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Davydov, M.V.; Melikhov, V.I.; Melikhov, O.I.
Proceedings of the International Conference Nuclear Energy for New Europe 20032003
Proceedings of the International Conference Nuclear Energy for New Europe 20032003
AbstractAbstract
[en] Assessments of the RELAP5/MOD3.2 computer code using critical heat flux data from three sets of experiments have been performed independently by analysts at the Electrogorsk Research and Engineering Center and the Idaho National Engineering and Environmental Laboratory. The experiments, performed at the KS-1 and V-200 facilities, investigated dryout at the top of rod bundles with geometry typical of VVER reactors. The two assessments were compared, investigating differences in the input models and explaining the resultant differences in the calculations. The differences between the two sets of calculations were generally much smaller than the differences between the calculations and the data. Both assessments found that the code calculations were in minimal agreement with the data, and recommended the development of a more applicable critical heat flux model for the code. (author)
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Ravnik, M.; Zagar, T. (Nuclear Society of Slovenia (Slovenia)) (eds.); Nuclear Society of Slovenia, Ljubljana (Slovenia). Funding organisation: Inst. Jozef Stefan, Ljubljana (Slovenia); NUMIP, Krsko (Slovenia); Ministry of Education, Science and Sport of Slovenia, Ljubljana (Slovenia); Westinghouse Electric Systems Europe S.A., Brussels (Belgium); Framatome, Paris (France); Agency for Radwaste Management, Ljubljana (Slovenia); Inetec, Zagreb (Croatia); Elmont, Krsko (Slovenia); Slovenian Nuclear Safety Administration, Ljubljana (Slovenia); Inst. of Metals and Technology, Ljubljana (Slovenia); Inst. of Metal Constructions, Ljubljana (Slovenia); Q Techna, Krsko (Slovenia); Faculty of Mechanical Engineering, Univ. of Ljubljana (Slovenia); Graduate Program Nucelar Engineering, Univ. of Ljubljana (Slovenia); 827 p; ISBN 961-6207-21-0; ; 2003; [6 p.]; International Conference Nuclear Energy for New Europe 2003; Portoroz (Slovenia); 8-11 Sep 2003; Also available from Slovenian Nuclear Safety Administration, Zelezna cesta 16, Ljubljana (SI) or Nuclear Society of Slovenia, Jamova 39, Ljubljana (SI); 3 refs., 5 figs.
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COMPUTER CODES, ENERGY TRANSFER, ENRICHED URANIUM REACTORS, FLUID MECHANICS, FUEL ASSEMBLIES, HOMOGENEOUS REACTORS, HYDRAULICS, MECHANICS, POWER REACTORS, PWR TYPE REACTORS, REACTOR COMPONENTS, REACTORS, RESEARCH AND TEST REACTORS, RESEARCH REACTORS, SOLID HOMOGENEOUS REACTORS, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
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