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Science and Technology of Nuclear Installations
Volume 2013 (2013), Article ID 725687, 12 pages
Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database
1Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802, USA
2RES/DSA/CDB, US Nuclear Regulatory Commission, Washington, DC 20555-0001, USA
Received 6 July 2012; Accepted 16 November 2012
Academic Editor: Diana Cuervo
Copyright © 2013 M. Avramova et al. This is an open access article distributed under the Creative Commons Attribution License, which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.
- H. Herkenrath, W. Hufschmidt, U. Jung, and F. Weckermann, “Experimental investigation of the enthalpy and mass flow distribution in 16-rod clusters with BWR-PWR geometries and conditions,” EUR 7575 EN, ISPRA, 1981.
- R. T. Lahey Jr., B. S. Shiralkar, and D. W. Radcliffe, Two-Phase Flow and Heat Transfer in Multi-Rod Geometries: Subchannel and Pressure Drop Measurements in a Nine-Rod Bundle for Diabatic and Adiabatic Conditions, 1970, GEAP-13049, GE.
- A. Rubin, et al., “OECD/NRC benchmark based on NUPEC PWR subchannel and bundle tests (PSBT). Volume I: experimental database and final problem specifications,” NEA/NSC/DOC, 2010.
- “CTF—a thermal-hydraulic subchannel code for LWRs transient analyses. User’s manual,” Technical Report, RDFMG, The Pennsylvania State University, 2009.
- C. Y. Payk, et al., “Analysis of FLECHT SEASET 163-rod blocked bundle data using COBRA-TF,” Tech. Rep. NRC/EPRI/Westinghouse-12, 1985.
- M. Avramova, D. Cuervo, K. Ivanov, et al., “Improvements and applications of COBRA-TF for stand-alone and coupled LWR safety analyses,” in Proceeding of the International Conference on the Physics of Reactors (PHYSOR '06), Vancouver, Canada, September 2006.
- TRACE V5. 0 Theory Manual, “Field Equations, Solution Methods, and Physical Models,” USNRC, Washington DC.
- R. T. Lahey and F. J. Moody, The Thermal Hydraulics of a Boiling Water Nuclear Reactor, American Nuclear Society (ANS), 1993.
- J. T. Rogers and R. G. Rosehart, “Mixing by turbulent interchange in fuel bundles, correlations and inferences,” ASME 72-HT-53, 1972.
- S. G. Beus, “A two-phase turbulent mixing model for flow in rod bundles,” Tech. Rep. WAPD-T-2438, Bettis Atomic Power Laboratory, 1970.
- M. Avramova, K. Ivanov, and L. E. Hochreiter, “Analysis of steady state and transient void distribution predictions for phase I of the OECD/NRC BFBT Benchmark using CTF/NEM,” in Proceedings of the 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-12 '07), Pittsburgh, Pa, USA, October 2007.
- M. Avramova, A. Rubin, et al., “OECD-NEA/US-NRC/NUPEC PWR Subchannel and bundle test (PSBT) benchmark, Volume II: final results of phase I on void distribution,” OECD/NEA Report, 2011.
- M. Valette and C. E. A. -Grenoble, Private Correspondence.