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Journal of Applied Science & Engineering

Dhaka University Journal of Applied Science & Engineering

Issue: Vol. 7, No. 1, January 2022
Title: Study of the Perturbation in Temperature Profile of an AGR Fuel Pin for Surface Roughness of Cladding by CFD Simulation in Ansys Fluent
Authors:
  • Sadek Hossain Nishat
    Department of Nuclear Engineering, University of Dhaka, Dhaka-1000
  • Farhana Islam Farha
    Department of Nuclear Engineering, University of Dhaka, Dhaka-1000
  • Md Hossain Sahadath
    Department of Nuclear Engineering, University of Dhaka, Dhaka-1000
DOI:
Keywords: AGR, CFD, Fluent, Fuel Rod, Ribs, Temperature.
Abstract:

The surface roughness of nuclear fuel cladding plays a crucial role in the thermal-hydraulic response of the Advanced Gas Cooled reactor (AGR). In the present work, the change in the temperature distribution from an isolated AGR fuel rod to primary coolant due to cladding roughness was studied by computational fluid dynamics (CFD) simulation in Ansys Fluent software. Square transverse ribs of the various pitch to height ratios (p/k) were considered as the surface roughness. Radial temperature profiles from fuel to coolant were generated. Lower fuel temperature was found for the fuel rod with a rough cladding surface as compared to the smooth cladding surface. The peak fuel temperature was determined and found to decrease with decreasing values of (p/k). Temperature drop across the fuel and from fuel to coolant was also studied.

References:
  1. N. E. Todreas, and M. S. Kazimi, “Nuclear Systems: Volume 1”, London: Taylor & Francis. Chapter 2, 1990
  2. A.E. Bergles, V. Nirmalan, G.H. Junkhan, R.L. Webb, “Bibliography on Augmentation of Convective Heat and Mass Transfer—II”, United States Department of Energy, 1983
  3. J. R Lamarsh, A.J. Baratta, “Introduction to Nuclear Engineering” Third ed, New Jersey: Prentice-Hall, 2001, Chapter 8
  4. . Keshmiri, “Three-dimensional simulation of a simplified advanced gas-cooled reactor fuel element”, Nuclear Engineering, and Design, vol. 241, pp. 4122– 4135, 2011
  5. A. Keshmiri, and J. Gotts, “Thermal-hydraulic analysis of four geometrical design parameters in rib-roughened channels”, Numerical Heat Transfer, Part A, pp 305–327, 2011
  6. A. Ooi, G. Iaccarino, P.A. Durbin, M. Behnia, “Reynolds averaged simulation of flow and heat transfer in ribbed ducts”, International Journal of Heat and Fluid Flow, vol. 23, pp. 750–757, 2002
  7. W. F Cope, “The Friction and Heat Transmission Coefficients of Rough Pipes” Froc. Inst. Mechanical Engineers, pp. 99–105, 1941
  8. T. S. Ravigururajan, “General correlations for pressure drop and heat transfer for single-phase turbulent flows in ribbed tubes” Iowa State University, pp. 18–19, 1986
  9. E. Nonbel, “Description of the Advanced Gas-Cooled Type of Reactor (AGR)” Nord. Nucl. Saf. Res. NKS/RAK2 (96) TR-C2.,1996.
  10. S. G. Popov, J. J. Carbajo, V. K. Ivanov, G. L. Yoder, “Thermophysical Properties of MOX and UO2 Fuels Including the Effects of Irradiation”. J. Russ. Res. Cent. Oak Ridge Natl. Lab. pp. 9–24. 2000
  11. C. S. Kim, “Thermophysical Properties of Stainless Steels”, Argonne Natl. Lab. pp. 2–13. 1975
  12. M. M. Rathore, “Thermal Engineering”, pp. 1067–1116, 2010
  13. Ansys, Release 17.0, Help System, Mechanical APDL Documentation, ANSYS, Inc
  14. S. Zainal, C. Tan, C. J. Sian, and T. J. Siang, “Ansys simulation for Ag / HEG Hybrid Nanofluid in Turbulent Circular Pipe”, vol. 23, no. 1, pp. 20–35, 2016
  15. J. J. Duderstadt, and L. J. Hamilton, “Nuclear Reactor Analysis”, New York: Wiley, Chapter 12. 1976.
  16. S. G. Popov, J.J. Carbajo, V.K. Ivanov and G.L. Yoder’ “Thermophysical Properties of MOX and UO2 Fuels Including the Effects of Irradiation”, Engineering Technology Division, Fissile Materials Deposition Program, Oak Ridge National Laboratory. ORNL/TM/-2000/351, 2000
  17. F. I. Farha and M.H. Sahadath, “Modeling and Simulation of Radial Temperature, Thermal Heat Flux, and Thermal Gradient Distribution in Solid and Annular Nuclear Fuel Element of Uranium Dioxide”, DUJASE, vol. 6, no.1, pp. 1-8, 2021
  18. F. I. Farha, M. H. Sahadath, S. H. Nishat, “Thermal‐hydraulic analysis of UO2 and MOX fuel considering different cladding materials at various burnup levels in pressurized water reactor”. Heat Transfer, Wiley LLC, pp.1‐17. 2021.