Abstract: The reactor core hot spot factor is an important parameter for thermal hydraulics design and safety analysis. Based on nodal expansion method used for solving neutron diffusion equations in hexagonal-z geometry, the orthogonal polynomial method was used to approximate to the neutron flux density distribution and the pin power distribution. Considering the features of hexagonal fuel subassemblies used in fast reactor, the small hexagonal integral method was used to calculate power of each pin. Based on NAS code, NAS-PIN module was encoded for calculation of pin power distribution. Comparing with the MCNP calculation results, the NAS-PIN calculation results are believable and can meet requirements for engineering design.