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Measurement of the Uranium-235 Content in a Spent Mtr Type Fuel Element Using the Delayed-Neutron Yield Technique

R. H. Augustson, C. N. Henry, C. R. Weisbin

Nuclear Technology / Volume 14 / Number 2 / May 1972 / Pages 197-199

Technical Note / Analysis / dx.doi.org/10.13182/NT72-A31136

The 235U content of a spent MTR fuel element enclosed in a lead transfer cask has been nondestructively measured by observing the delayed-neutron emissions following high energy neutron induced fission. Fourteen MeV neutrons from a (D,T) neutron generator moderated through 8 in. of lead were used as the interrogating radiation. A DTF-IV calculation was used to correct the data for the contributions due to other fissionable isotopes. The nondestructive measurement agreed to within 1.5% with the 235U content as predicted by reactor burnup calculations.