Neutron activation analysis with PE moderated252Cf neutron source

Abstract

The 1.369 MeV γ-rays emitted from24Na→24Mg+β−+γ after23Na(n,γ)24Na reaction were counted by high purity germanium (HPGe) detector and the half-life of24Na was derived. This process was simulated by Monte Carlo Neutron and Photon Transport Code (MCNP-4A).252Cf neutron source was moderated by a polyethylene (PE) cylinder to increase the cross section of neutron absorption reaction and to decrease the biological hazard. NaCl powder of 20 cm3 and 40 cm3 volume in cylindrical polypropylene capsules were irradiated by the neutrons passed through the moderator. MCNP-4A was used to determine the optimum size of PE moderator, to assume the realistic geometry of the HPGe detector, and to assume the absolute efficiencies of the detector. The count rates for 1.369 MeV γ-rays in the HPGe detector were calculated by MCNP-4A for 20 cm3 and 40 cm3 NaCl. The accumulated counts calculated tumed out to be higher than those actually measured by 31% with a relative error of 3%. The half-life of24Na measured within 4% and 1% for 20 cm3 and 40cm3 NaCl agrees with that of the reference. So, we can say that the result of MCNP-4A has about 30% of accuracy and 3% of precision in simulating the neutron activation analysis.