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چکیده
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The neutron as an unbranched member of the nucleus structure. Plays a crucial role in the study of nuclear forces and the issue of neutron diffusion in a neutron system, Such as nuclear reactors, is of particular importance. Whit the advent of nuclear science and technology and increasing concerns about the safetly of reactors, the issue of controlling power and preventing an increase in the core temperature of the reactor and subsequent melting of the heart is of great importance. The diffusion equation can be used to calculate the neutron flux and to determine how they are distributed in a neutron environment and in environments where neutron absorption is not high. After solving the equation, flux is obtained analytically and the result of the calculation is compared with the results of the simulation. The MCNP code is used to model the 3D-centered configuration of the core, which includes all fuel rods, fixed and movable (control), central channel, sensor channel, coolant (moderator) and so on. The purpose of this study is to obtain the flux graph from the analytical method and the simulation chart, in the next step the adaptation of these two graphs and then the calculation of the neutron coefficient are calculated.
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