Neutron flux characterization of the beam port of the Missouri University of Science and Technology Reactor - AKYÜREK, TAYFUN (SPRINGER, 2019)
Neutron activation analysis was used to determine the thermal and epithermal neutron flux and its spatial distribution at the beam port of the Missouri University of Science and Technology Research Reactor. Gold foil irradiations were conducted and the determined flux was compared to Monte Carlo radiation transport predictions using the Monte Carlo N-particle code. The average experimental thermal (E<0.55eV) and epithermal (0.55eV<E<100keV) fluxes were (5.7 +/- 0.3)x10(6)cm(-2)s(-1) and (7.3 +/- 0.3)x10(5)cm(-2)s(-1), respectively and were found to be in good qualitative agreement with the MCNP simulations.
Keywords: Foil activation; Neutron activation analysis; Neutron flux characterization; MCNP; Monte Carlo; Beam port; SELF-SHIELDING FACTORS; CORE; PGAA