Date of Award


Document Type


Degree Name

Master of Science (MS)


Environmental Engineering and Earth Sciences

Committee Member

Dr. Nicole E. Martinez, Committee Chair

Committee Member

Dr. Timothy A. DeVol

Committee Member

Dr. Mark A. Blenner


Determination of neutron dose can be challenging and requires knowledge of neutron energy and neutron flux. A plutonium-239/beryllium (239PuBe) alpha-neutron source was used to irradiate bacterial samples to create neutron dose response. The goal of this project was to characterize the thermal neutron flux of the 239PuBe alpha-neutron source and model the neutron dose using version MCNPX of the Monte-Carlo N-Particle transport codes. The 37 GBq 239PuBe alpha-neutron source was placed in a neutron “howitzer,” that is, a 2-ft diameter moderating barrel with four radial irradiation ports. Multi-foil activation was used at various distances to determine thermal neutron flux, which was then used to verify a MCNPX code representing the system. Dysprosium thermal foils were used with cadmium covers. The MCNPX code was then adapted for dosimetric modeling. That is, the F5 tally, with a dose function, was used in place of the F4 tally. The four irradiation ports were found to have average thermal neutron fluxes of 5334 ± 829, 2928 ± 451, 1289 ± 199, and 1211 ± 186 neutrons cm-2 s-1 at 3.58, 9.04, 12.8, and 13.7 cm from the 239PuBe alpha-neutron source, respectively. The adapted MCNPX code calculated theoretical total ambient dose equivalent rates of 1717 ± 90.2, 703 ± 37.0, 286 ± 15.0, and 174 ± 9.18 mrem hr-1 at 4, 8, 14, and 18 cm from the 239PuBe alpha-neutron source, respectively. The theoretical direct (uncollided) ambient dose equivalent rates at the same distances were 837 ± 44.0, 272 ± 14.3, 100 ± 5.29, and 63.1 ± 3.32 mrem hr-1, respectively. Rough estimates of the absorbed dose rates were made from the ambient dose equivalent rates and a recommendation of 23.6 cm from the PuBe source was made to achieve an absorbed dose rate of roughly 10 mGy d-1.



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