Abstract
A deuterium‐deuterium accelerator-type neutron generator was installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). This generator is shielded by different hydrogenated and non-hydrogenated materials to reduce the dose rates in the vicinity of the facility. In the work presented in this paper, both SCALE6 and MCNP5 radiation transport codes were used to conduct two independent simulations. The new shielding analysis tool of SCALE6—MAVRIC, with the automatic variance reduction technique of SCALE6, was utilized to estimate and compare the dose rates from the unbiased MCNP simulation. The ultimate goal of this study was to compare the computational effectiveness offered by employing the MAVRIC sequence in the modeling of the neutron generator facility at Missouri S&T.
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