Evaluation and benchmarking of gamma dose rate employing different nuclear data libraries for MCNP code at the decommissioning stage of Ignalina NPP

A comparative study was performed to reveal the differences of three nuclear data libraries for gamma dose rate calculations when applied to heterogeneous environment in the case of decommission of the Ignalina Nuclear Power Plant (INPP). The following libraries were investigated by employing the Monte Carlo n-particle transport code (MCNP): ENDF/B-VII, JEFF-3.1 and JENDL-3.3, based on the experiments performed for gamma radiation dose rate measurements inside the emergency core cooling system (ECCS) tank with surface radioactive contamination up to 54 Bq/cm2. MCNP precise simulation and the benchmark between the libraries highlighted the differences of results for the selected case of this investigation. The results revealed that the ENDF library is trustworthy for various dose and shielding calculations and similar applications since it showed a statistically satisfied agreement between the simulation results and experimental data.