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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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

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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).