A sensitivity analysis system is developed for assessing the economic implications of uncertainties in nuclear data and related computational methods for light water power reactors. Results of the sensitivity analysis indicate directions for worthwhile improvements in data and methods. Benefits from improvements in data and methods are related to reduction of margins provided by designers to ensure meeting reactor and fuel objectives. The sensitivity analysis system relates costs to uncertainties in nuclear data and methods by two sequences of operations broken at the few-group data level. The first determines the sensitivity of reactor fuel cycle cost to uncertainties in few-group microscopic cross sections. Then, for important cases, further analysis relates few-energy-group cell-averaged microscopic cross sections to uncertainties in basic nuclear data and in related computational methods. Sensitivity analyses are carried out using the batch depletion code FASTCELL, the core analysis code FASTCORE, and the reactor cost code COSTR. FASTCELL depletes a cell using methods comparable to industry cell codes except for a few-group treatment of cell flux distribution. FASTCORE is used with the Haling strategy of fixed power sharing among batches in the core. COSTR computes costs using components and techniques as in industry costing codes, except that COSTR uses fixed payment schedules. Sensitivity analyses are carried out for large commercial boiling and pressurized water reactors. Each few-group nuclear parameter is changed, and initial enrichment is also changed so as to keep the end-of-cycle core multiplication factor unchanged, i.e., to preserve cycle time at the demand power. Sensitivities of eqilibrium fuel cycle cost are determined with respect to ∼300 few-group nuclear parameters, both for a normal fuel cycle and for a throwaway fuel cycle. Particularly large dollar implications are found for thermal and resonance range cross sections in fissile and fertile materials. Sensitivities constrained by adjustment of fission neutron yield so as to preserve agreement with zero exposure integral data also are computed.