Numerical study on CTF code to predict void fraction in PWR sub channel conditions
Main Article Content
Abstract
CTF is a version of the widely used COBRA-TF code with capability of 3D simulation for core sub channel thermal hydraulics behavior. Recently, CTF is reviewed and the consideration of CTF to predict void fraction in PWR sub channel conditions such as subcooled region still need more investigation. Due to the fact that the Chen’s correlation of heat transfer coefficient is developed for relatively low pressure and high quality conditions associated with forced convection vaporization, and is not strictly valid for PWR operation conditions, so that, in this study, some runs of single channel in the benchmark based on NUPEC PWR Sub channel and Bundle Tests (PSBT) are used to investigate void fraction prediction by CTF in subcooled region and also to verify some remarkable notice of CTF from other authors. The goal of the study is to evaluate deviation for CTF void fraction prediction in PWR sub channel conditions.
Article Details
Keywords
CTF, COBRA-TF, void fraction, PWR, PSBT, Chen’s correlation
References
[2] Adam Joseph Rubin., “OECD/NRC Benchmark based on NUPEC PWR Sub channel and Bundle.
Tests (PSBT) - Analysis using sub channel code CTF and system code TRACE”, A Thesis in Nuclear Engineering for the Degree of Master of Science, The Pennsylvania State University, May 2011.
[3] M. Avramova, A. Velazquez-Lozada, and A. Rubin., “Comparative Analysis of CTF and Trace
Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database”, Hindawi Publishing Corporation, Science and Technology of Nuclear Installations, Volume 2013, Article ID 725687, pp. 2-5, Accepted 16 November 2012.
[4] Murray Cameron Thames., “Application and Assessment of a 9-equation Sub channel Methodology to Rod Bundles”, A thesis in Nuclear Engineering for the Degree of Master of Science, North Carolina State University, pp. 2-6, 44-48, 2014.
[5] J. Michael Doster., “Assessment of the Performance of COBRA-TF for the Prediction of Subcooled Boiling Conditions in Rod Bundles”, CASL-U-2013-0201-000, September 30, p. 24, 2013.
[6] Robert K. Salko., “Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients”, A dissertation in Nuclear Engineering for the Degree of Doctor of Philosophy, The Pennsylvania State University, pp. 7-17, May 2012.
[7] Subcooled Boiling Data from Rod Bundles, EPRI, Palo Alto, CA; 2002. 1003383.
[8] S. G. Beus, “A two-phase turbulent mixing model for flow in rod bundles,” Tech. Rep. WAPD-T-2438, Bettis Atomic Power Laboratory, 1970.
[9] N. Kurul and M. Z. Podowski, "On the modeling of multi dimensional effects in boiling channels”, Proceedings of the 27th National heat transfer Conference, Minneapolis, July 1991.
[10] H. Anglart, O. Nylund, N. Kurul, and M. Z. Podowski, "CFD prediction of flow and phase distribution in fuel assemblies with spacers," Proceedings of the NURETH-7, Saratoga Springs, New York, 1995, published in: Nuclear Eng. & Design (NED), Vol. 177, pp. 215-228, 1997.
[11] G. Rabello dos Anjos, Jacopo Buongiorno., “Bubble Condensation Heat Transfer in Subcooled Flow Boiling at PWR Conditions: a Critical Evaluation of Current Correlations”, Massachusetts Institute of Technology Cambridge, MA, USA, pp. 8-11, September 2013.