Indian Journal of Science and Technology

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The MCNP-4C Monte Carlo code was used to model a 2 MW thermal VVR-S research reactor. The neutron with continuous energy cross sections of the ENDF/B-VI library was applied to MCNP-4C to calculate the thermal and fast neutron fluxes. The computed neutron flux showed that the MCNP-4C can be used in the reactors similar to VVR-S reactor.

Keywords

MCNP, Multiplication Factor, Neutron Flux, VVR-S
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