2015•Indian Journal of Science and TechnologyOpen access

Computing the Neutron Flux in VVRS-2 Reactor using MCNP-4C

M. Rahgoshay, Masoud Khaleghi, Mehdi Hashemi-Tilehnoee

Abstract

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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Key concepts: Neutron flux, Research reactor, Neutron temperature, Nuclear physics, Nuclear engineering, Neutron, Monte Carlo method, Flux (metallurgy)