Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

Nuclear Engineering and Design

Volumn 240, Issue 10, 2010, Pages 3087-3094

  Ammirabile, Luca ^a  

^a TNO   (Netherlands)

Author keywords

[No Author keywords available]

Indexed keywords

CHANNEL CODE; CODE CAPABILITY; COMPUTATIONAL TOOLS; CORE DESIGN; COUNTER CURRENT; CRITICAL PRESSURES; DENSITY VARIATIONS; EXPERIMENTAL DATA; FUEL ASSEMBLY; FUEL ASSEMBLY DESIGN; GENERATION IV; HEAT TRANSFER AND PRESSURE DROP; HEAT TRANSFER ENHANCEMENT; HIGHER EFFICIENCY; HYDRIDE FUEL; NEUTRONIC CODES; REACTOR DESIGNS; SUPER-CRITICAL REGIONS; SUPERCRITICAL WATER; SUPERCRITICAL WATER CONDITIONS; SUPERCRITICAL WATER REACTORS; TECHNICAL CHALLENGES; THERMAL-HYDRAULIC ANALYSIS; WATER PROPERTIES;

CHANNEL CODING; DESIGN; HEAT TRANSFER; HYDRAULIC TOOLS; NUCLEAR FUELS; THERMODYNAMIC PROPERTIES;

LIGHT WATER REACTORS;

EID: 77957717240     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2010.05.025     Document Type: Conference Paper

References (20)

1. Basile, D., Beghi, M., Chierici, R., Salina, E., Brega, E., 1999. COBRA-EN: an upgraded version of the COBRA-3C/MIT code for thermal hydraulic transient analysis of light water reactor fuel assemblies and cores. Report 1010/1, ENELCRTN, Milano.
2. Bishop, A.A., Sandberg, R.O., Tong, L.S., 1964. Forced convection heat transfer to water at near critical temperatures and supercritical pressures, WCAP-2056-P, Part-III-B.
3. L. Cao, Y. Oka, Y. Ishiwatari, and Z. Shang Core design and subchannel analysis of a superfast reactor Journal of Nuclear Science and Technology 45 2008 138 148
4. Cheng, X., Schulenberg, T., 2001. Heat transfer at supercritical pressures - literature review and application to an HPLWR. Scientific Report FZKA 6609, FzK.
5. X. Cheng, T. Schulenberg, D. Bittermann, and P. Rau Design analysis of core assemblies for supercritical pressure conditions Nuclear Engineering and Design 223 2003 279 294
6. Dittus, F.W., Boelter, L.M.K., 1930. Heat Transfer in Automobile Radiators of Tubular Type. Publ. Eng. 2 (13), Univ. California, Berkley, pp. 443-461.
7. Greenspan, E., Todreas, N., Petrovic, B., 2006. Use of Solid Hydride Fuel for Improved LWR Core Designs. Univ. of California, Tech. Report DOE/SF/22615-1.
8. J. Hofmeister, C. Waata, J. Starflinger, T. Schulenberg, and E. Laurien Fuel assembly design study for a reactor with supercritical water Nuclear Engineering and Design 237 2007 1513 1521
9. Y. Jiyang, W. Songtao, and J. Baoshan Development of sub-channel analysis code for CANDU-SCWR Progress in Nuclear Energy 49 2007 334 350
10. D.S. Rowe COBRA-3C: A Digital Computer Program for Steady-State and Transient Thermal-Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements, BNWL-1965, Battelle 1973 Pacific Northwest Laboratory Richland, Washington
11. Stewart, C.W., Cuta, J., Montgomery, S., Kelly, J., Basehore, K., George, T., Rowe, D., 1983. VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, EPRI NP-2511-CCM, Electric Power Research Institute Research Project 1584-1.
12. I. Toumi, A. Bergeron, D. Gallo, E. Royer, and D. Caruge FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications Nuclear Engineering and Design 200 2000 139 155
13. H. Tsige-Tamirat, L. Ammirabile, and M.A. Fütterer On use of hydride fuel in HPLWR Proceedings of the 4th International Symposium on Supercritical Water-Cooled Reactors Heidelberg, Germany, March 8-11 2009
14. U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum A Technology Roadmap for Generation IV Nuclear Energy Systems 2002 http://www.gen-4.org/PDFs/GenIVRoadmap.pdf
15. W. Wagner, and A. Kruse The industrial standard IAPWS-IF97 for the thermodynamic properties and supplementary equations for other properties Properties of Water and Steam 1997 Springer Heidelberg, Germany
16. C.L. Wheeler, C.W. Stewart, R.J. Cena, D.S. Rowe, and A.M. Sutey COBRA-IV-I: An Interim Version of COBRA for Thermal-Hydraulic Analysis of Rod Bundles Nuclear Fuel Elements and Cores, BNWL-1962, Battelle 1976 Pacific Northwest Laboratory Richland, Washington
17. X-5 Monte Carlo Team, 2003. MCNP - A General Monte Carlo N-Particle Transport Code, Version-5. Los Alamos National Laboratory, Tech. Report, LA-UR-03-1987.
18. K. Yamagata, K. Nishikawa, S. Hasegawa, T. Fujii, and S. Yoshida Forced convection heat transfer to supercritical water flowing in tubes International Journal of Heat and Mass Transfer 15 1972 2575 2593
19. J. Yoo, Y. Oka, Y. Ishiwatari, J. Yang, and J. Liu Subchannel analysis of supercritical light water-cooled fast reactor assembly Nuclear Engineering and Design 237 2007 1096 1105
20. Y.J. Yoo, D.H. Hwang, and D.S. Sohn Development of a subchannel analysis code MATRA applicable to PWRs and ALWRs Journal of the Korean Nuclear Society 31 1999 314 327