The estimation of the neutron fluence at the Reactor Pressure Vessel (RPV) is classically carried out by a two-step approach. The first step is to estimate the full core neutron source term whether the second step of the calculation consists in the transport of neutrons from the core (source term) to the RPV using the neutron fission distribution determined in the previous step. For this purpose, the neutron fission distribution is to be accurately determined at the fuel pin level for the assemblies on the border of the core. To achieve this goal, two methods are evaluated in this study. The first method considered is a full core 2D Monte Carlo calculation using the MNCP6 code. The second method is based on a deterministic approach using the CASMO5 multi-segment option, allowing a full 2D transport calculation at the pin level with an expected accuracy similar to a stochastic method. The comparison of the two methods shows an overall good agreement with differences within the statistical uncertainty for different cores: homogeneous UOX core, mixed UOX-MOX loading and the effect of the hafnium rods used in the assemblies in the periphery of the core. The modelling limitation and the associated calculational time are discussed for the comparison of the two approaches.