2009•Progress in Nuclear EnergyRequires access

Calculation and comparison of the neutron energy flux spectra in the Syrian MNSR irradiation sites using the MCNP-4C code

K. Khattab, I. Sulieman

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Key concepts: Neutron flux, Neutron temperature, Neutron source, Neutron, Irradiation, Flux (metallurgy), Neutron cross section, Spectral line

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Calculation and comparison of the neutron energy flux spectra in the Syrian MNSR irradiation sites using the MCNP-4C code — Research Paper | ScholarLens