Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지

Effect of Spacer Grids on CHF at PWR Operating Conditions

  • Published : 2001.06.01
Download PDF
⟨ Previous Next ⟩

Abstract

The CHF in PWR rod bundles is usually predicted by the local flow correlation approach based on subchannel analysis while difficulty exists due to the existence of spacer grids especially with mixing vanes. In order to evaluate the effect of spacer grids on CHF, the experimental rod bundle data with various types of spacer grids were analyzed using the subchannel code, COBRA-IV-i. For the Plain grid data, a CHF correlation was described as a function of local flow conditions and heated length, and then the residuals of the CHF in mixing vaned grids predicted by the correlation were examined in various kinds of grids. In order to compensate for the residual, three parameters, distances between grids and from the last grids to the CHF site, and equivalent hydraulic diameter were introduced into a grid parameter function representing the remaining effect of spacer grids predicted most of the CHF data points in plaing grids within $\pm$20 percent error band. Good agreement with the CHF data was also shown when the grid parameter function for mixing vaned grids of a specific design was used to compensate for the residuals of the CHF data predicted by the correlation.

Keywords

References

  1. D.H. Hwang, et al., Development of a rod bundle correction method and its application to predicting CHF in rod bundles, Nucl. Engrg. & Des. 139, 205-220 (1993) https://doi.org/10.1016/0029-5493(93)90157-5
  2. J. Weisman, Methods for detaile d thermal and hydraulic analysis of water-cooled reactors, Nucl. Sci. & Engrg., 57, 255-276 (1975)
  3. P.G. Barnett, An investigation into the validity of certain hypotheses implied by various burnout correlation, U.K. AEA report, AEEW-R214 (1963)
  4. C.L. Wheeler, et al., COBRA-IV-I: An interim version of COBRA for thermal hydraulic analysis of rod bundle nuclear fuel elements and cores, BNWL-1962, (1976)
  5. L.S. Tong, An evaluation of the departure from nucleate boiling in bundles of reactor fuel rods, Nucl. Sci. Engrg. 33, 7-15 (1968)
  6. E.R. Rosal, et al., High pressure rod bundle DNB data with axially non-uniform heat flux, Nucl. Engrg. & Des. 106, 209-220 (1974)
  7. J. Yates, K. Galbraith, and R. Collingham, DNB and post-DNB heat transfer data in 25 rod bundles, XN-75-54, Exxon Nuclear Corp. 1975, presented at 3rd Water Reactor Satety Information Mtg. Washington, D.C., Sep. (1975)
  8. D.C. Groeneveld and W.W. Yousef, Springs,Spacing devices for nuclear fuel bundles: A survey of their effect on CHF, Post-CHF heat transfer and pressure drop, ANS/ASME Int. Topl. Mtg. Nuclear reactor thermal-hydraulics, Saratoga, Spring, N.Y., Oct. (1980)
  9. F. E. Motley, et al., New Westinghouse correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids, Westinghouse report, WCAP-8263, (1976)
  10. H.B. Giap, et al., A new methodology for developing an empirical DNB correlation
  11. F.D. Lawrence, et al., Critical Heat Flux in PWR fuel assemblies, AIChE-ASME Heat Transfer Conf.,, Salk Lake City, Utah, Aug. 15-17 (1977)
  12. J.C.M. Leung and R.E. Henry, Prediction of CHF in rod bundles using round tube CHF correlation, ANS Trans. 33, 963-966 (1979)
  13. J. Weisman and S.H. Yang, A theoretically based in critical heat flux prediction for rod bundles at PWR conitions, Nucl. Engrg. & Des. 85, 239-250 (1985) https://doi.org/10.1016/0029-5493(85)90289-4
  14. W.S. Lin, et al., 'Bundle critical power predictions under normal and abnormal conditions in pressurized water reactors, Nucl. Tech. 98, 354-365 (1992)
  15. K. Rheme, The structure of turbulence in rod bundles and the implications on natural mixing between the subchannels, Int. J. Heat Mass Transfer, 35, 567-581 (1992) https://doi.org/10.1016/0017-9310(92)90291-Y
  16. F. de Crecy, The effect of grid assembly mixing vanes on critical heat flux value and azimuthal location in fuel assemblies, NURETH-6, Grenoble (1993)
  17. L.S. Tong, 'Boiling crisis and critical heat flux', U.S. AEC. TID-25887 (1972)
  18. D.G. Reddy and C.F. Fighetti, Parametric study of CHF data Vol.3: Critical heat flux data, EPRI-NP-2609 (1982)
  19. D.G. Reddy and C.F. Fighetti, Parametric study of CHF data Vol.2: A generalized subchannel CHF correlation for PWR and BWR fuel assemblies, EPRI-NP-2609, (1982)
  20. R.V. Macbeth, Burnout analysis, Part 4: Application of a local condition hypothesis to world data for uniformly heated round tubes and rectangular channels, U.K. AEA report, AEEW-R267 (1963)
  21. SAS user's guide: Statistics, version 5 edition, SAS Institution Inc. (1985)
  22. L.C. Edwin, et al., Statistics manual, Dover Publications Inc., New York (1956)
  23. F.H. Bowditch and D.J. Mogford, 'An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements,' UKAEA report: AEEW-R2050 (1987)
  24. Vogel, Mixing input data for SIEMENS test sections used by MIX84 code until 1989, KWU B123/1989/e377 (1989)