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A Monte Carlo Neutron Photon (MCNP) transport code has been employed to simulate CR-39 plastic track detector as neutron dosimeter at various neutron energies. In each simulation a monoenergetic neutron source of a particular energy was embedded in the center of a sphere made of CR-39. Surrounding the source there were concentric shells of 2 μm CR-39 track detectors. The code, MCNP, was run on personal...
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