Dhaka University Journal of Applied Science & Engineering

• Farhana Islam Farha
  Department of Nuclear Engineering, University of Dhaka, Dhaka-1000
• Md Hossain Sahadath
  Department of Nuclear Engineering, University of Dhaka, Dhaka-1000

The perturbation in temperature profile and fissile content due to variation in the central gap of an isolated annular cylindrical fuel rod of a pressurized water reactor was studied by analytical calculation. Four different models namely UO2+Zircaloy-4+He, UO2+ Zr-1%Nb +He, MOX+Zircaloy-4+He, and MOX+ Zr-1%Nb +He were considered. The radial temperature profile was generated for different ratios (α) of outer to inner fuel radius. Lower fuel temperature was observed for small values of α and vice versa. The peak fuel temperature and temperature drop across the fuel pellets were calculated. Models of MOX fuel showed higher peak fuel temperature and large temperature drop than the models of UO2 for the same fuel cladding. Zr-1%Nb cladding results in a slightly higher fuel temperature than Zircaloy-4 for the same fuel composition. The faster changes of these parameters with α were found for UO2 than MOX fuel. The change in fissile loading with α was also studied and a sharp increase is observed if exceeds α=2.50

1. J. J Duderstadt and L. J. Hamilton, Nuclear Reactor Analysis, Wiley, New York, 1976.
2. Y. Oka and H. Madarame, Nuclear Reactor Design, New York: Springer, 2013.
3. P. Hejzlar and M. S. Kazimi, “Annular fuel for high-power-density pressurized water reactors: Motivation and overview”,Nuclear Technology, vol. 160, no. (1), pp. 2–15, 2007.
4. M. A. Mozafari and F. Faghihi, “Design of annular fuels for a typical VVER-1000 core: Neutronic investigation, pitch optimization, and MDNBR calculation”,Annals of Nuclear Energy, vol. 60, pp. 226–234, 2013.
5. N. M. Zaidabadi, and G. R. Ansarifar, “Design of a Small Modular Nuclear Reactor with dual cooled annular fuel and investigation of the fuel inner radius effect on the power peaking factor and natural circulation parameters”, Annals of Nuclear Energy, vol. 138, 107185, 2020.
6. J. R Lamarsh and A. J. Baratta, Introduction to Nuclear Engineering, New Jersey: Prentice-Hall, 2001.
7. N. E. Todreas and M. S. Kazimi, Nuclear Systems: Volume 1. London: Taylor & Francis, pp. 19-36, 1990.
8. S.G. Popov and 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.
9. W.G. Luscher, K.J. Geelhood and P. Raynaud, Material Property Correlations: Comparisons between FRAPCON-3.5, FRAPTRAN-1.5, and MATPRO. Office of Nuclear Regulatory Research, U.S.NRC. NUREG/CR-7024, Rev. 1. PNNL-19417, Rev. 1. 2014.
10. R. D. McCarty. Thermophysical Properties of Helium. National Bureau of Standards Report. NBS-TN-631, 1972.
11. F. I. Farha and 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, pp. 1‐17, 2021. DOI: 10.1002/htj.22225.