Mathew Ademola Jayeola, Musbaudeen Kewulere Fasasi, Adebimpe Amos Amosun, Ayodeji Olalekan Salau, Babatunde Michael Ojo

Numerical Computation of Fission-Product Poisoning Build-up and Burn-up Rate in a Finite Cylindrical Nuclear Reactor Core

All fission products are classified as reactor poisons because they absorb neutrons to some extent, most of which buildup slowly as the fuel burns up and eventually constitutes a long term reactivity effect in the core. Amidst the numerous fission fragments produced per fission, the presence of Xenon-135 and Samarium-149 has the greatest effect on a reactor core multiplication factor because of their large absorption cross-sections. In this study, we present a modified one-group time independent neutron diffusion equation using the method of Eigen functions and also provided an algorithm to calculate the temperature variations of the neutron fluxes. The solution obtained from the diffusion equation was used to determine the initial thermal neutron flux needed for the reactor startup. The four basic fission-product poisoning buildup and burn-up rate equations were solved using direct integration method and constant flux approximation over a particular time interval. Furthermore, a computer algorithm called Java code for Fission-Product Poisioning Build-up and Burn-up (Jac-FPPB) code was designed to calculate the temperature variations of the neutron fluxes, fission -isotopes cross sections and the atom concentrations of the fission products over a given time interval. The result from Jac-FPPB code showed that the neutron fluxes and neutron energies increase as the temperature of the fuel increases. In addition, the computed atom concentrations of each fission isotopes at any given time interval showed that the isotopes increasingly build up steadily at the initial time interval and rises to a constant level where the buildup rate of the isotopes approximately equals its burn up rate. This study concluded that the designed algorithm (JaC-FPBB code) proved efficient as it could compute the build-up and burn-up rates for the two important fission fragments in a nuclear reactor core. The code is easily accessible and could serve as a tool for the development of nuclear energy in developing countries, especially Nigeria.

Keywords:

Numerical computation Cylindrical reactor core, Fission-Products poisoning, Neutron diffusion equation, Samarium-149, Xenon-135,

PDF

___

Arshad, M. (1994). Study Of Xenon And Samarıum behavıour in the LEU Parr-1 Cores. Pakistan: Nuclear Engineering Division, Pakistan Institute of Nuclear Science & Technology, 4(28), pp. 1-21.
Briesmeister, J. F. (1993). MCNP - A General Monte Carlo N-Particle Transport Code, Los Alamos National Laboratory, New Mexico, USA, pp. 4-12.
Berthou, V., Degueldre, C., Magill, J. (2003). Transmutation Characteristics in Thermal and Fast Neutron Spectra: Application to Americium, Journal of Nuclear Materials, pp. 156-162.
England, T.R., Wilson. W.B., Stamatelatos, M.G. (1976). Fission product data for thermal reactors- A data set for EPRI-CINDER using ENDF/B-IV. USA: Los Alamos Scientific Laboratory, 8(9), 81p.
Fermi, E. (1940). Nuclear disintegrations, In: Electrical Engineering, 59(2), 57-58.
Hermann, O. W., and Wesfall, R. M. (1998). ORIGEN-S: Scale System Module to Calculated Fuel Depletion Actinide Transmutation, Fission Product Build-up, Decay and Associated Radiation Source Terms. Tennessee, USA: Oak Ridge National Laboratory, 6(2), pp. 1-122.
Marcum, W. and Spinrad, B. I. (2013). Nuclear reactor device, pp. 1-8. [Online]. Available: www.britannica.com/technology/nuclear-reactor
Kord, S., and Lulu, L. (2012). Nuclear Reactor Physics, Massachusetts Institute of Technology, 372p.
Lamarsh, J. R. (1966). Introduction to Nuclear Reactor Theory. New york: Addison-Wesley Publishing Company, pp. 255-256.
Lamarsh, J. R. (2001). Introduction to Nuclear Reactor Theory. New york: Addison-Wesley Publishing Company, 501p.
Lamarsh, J. R, and Baratta, A. J. (2001). Introduction to Nuclear Engineering. Newjersy: Prentice-Hall, Inc. Publishing Company, 3, pp. 272-281.
Parma E. J. (2002). BURNCAL- A Nuclear Reactor Burnup Code Using MCNP Tallies. California: Sandia National Laboratories, pp. 3-102.
Stacey, W. M. (2007). Nuclear Reactor Physics, New york: Wiley-VCH Publishing Company, 2, pp. 211-215.
Yesilyurt, G., Clarno, K. T., Gauld, I. C. (2011). Modular Orıgen-S for Multı-Physıcs Code Systems. International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Rio de Janeiro, Brazil, pp. 1-14.