Nondestructive fissile isotopic measurement technique for uranium-233-uranium-235 fuels using prompt and delayed fission neutron counting

Assay and analysis procedures were developed for nondestructive fissile isotopic measurement of mixed 233U-235U fuel samples. For 233U, the number of delayed neutrons released per fission is about half that for 235U, although the number of prompt neutrons is approximately the same. By separately counting prompt and delayed neutrons released by a sample exposed to neutron irradiation, the amounts of 233U and 235U present in the sample can be determined. Equations of delayed and prompt neutron counts versus 233U and 235U contents are solved simultaneously for the 233U and 235U contents of a sample