Nuclear cell homogenization code for thermal reactor

Like in the most nuclear cell homogenization, integral transport equation is also used to solve the neutron transport problem, especially by using collision probability method. Collision probability method has advantage that for relatively simple geometry moreover by its symmetrical feature it is possible to treat angular integration analytically. Neutron transport was solved by physical neutron characteristic in that energy range. Spectrum neutron calculation used 70 groups' energy, and for thermal energy range was divided to 48 energy groups. Calculation for fast energy range we used microscopic cross section data from SLAROM library, and for thermal energy range we used experiment data from ENDF/B VI then we interpreted with Code System NJOY7.0. For example calculation we used cell with one-dimensional cylindrical geometry, which divides to 3 regions spatial, 1 region for fuel, 1 region for cladding, and another for coolant. Treatment for calculation nuclear cell homogenization used linear equation