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MCNP-A General Monte Carlo N-Particle Transport Code

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On normalization of fluxes and reaction rates in MCNP criticality calculations

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Engineering, High-energy Nuclear Reaction, Physics +10

Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization

Luka Snoj, Andrej Trkov, Radojko Jačimović et al. · Applied Radiation and Isotopes · 2010 · 86 citations

Engineering, Nuclear Engineering, Nuclear Safety +10