The neutron absorption and fission cross sections for 239Pu and 235U have been measured over the neutron energy range from 0.02 eV to 200 keV. In addition, the neutron capture cross section for 197Au was measured from 10 to 50 keV and the fission cross section of 233U was measured from 0.1 to 100 keV. Normalization of the 239Pu and 235U data was made over the energy region from 0.02 to 0.4 eV to the ENDF/B-III neutron cross sections for these isotopes, Mat 1159 and 1157, respectively. The capture cross section for 197Au was normalized using the saturated resonance method for the 4.9-eV resonance. For 233U fission, the normalization was made using the results of Weston et al. The neutron flux was measured using the 10B(n,α) reaction; the energy variation used for this reaction was that given in ENDF/B-III. The pulsed-neutron beam for these measurements was generated using the Oak Ridge Electron Linear Accelerator. A large liquid scintillator about 40 m from the neutron source was used to detect the prompt gamma-ray cascades resulting from neutron absorption in the sample. The time interval between the burst of neutrons and the detection of the absorption event was used to establish the neutron energy scale. The sample of the fissile isotopes was contained in multiplate (pulse) ionization chambers and those neutron absorption events detected in coincidence with a pulse from the ionization chamber were defined as fission events. In general for 239Pu and 235U, these experiments indicated lower neutron fission cross sections than contained in ENDF/B-III for energies above 10 keV. The measured values of the ratio α, neutron capture-to-neutron fission, for 239Pu agree within errors with those derived from ENDF/B-III, Mat 1159. For the present measurements, the uncertainty on α for 239Pu is ∼11% at 10 keV and increases to ∼30% at 100 keV. The experimental results for the neutron capture cross section for 197Au are ∼15% lower than the ENDF/B-III values. The measurements of the ratio of the neutron fission cross section for 233U to that for 235U are generally higher than the ENDF/B-III values by ∼5%.