Three dimensional radiation transport analyses in PWR with TORT and MCNP

Three dimensional (3D) neutron and gamma calculations for structural materials inside the reactor vessel in a commercial PWR were performed using the 3D transport code TORT and the Monte Carlo code MCNP to assess the accuracy of calculations using these codes and libraries. Comparisons with two dimensional DORT calculations with various libraries and surveillance dosimetry measurements indicated that TORT and MCNP calculations give similar agreements with surveillance measurements to DORT calculations. Influences of the cross section data, ENDF/B-IV, ENDF/B-VI and JENDL3.2 on attenuation of the fast flux and dpa rate in the reactor vessel, relative contributions of gamma-rays and thermal neutrons to dpa were discussed.