Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지

Wire-wrap Models for Subchannel Blockage Analysis

  • Ha K.S. (Korea Atomic Energy Research Institute) ;
  • Jeong H.Y. (Korea Atomic Energy Research Institute) ;
  • Chang W.P. (Korea Atomic Energy Research Institute) ;
  • Kwon Y.M. (Korea Atomic Energy Research Institute) ;
  • Lee Y.B. (Korea Atomic Energy Research Institute)
  • Published : 2004.04.01
Download PDF
⟨ Previous Next ⟩

Abstract

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

Keywords

References

  1. Kim, W. S., et aI., 2002, A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly, Annals of Nuclear Energy, 29, 303-321 https://doi.org/10.1016/S0306-4549(01)00041-X
  2. Davis, A. L. et aI., 1979, SABRE I - A computer program for the calculation of three dimensional flows in rod clusters, AEEW-R 1057
  3. Ninokata, H., et aI., 1987, Distributed resistance modeling of wire-wrapped rod bundles, Nuclear Engineering and Design, 104,93-102 https://doi.org/10.1016/0029-5493(87)90306-2
  4. Stewart, C. W., et aI., 1977, COBRA-IV: The model and the method, BNWL-2214
  5. Gunter, A. Y., and Shaw, W. A., 1945, A general correlation of friction factors of various types of surfaces in crossflow, Transaction of ASME, 67, 643-660
  6. Suh, K. Y., and Todreas, N. E., 1987, An experimental correlation of cross-flow pressure drop for triangular array wire-wrapped rod assemblies, Nuclear Technology, 76, 229-240
  7. Fontana, M. H., et al., 1974, Temperature distribution in the duct wall and at the exit of 19-rod simulated LMFRB fuel assembly. Nuclear Technology, 24, 176-200
  8. Domanus, H. M., et aI., 1980, Applications of the COMMIX code using the porous medium formulation, Nuclear Engineering and Design, 62,81-100 https://doi.org/10.1016/0029-5493(80)90022-9
  9. Macdougall, J. D. and Lillington, J. N., 1984, The SABRE code for fuel rod cluster thermohydraulics, Nuclear Engineering and Design, 82, 171-190 https://doi.org/10.1016/0029-5493(84)90210-3
  10. Kirsch, D., 1974, Investigations on the flow and temperature distribution downstream of local coolant blockages in rod bundle subassemblies, Nuclear Engineering and Design, 31, 266-279 https://doi.org/10.1016/0029-5493(75)90147-8