Neutron shielding calculation for VVER NPP

There are two methods for neutron transport (shielding) calculation used in Energoproject, Prague, the method of discrete ordinates (code TORT-DORT) and the Monte Carlo method (codes MCNP and module within the code SCALE). The task concerning neutron dose rates calculation near casks with VVER spent fuel are presented as an example. Measured neutron dose rates of real loaded C-30 casks for VVER spent fuel assemblies are compared with calculated values in the frame of the international benchmark calculation task. A part of the task realized by the Atomic Energy Research (AER) organization concerning neutron shielding is calculated. The cask C-30 is used in Slovak Jaslovske Bohunice NPP for transport of spent fuel assemblies to the storage facility. The benchmark task has been calculated by the two-dimensional code DORT originated from Oak Ridge National Laboratory. The code solves transport problems using the method of discrete ordinates (SN - method). Calculated neutron dose rates in azimuth and vertical directions show good agreement with the experiment within the range of the measurement errors. In comparison with the other codes the results of DORT are approximately 20% lower. There have been analysed differences between one- and two- dimensional approach and influence of the flux-to-dose rate conversion factors set