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Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J.
Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-51995
Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-51995
AbstractAbstract
[en] An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident
Primary Subject
Source
Block, R.C.; Feiner, F. (American Nuclear Society, La Grange Park, IL (United States)) (comps.); Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); Japan Society of Multiphase Flow, Osaka (Japan); 862 p; Sep 1995; p. 336-352; 7. international meeting on nuclear reactor thermal hydraulics; Saratoga, NY (United States); 10-15 Sep 1995; Also available from OSTI as TI95017077; NTIS; GPO
Record Type
Report
Literature Type
Conference; Numerical Data
Report Number
Country of publication
DATA, ENERGY TRANSFER, ENRICHED URANIUM REACTORS, EVALUATION, FAILURES, HEAVY WATER COOLED REACTORS, HEAVY WATER MODERATED REACTORS, INFORMATION, NUMERICAL DATA, PHWR TYPE REACTORS, PIPELINES, POWER REACTORS, PRESSURE TUBE REACTORS, REACTORS, THERMAL REACTORS, WATER COOLED REACTORS, WATER MODERATED REACTORS
Reference NumberReference Number
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