Evaluation of modern techniques for the calculation of ionizing radiation shielding

The work is aimed to review the assessment of recent developments of simulation tools concerning the determination of the shielding of ionizing radiation and local dose rates in the vicinity of casks and facilities for nuclear fuel and nuclear wastes. The combination of deterministic and Monte Carlo methods in the MAVRIC code allows a simplified analysis of the complex problems in the fields of radiation transport. The capabilities of the program packages were studied using benchmark experiments from the SINDBAD and ICSBEP data bases. The quality of the radiation source term definition has a major impact on the quality of shielding calculations of nuclear matter containing casks. The source term generation includes the numerical determination of radiation inventories, and the calculation of neutron and gamma spectra. Using a detailed 3D generic cask model with MSNP 5 1.51 and MAVRIC/SCALE6.1.2 different sources of potential uncertainties in the radiation transport calculation were analyzed and quantified, including the variation of input parameters.