As many French nuclear 900 MWe Pressurized Water Reactors (PWRs) approach the end of their design lifetime (40 years), it is of great importance, in the perspective of a possible extension of operation, to be able to accurately characterize the structural integrity of the reactor pressure vessel for safety reasons. Therefore, the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) consider the possibility to develop a best-estimate calculation scheme for the fast neutron fluence (neutrons with energies greater than 1 MeV) at the vessel of PWRs, which is used to predict the material embrittlement. However, before building this type of model, one prior key point is to determine which parameters need to be considered. The aim of this paper is to assess the need to take into account the power history of the reactor core and the induced variation of its operational parameters (core power, boron concentration and control rods insertion) in order to perform fast neutron fluence calculations. The analysis has been performed using a calculation scheme coupling a deterministic approach (CASMO5 and PARCS codes) to evaluate the core fission distribution and a Monte-Carlo modeling (MCNP6 code) to estimate the attenuation of the neutrons source from the core to the vessel.