Session has timed out. Please sign back in to continue.

References

Figures

Grahic Jump Location
Fig. 1

Locations of test segments removed from parent rods

Grahic Jump Location
Fig. 2

Power input as function of time for each TK test case

Grahic Jump Location
Fig. 3

Minimum DNBR with W-3 correlation for each TK test case

Grahic Jump Location
Fig. 4

Minimum DNBR with Biasi correlation for each TK test case

Grahic Jump Location
Fig. 5

Fuel pellet rim temperatures at axial midpoint

Grahic Jump Location
Fig. 6

Clad surface temperatures at axial midpoint with W-3 correlation

Grahic Jump Location
Fig. 7

Clad surface temperatures at axial midpoint with Biasi correlation

Grahic Jump Location
Fig. 8

Heat fluxes at axial midpoint with W-3 correlation

Grahic Jump Location
Fig. 9

Heat fluxes at axial midpoint with Biasi correlation

Tables

Errata

Discussions

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In