Determination of neutron flux in the void of a reactor cell by equivalent region method

Full text: This paper deals with the flux distribution in case of void in the reactor cell with natural uranium fuel rods (ru = 1.29 cm), reactor channel (r1 = 3.5 cm) and graphite moderator (rm = 8.9 cm). The void is replaced by equivalent material having only neutron absorption cross section. A homogeneous neutron source is placed in this region. The source is defined to have the number of neutron from the source is equal to the number of absorbed neutrons per unit volume. The atoms of this equivalent medium do not change the direction or energy of neutrons. One-group P3 approximation of spherical harmonics method was used. Equivalent region was divided into 5 subregions with equal absorption of 0.1 cm-1. A few iterations were done in each subregion to minimize the error in achieving the equality of the number of created and absorbed neutrons per unit volume