Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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3-D CFD Analysis of the CANDU-6 Moderator Circulation Under Nnormal Operating Conditions

  • Published : 2004.12.01
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Abstract

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

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References

  1. L.N. Carlucci and 1. Cheung, 'The Effects of Symmetric/Asymmetric Boundary Conditions on the Flow of an Internally Heated Fluid,' Numerical Methods for Partial Differential Equations, 2, 47-61 (1986) https://doi.org/10.1002/num.1690020105
  2. R.G. Huget, J.K. Szymanski, and W.I. Midvidy, 'Status of Physical and Numerical Modelling of CANDU Moderator Circulation,' Proceedings of 10th Annual Conference of the Canadian Nuclear Society, Ottawa (1989).
  3. R.G. Huget, J.K. Szymanski, and W.I. Midvidy, 'Experimental and Numerical Modelling of Combined Forced and Free Convection in a Complex Geometry with Internal Heat Generation,' Proceedings of 9th International Heat Transfer Conference, Vol. 3, 327 (1990)
  4. W.M. Collins, 'PHOENICS2 Model Report for Wolsong 2/3/4 Moderator Circulation Analysis,' Wolsong NPP 2/3/4, 86-03500-AR-053, Revision 0 (1995)
  5. P. Seodijono, W.M. Collins, and T. De, 'Moderator Analysis for In-Core and Out-of-Core Loss of Coolant Accident (LOCA),' Wolsong NPP 2/3/4, 86-03500-AR-052, Revision 0 (1995)
  6. N.E. Todreas and M.S. Kazimi, Nuclear System II: Elements of Thermal Hydraulic Design, Chap. 5, Hemisphere Publishing Corporation (1990)
  7. G.!. Hadaller, R.A. Fortman, J. Szymanski, W.I. Midvidy and D.J.' Train, 'Frictional Pressure Drop for Staggered and In Line Tube Bank with Large Pitch to Diameter Ratio,' Preceedings of 17th CNS Conference, Frederiction, New Brunswick, Canada (1996)
  8. K. Aydogdu, K.Y. Kim, and Y.I. Kim, 'Radiation Heating Report,' Wolsong NPP 2/3/4, 86-03320-AR-004, Revision 2 (1994)
  9. Churl Yoon, Bo Wook Rhee, and Byung-Joo Min, '3-D CFD Analysis of the CANDU-6 Moderator Transient for the 35% RIH Break with Loss of ECC Injection,' The 10th International Topiclal Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), Seoul, Korea, October (2003)
  10. Churl Yoon, Bo Wook Rhee, and Byung-Joo Min, 'Development and Validation of the 3-D CFD Model for CANDU-6 Moderator Temperature Predictions,' Accepted for the publication in Nuclear Technology, vol. 148, no.3, 259-267 (2004)