Nondestructive fissile isotopic measurement technique for /sup 233/U--/sup 235/U fuels using prompt and delayed fission--neutron counting [electronic resource].
- Oak Ridge, Tenn. : Oak Ridge National Laboratory, 1979.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy.
- Additional Creators:
- Oak Ridge National Laboratory
United States. Department of Energy. Office of Scientific and Technical Information
- Assay and analysis procedures were developed for nondestructive fissile isotopic measurement of mixed /sup 233/U--/sup 235/U fuel samples. For /sup 233/U much fewer delayed neutrons are released per fission than for /sup 235/U, 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 /sup 233/U and /sup 235/U present in the sample can be estimated. Equations of delayed- and prompt-neutron counts versus /sup 233/U and /sup 235/U contents are solved simultaneously for the /sup 233/U and /sup 235/U contents of a sample. Eleven samples containing mixtures of /sup 233/U and /sup 235/U from no /sup 233/U to nearly 100% were prepared and assayed in prompt- and delayed-neutron assay devices. Constants for calibration equations, which are derived in this report, were fitted to data from nine of the samples. The maximum differences between counts calculated by the calibration equations and measured counts were 2.3% for delayed neutrons and 1.2% for prompt neutrons, indicating a good selection of the form for the calibration equations. The two remaining samples were treated as unknowns, and the uranium contents of these samples were estimated by simultaneously solving the two calibration equations. The maximum difference between estimated /sup 233/U or /sup 235/U content and actual content for either sample was 1.5%.
- Published through SciTech Connect.
Allen, E.J.; McNeany, S.R.
- Funding Information:
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