Analysis of a neutron scattering and gamma-ray production integral experiment on aluminum for neutron energies from 1 to 15 MeV

Technical Report · Sun Jan 01 04:00:00 EST 1978

DOI:https://doi.org/10.2172/5183366· OSTI ID:5183366

Cramer, S. N.; Oblow, E. M.

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.

• Research Organization: Oak Ridge National Lab., Tenn. (USA)
• Sponsoring Organization: USDOE
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 5183366
• Report Number(s): ORNL/TM-6146
• Country of Publication: United States
• Language: English

Related Subjects

654003* -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ALUMINIUM
CROSS SECTIONS
DATA ANALYSIS
ELEMENTS
GAMMA SPECTRA
METALS
MONTE CARLO METHOD
NEUTRAL-PARTICLE TRANSPORT
NEUTRON SPECTRA
NEUTRON TRANSPORT
RADIATION TRANSPORT
SCATTERING
SPECTRA