2008•Applied Radiation and IsotopesRequires access

Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code

K. Khattab, I. Sulieman

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Key concepts: Irradiation, Neutron temperature, Neutron source, Neutron, Radiochemistry, Neutron flux, FOIL method, Materials science

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Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code — Research Paper | ScholarLens