Development and Testing of Neutron Dosimetry Techniques for Accelerator-Based Irradiation Facilities

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Abstract
Integral tests have been made for 27 nuclear activation cross sections required for neutron dosimetry at accelerator-based irradiation facilities. Thirteen materials were irradiated in a well-defined geometry using neutrons produced by stopping 14- to 16-MeV deuterons in a thick beryllium target. The neutron spectrum was measured by time-of-flight spectrometry. Comparison of the activation integrals with the differential neutron spectrum provided an assessment of the nuclear data.A small fission chamber was used to monitor the irradiation, map flux gradients, and study attenuation and scattering effects. Comparison of the fission rates to fission product activations in the foil materials was also used to measure cumulative fission yields for 235U and 238U.