Monte Carlo Simulation of the Greek Research Reactor Neutron Irradiation Positions Using Mcnp

Prediction of neutron flux at the irradiation devices of a research reactor facility is essential for the design and evaluation of experiments involving material irradiations. A computational model of the Greek Research Reactor (GRR-1) was developed using the Monte Carlo code MCNP with continuous energy neutron cross-section data evaluations from ENDF/B-VI library. The model included detailed geometrical representation of the fuel and control assemblies, beryllium reflectors, irradiation devices and the graphite pile. The MCNP model was applied to predict neutron flux at the in-pool irradiation positions and the graphite pile. The MCNP estimated neutron fluxes were compared with measurements using activation foils and a good agreement between calculated and experimental results was observed.