Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach

ABSTRACT

Nuclear fuel cell calculation is one of the most complicated steps of neutron transport problems in the reactor core. A few numerical methods use neutron flat flux (FF) approximation to solve this problem. In this approach, neutron flux spectrum is assumed constant in each region. The solution of neutron transport equation using collision probability (CP) method based on non flat flux (NFF) approximation by introducing linear spatial distribution function implemented to a simple cylindrical annular cell has been carried out. In this concept, neutron flux spectrum in each region is different each other because of an existing of the spatial function. Numerical calculation of the neutron flux in each region of the cell using NFF approach shows a fairly good agreement compared to those calculated using existing SRAC code and FF approach. Moreover, calculation of the neutron flux in each region of the nuclear fuel cell using NFF approach needs only 6 meshes which give equivalent result when it is calculated using 24 meshes in FF approach. This result indicates that NFF approach is more efficient to be used to calculate the neutron flux in the regions of the cell than FF approach.

Nuclear fuel cell calculation is one of the most complicated steps of neutron transport problems in the reactor core. A few numerical methods use neutron flat flux (FF) approximation to solve this problem. In this approach, neutron flux spectrum is assumed constant in each region. The solution of neutron transport equation using collision probability (CP) method based on non flat flux (NFF) approximation by introducing linear spatial distribution function implemented to a simple cylindrical annular cell has been carried out. In this concept, neutron flux spectrum in each region is different each other because of an existing of the spatial function. Numerical calculation of the neutron flux in each region of the cell using NFF approach shows a fairly good agreement compared to those calculated using existing SRAC code and FF approach. Moreover, calculation of the neutron flux in each region of the nuclear fuel cell using NFF approach needs only 6 meshes which give equivalent result when it is calculated using 24 meshes in FF approach. This result indicates that NFF approach is more efficient to be used to calculate the neutron flux in the regions of the cell than FF approach.

Cite this paper

M. Ali Shafii, Z. Su’ud, A. Waris and N. Kurniasih, "Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach,"*World Journal of Nuclear Science and Technology*, Vol. 2 No. 2, 2012, pp. 49-53. doi: 10.4236/wjnst.2012.22008.

M. Ali Shafii, Z. Su’ud, A. Waris and N. Kurniasih, "Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach,"

References

[1] Z. Su’ud, “Computer Code for Homogenization of Nuclear Fuel Cell in the Fast Reactor,” Proceeding of a Workshop in Computational Science and Nuclear Technology, Bandung, 24-25 February 1998, pp. 110-115.

[2] Z. Su’ud, Y. K. Rustandi and R. Kurniadi, “Parallel Computing in the Calculation of a Constant Group of Nuclear,” Proceeding of the seventh seminar of Technology and Safety of NPP and Nuclear Facilities, Bandung, 19 February 2002, pp. 17-22.

[3] M. A. Shafii and Z. Su’ud, “Development of Cell Homogenization Code Using General Geometry Approach,” International Conference on Advances in Nuclear Science and Engineering, Bandung, 13-14 November 2007, pp. 403-406.

[4] M. A. Shafii, Z. Su’ud, A. Waris and N. Kurniasih, “Development of Cell Homogenization Code with Collision Probability Method,” International Conference of Mathematics and Natural Science, Bandung, 28-30 October 2008, pp. 169-175.

[5] K. Okumura, T. Kugo, K. Kaneko and K. Tsuchihashi, “SRAC 2006: A Comprehensive Neutronics Calculation Code System,” Japan Atomic Energy Agency, Ibaraki, 2007.

[6] M. Nakagawa and K. Tsuchihashi, “SLAROM: A Code for Cell Calculation of Fast Reactor,” Japan Atomic Energy Research Institute, Ibaraki, 1984.

[7] S. S. Rao, “Finite Element Method in Engineering,” Pergamon Press, New York, 1983.

[8] T. Hazama, “Private Communication,” 2008.

[9] T. Hazama, G. Chiba and K. Sugino, “Development of a Fine and Ultra-Fine Group Cell Calculation Code SLAROM-UF for Fast Reactor Analyses,” Journal of Nuclear Science and Technology, Vol. 43, No. 8, 2006, pp. 908-918. doi:10.3327/jnst.43.908

[1] Z. Su’ud, “Computer Code for Homogenization of Nuclear Fuel Cell in the Fast Reactor,” Proceeding of a Workshop in Computational Science and Nuclear Technology, Bandung, 24-25 February 1998, pp. 110-115.

[2] Z. Su’ud, Y. K. Rustandi and R. Kurniadi, “Parallel Computing in the Calculation of a Constant Group of Nuclear,” Proceeding of the seventh seminar of Technology and Safety of NPP and Nuclear Facilities, Bandung, 19 February 2002, pp. 17-22.

[3] M. A. Shafii and Z. Su’ud, “Development of Cell Homogenization Code Using General Geometry Approach,” International Conference on Advances in Nuclear Science and Engineering, Bandung, 13-14 November 2007, pp. 403-406.

[4] M. A. Shafii, Z. Su’ud, A. Waris and N. Kurniasih, “Development of Cell Homogenization Code with Collision Probability Method,” International Conference of Mathematics and Natural Science, Bandung, 28-30 October 2008, pp. 169-175.

[5] K. Okumura, T. Kugo, K. Kaneko and K. Tsuchihashi, “SRAC 2006: A Comprehensive Neutronics Calculation Code System,” Japan Atomic Energy Agency, Ibaraki, 2007.

[6] M. Nakagawa and K. Tsuchihashi, “SLAROM: A Code for Cell Calculation of Fast Reactor,” Japan Atomic Energy Research Institute, Ibaraki, 1984.

[7] S. S. Rao, “Finite Element Method in Engineering,” Pergamon Press, New York, 1983.

[8] T. Hazama, “Private Communication,” 2008.

[9] T. Hazama, G. Chiba and K. Sugino, “Development of a Fine and Ultra-Fine Group Cell Calculation Code SLAROM-UF for Fast Reactor Analyses,” Journal of Nuclear Science and Technology, Vol. 43, No. 8, 2006, pp. 908-918. doi:10.3327/jnst.43.908