Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle

N. Tarasova, V. Kisina, Timo Vanttola, Olli Tiihonen, A.S. Konjkov, D.L. Prozerov

    Research output: Book/ReportReport

    Abstract

    An experimental program was carried out, where heat transfer and hydraulic resistance were measured in a facility that simulated a fuel assembly of a Soviet VVER-440 nuclear power reactor. The experiments were performed in a test loop with four different rod bundles that consisted of 19 full height, directly electrically heated rods. Most of the tests were conducted close to nominal conditions of the simulated reactor. Heat transfer crisis of dryout type was measured at mass fluxes from 15 to 40 % of the nominal flow of the reactor, when inlet temperature of flow and heat flux were varied in the neighbourhood of the nominal reactor values. Occurrence of heat transfer crisis was independent of heat flux in the covered large steam qualities and a small mass fluxes. No crisis correlation predicted this phenomenon, but the GE and the Hsu-Beckner formulas obtained on an average similar values to those of the experiments. In slow flow decay transients the crisis appeared later than in the stationary experiments. Single phase heat transfer coefficient in the rod bundles was found to be close to the Dittus-Boelter correlation. Two-phase heat transfer coefficient was smaller than expected, but the measurements were disturbed by surface fouling. Single phase friction of turbulent flow was the same as had earlier been obtained in similar conditions. Two phase friction was measured up to large steam qualities beyond dryout in small mass fluxes. Local minimum of friction was observed around the crisis point. The Baroczy curves predict two-phase friction best of the correlations that were tested against the experimental data.
    Original languageEnglish
    Place of PublicationEspoo
    PublisherVTT Technical Research Centre of Finland
    Number of pages72
    ISBN (Print)951-38-1683-4
    Publication statusPublished - 1983
    MoE publication typeD4 Published development or research report or study

    Publication series

    SeriesValtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports
    Number154
    ISSN0358-5077

    Fingerprint

    Hydraulics
    Friction
    Heat transfer
    Mass transfer
    Heat transfer coefficients
    Heat flux
    Steam
    Experiments
    Fouling
    Nuclear energy
    Turbulent flow
    Temperature

    Keywords

    • heat transfer crisis
    • reactor safety
    • rod bundle
    • subchannel analysis
    • transients
    • pressure loss

    Cite this

    Tarasova, N., Kisina, V., Vanttola, T., Tiihonen, O., Konjkov, A. S., & Prozerov, D. L. (1983). Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle. Espoo: VTT Technical Research Centre of Finland. Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports, No. 154
    Tarasova, N. ; Kisina, V. ; Vanttola, Timo ; Tiihonen, Olli ; Konjkov, A.S. ; Prozerov, D.L. / Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle. Espoo : VTT Technical Research Centre of Finland, 1983. 72 p. (Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports; No. 154).
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    abstract = "An experimental program was carried out, where heat transfer and hydraulic resistance were measured in a facility that simulated a fuel assembly of a Soviet VVER-440 nuclear power reactor. The experiments were performed in a test loop with four different rod bundles that consisted of 19 full height, directly electrically heated rods. Most of the tests were conducted close to nominal conditions of the simulated reactor. Heat transfer crisis of dryout type was measured at mass fluxes from 15 to 40 {\%} of the nominal flow of the reactor, when inlet temperature of flow and heat flux were varied in the neighbourhood of the nominal reactor values. Occurrence of heat transfer crisis was independent of heat flux in the covered large steam qualities and a small mass fluxes. No crisis correlation predicted this phenomenon, but the GE and the Hsu-Beckner formulas obtained on an average similar values to those of the experiments. In slow flow decay transients the crisis appeared later than in the stationary experiments. Single phase heat transfer coefficient in the rod bundles was found to be close to the Dittus-Boelter correlation. Two-phase heat transfer coefficient was smaller than expected, but the measurements were disturbed by surface fouling. Single phase friction of turbulent flow was the same as had earlier been obtained in similar conditions. Two phase friction was measured up to large steam qualities beyond dryout in small mass fluxes. Local minimum of friction was observed around the crisis point. The Baroczy curves predict two-phase friction best of the correlations that were tested against the experimental data.",
    keywords = "heat transfer crisis, reactor safety, rod bundle, subchannel analysis, transients, pressure loss",
    author = "N. Tarasova and V. Kisina and Timo Vanttola and Olli Tiihonen and A.S. Konjkov and D.L. Prozerov",
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    Tarasova, N, Kisina, V, Vanttola, T, Tiihonen, O, Konjkov, AS & Prozerov, DL 1983, Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle. Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports, no. 154, VTT Technical Research Centre of Finland, Espoo.

    Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle. / Tarasova, N.; Kisina, V.; Vanttola, Timo; Tiihonen, Olli; Konjkov, A.S.; Prozerov, D.L.

    Espoo : VTT Technical Research Centre of Finland, 1983. 72 p. (Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports; No. 154).

    Research output: Book/ReportReport

    TY - BOOK

    T1 - Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle

    AU - Tarasova, N.

    AU - Kisina, V.

    AU - Vanttola, Timo

    AU - Tiihonen, Olli

    AU - Konjkov, A.S.

    AU - Prozerov, D.L.

    PY - 1983

    Y1 - 1983

    N2 - An experimental program was carried out, where heat transfer and hydraulic resistance were measured in a facility that simulated a fuel assembly of a Soviet VVER-440 nuclear power reactor. The experiments were performed in a test loop with four different rod bundles that consisted of 19 full height, directly electrically heated rods. Most of the tests were conducted close to nominal conditions of the simulated reactor. Heat transfer crisis of dryout type was measured at mass fluxes from 15 to 40 % of the nominal flow of the reactor, when inlet temperature of flow and heat flux were varied in the neighbourhood of the nominal reactor values. Occurrence of heat transfer crisis was independent of heat flux in the covered large steam qualities and a small mass fluxes. No crisis correlation predicted this phenomenon, but the GE and the Hsu-Beckner formulas obtained on an average similar values to those of the experiments. In slow flow decay transients the crisis appeared later than in the stationary experiments. Single phase heat transfer coefficient in the rod bundles was found to be close to the Dittus-Boelter correlation. Two-phase heat transfer coefficient was smaller than expected, but the measurements were disturbed by surface fouling. Single phase friction of turbulent flow was the same as had earlier been obtained in similar conditions. Two phase friction was measured up to large steam qualities beyond dryout in small mass fluxes. Local minimum of friction was observed around the crisis point. The Baroczy curves predict two-phase friction best of the correlations that were tested against the experimental data.

    AB - An experimental program was carried out, where heat transfer and hydraulic resistance were measured in a facility that simulated a fuel assembly of a Soviet VVER-440 nuclear power reactor. The experiments were performed in a test loop with four different rod bundles that consisted of 19 full height, directly electrically heated rods. Most of the tests were conducted close to nominal conditions of the simulated reactor. Heat transfer crisis of dryout type was measured at mass fluxes from 15 to 40 % of the nominal flow of the reactor, when inlet temperature of flow and heat flux were varied in the neighbourhood of the nominal reactor values. Occurrence of heat transfer crisis was independent of heat flux in the covered large steam qualities and a small mass fluxes. No crisis correlation predicted this phenomenon, but the GE and the Hsu-Beckner formulas obtained on an average similar values to those of the experiments. In slow flow decay transients the crisis appeared later than in the stationary experiments. Single phase heat transfer coefficient in the rod bundles was found to be close to the Dittus-Boelter correlation. Two-phase heat transfer coefficient was smaller than expected, but the measurements were disturbed by surface fouling. Single phase friction of turbulent flow was the same as had earlier been obtained in similar conditions. Two phase friction was measured up to large steam qualities beyond dryout in small mass fluxes. Local minimum of friction was observed around the crisis point. The Baroczy curves predict two-phase friction best of the correlations that were tested against the experimental data.

    KW - heat transfer crisis

    KW - reactor safety

    KW - rod bundle

    KW - subchannel analysis

    KW - transients

    KW - pressure loss

    M3 - Report

    SN - 951-38-1683-4

    T3 - Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports

    BT - Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle

    PB - VTT Technical Research Centre of Finland

    CY - Espoo

    ER -

    Tarasova N, Kisina V, Vanttola T, Tiihonen O, Konjkov AS, Prozerov DL. Experiments on heat transfer crisis and hydraulic resistance in a 19-rod bundle. Espoo: VTT Technical Research Centre of Finland, 1983. 72 p. (Valtion teknillinen tutkimuskeskus. Tutkimuksia - Research Reports; No. 154).