<![CDATA[The Research on solution of the neutron diffusion equation with a PWR reactor using recycled uranium fuel at 1⁄6 section of the reactor core with a hexagonal IGT-6 geometry. The purpose of this research is to determine the distribution of the neutron flux in the PWR of recycled uranium fuel. The solution is done by computational simulation using the Dev-C++ programming. The parameters used in this study determine the specifications of the reactor core, determine the volume fraction, determine the atomic density, calculate the macroscopic cross-section with the PIJ module, calculate the neutron diffusion equation, calculate ϕ (x,y) using the Gauss Seidel method. The results obtained in this study are the neutron diffusion equation without a source obtaining the highest relative neutron flux value in group 1 of 4,5729×〖10〗^(-2), with a fission source obtaining the highest relative neutron flux value in group 3 of 7,3327×〖10〗^(-4), with fission and scattering sources obtaining the highest relative neutron flux value found in group 2 of 1,5157×〖10〗^(-3), and 3,200 MW of power is added to the source fission, the value of the neutron flux does not change. This is because the addition of power does not affect the value of the neutron flux.]]>
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