Metallic uranium was placed in a paraffin pile containing two B10F3 neutron counters. The mean life of thermal neutrons in this pile was measured (140 μsec). Neutron counter pulses occurring within a time interval of this order can be assumed to come from an individual fission event and statistical analysis makes it possible to obtain information concerning the actual number of neutrons released in one fission event. The efficiency of the pile for neutron counting was found to be 2.5% using the known neutron rate from spontaneous fission of uranium. Without further assumption, the method does not give the mean multiplicity directly, but the higher moments of the multiplicity spectrum. A Poisson distribution for the multiplicity spectrum fits our data well and gives an average of 2.3 ± 0.2 neutrons per spontaneous fission. This value is in good agreement with the mean multiplicity from other methods.