Author(s)

Adilson Costa da Silva, Aquilino Senra Martinez, Alessandro da Cruz Gonçalves

Nuclear Engineering Department, Federal University of Rio de Janeiro, Rio de Janeiro, Brazil.

In this electronic article we use the one-dimensional multigroup neutron diffusion equation to reconstruct the neutron flux in a slab reactor from the nuclear parameters of the reactor, boundary and symmetry condition, initial flux and k_eff. The diffusion equation was solved analytically for one single homogeneous fuel region and for two regions considering fuel and reflector. To validate the method proposed, the results obtained in this article were compared using reference methods found in the literature.

Neutron Diffusion Equation; Neutron Flux; Slab Reactor

A. Silva, A. Martinez and A. Gonçalves, "Reconstruction of the Neutron Flux in a Slab Reactor," World Journal of Nuclear Science and Technology, Vol. 2 No. 4, 2012, pp. 181-186. doi: 10.4236/wjnst.2012.24028.