Pengembangan Kode Komputer untuk Homogenisasi Sel Bahan Bakar Nuklir yang Diperkaya untuk Reaktor Termal

Abstract

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. This method has the advantage that for relatively simple geometry the angular integration may be carried out analytically. Neutron transport equation was solved in accordance with the physical neutron characteristic in its energy range. The neutron spectrum calculation used 70 energy group and for thermal energy range the energy was divided into 48 energy groups.For calculation in the fast energy range we used microscopic cross section data from SLAROM library, while for thermal energy range we used experiment data from ENDF/B VI and then we
interpreted it with the used of Code System NJOY97.0. As an example calculation we considered one-dimensional cylindrical cell which was divided into 3 regions i.e, for fuel, cladding, and coolant. The nuclear cell homogenization calculation was treated by the use of linear equation.