الفهرس 
IntroductionOnly 14 pages are availabe for public view
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Abstract
MCNPX code based on Mont Carlo Method , is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. The model is used to calculate the distribution of pin by pin power and flux inside the assembly. The effect of axial variation of water (coolant) density, and of control rods motion on the neutron flux and power distribution is analyzed. The effect of addition of Gd2 O3 to natural uranium (0.711%) on both the thermal neutron flux and normalized power are analysed. The concentration of U235, U238, Pu239, and its isotopes is also calculated at burn-up 50 GWD/T.