Moderation of neutron spectra

Most of the accelerators that produce the various microenergetic neutron sources required for low-energy neutron dosimetry studies have been shut down. One alternative to accelerator-produced sources is the use of fission neutron or (α,n) sources with unique neutron spectra. The problem with this solution is that maintenance of these sources is impractical. To help overcome this impracticality, the authors propose the use of moderating materials to produce a variety of spectra using a minimum number of sources. In the study, they performed Monte Carlo transport calculations under the following conditions: transporting neutrons from bare 252Cf or 241Am-Be sources from the center of various-sized spheres; tallying neutron spectra at 50 cm from the source. Of the twelve different moderating materials they studied, they found pure copper to be an ideal moderator. In this paper, they present flux-weighted energies, neutron spectra, and dose information for both 252Cf and 241Am-Be sources in bare and six-moderator configurations