A method of calculating the neutron source strength in irradiated fast flux test facility (FFTF) fuel has been developed. This method has been used to perform calculations in support of the reactivity monitoring of the FFTF reactor by the modified source multiplication method during refueling operations. Isotope buildup and depletion in FFTF fuel as a function of irradiation were evaluated with the ORIGEN and ALCHEMY codes using updated libraries of effective cross sections and half-lives. More accurate evaluations of isotopic density changes in fuel than previously possible were made at the Hanford Engineering Development Laboratory using the ENDF/B-V cross sections. Libraries of oneenergy-group effective cross sections for capture, fission, and (n,2n) reactions were developed by spectrum averaging 12-energy-group cross sections with typical 12-energy-group spectra in the inner and outer driver regions of FFTF Cores 1 and 2 at the beginning of life, the beginning of cycle 4, and end of cycle 4. The calculational results of isotope depletion and buildup for inner driver and outer driver fuel were used with recently evaluated neutron yields for spontaneous fission and (α,n) reactions. These provided for more accurate neutron source level evaluations. Neutron source strengths in FFTF cores 1 and 2 fuels, as a function of irradiation, were calculated and used in reactivity calculations for a sequence of core configurations representative of a typical FFTF refueling plan. The results of such calculations are presented.