Abstract
Monte Carlo calculations were made to analyze the results of an integral experiment with an aluminum sample to determine the adequacy of ENDF/B-IV neutron scattering and gamma-ray production cross-section data for aluminum. The experimental results analyzed included energy-dependent NE-213 detector neutron and gamma-ray count rates at a scattering angle of 125 deg and pulse-height spectra for scattered neutrons and gamma-rays. The experiments were carried out with the ORELA 1- to 20-MeV pulsed neutron source. The pulse-height data were unfolded to generate secondary neutron and gamma-ray spectra at 125 deg as a function of incident neutron energy. Multigroup Monte Carlo calculations using the MORSE code and ENDF/B-IV cross sections were made to analyze all reported results. Discrepancies between calculated and measured responses were found for secondary neutron scattering data above 10 MeV and for gamma-rays produced at energies between 4 and 7 MeV. A detailed analysis has not yet been performed to determine the reasons for these discrepancies.
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