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Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

[+] Author Affiliations
M. Zavisca, M. Ghaderi, M. Khatib-Rahbar

Energy Research, Inc., Rockville, MD

S. Bajorek

U.S. Nuclear Regulatory Commission, Rockville, MD

Paper No. ICONE14-89779, pp. 795-802; 8 pages
doi:10.1115/ICONE14-89779
From:
  • 14th International Conference on Nuclear Engineering
  • Volume 2: Thermal Hydraulics
  • Miami, Florida, USA, July 17–20, 2006
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 0-7918-4243-6 | eISBN: 0-7918-3783-1
  • Copyright © 2006 by ASME

abstract

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U. S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents the results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990s. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models.

Copyright © 2006 by ASME

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