Modeling the response of a Harshaw albedo neutron dosimeter using the Monte Carlo n-particle transport code, MCNP

Currently Oak Ridge National Laboratory utilizes the Harshaw albedo neutron dosimeter to estimate doses resulting from exposure to neutron fields. The characterization of this dosimeter's neutron response is primarily based on Alsmiller's and Barish's work that characterizes a Hoy albedo neutron dosimeter that utilizes ⁶LiF and ^{7LiF like the Harshaw dosimeter. However, the Hoy dosimeter's geometry and chemical constituents are very different from the Harshaw albedo neutron dosimeter which can significantly change the dosimeter response. Therefore, this project was instigated for the purpose of characterizing the relative neutron response curve of the Harshaw albedo neutron dosimeter using the MCNP Transport Code. This is achieved by modeling the Harshaw dosimeter and simulating the transport and subsequent interactions of neutrons being emitted by a neutron source parallel to the dosimeter. The results of forty-nine runs are used to characterize the response of the dosimeter to energies ranging from thermal (0.025 keV) to 20 MeV. As expected, the Harshaw's response curve shows the dosimeter being most responsive to thermal neutrons and decreasing in its responsiveness as the neutron energy is increases. These results are similar to those of Alsmiller and Barish as compared to the Harshaw dosimeter. Also, this study is significant in that it validates the usefulness of MCNP to model neutron dosimetry.}