Application of MCNP for neutronic calculations at VR-1 training reactor
Czech Technical University in Prague, Department of Nuclear Reactors, V Holesovickach 2, Prague 8, 180 00
The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.
Key words: Monte Carlo / MCNP / nuclear reactor analysis / critical experiment
© Owned by the authors, published by EDP Sciences, 2014