Design of a semiportable shielding for a Cf-252 neutron source using MCNP-5
A 10-µg Cf-252 spontaneous fission neutron source has been procured for use as a standard source for neutron dosimetry and for replacement of a depleted Cf-252. A radiation shielding container was designed and optimized using MCNP-5 radiation transport code to increase safety and versatility of transportation of the new source. Shielding materials considered for the model were acrylic, concrete, paraffin (plain and 5% borated), polyethylene, and water. The simulation results using F4 tally show that the use of borated paraffin resulted in the highest decrease in overall radiation dose rate. Using borated paraffin as the shielding material, a 3D dose rate mesh across the container was generated by FMESH4 tally. The overall design and the dose rate results show that the container can be used for both short distance transportation and temporary source storage.
By submitting their manuscript to the Samahang Pisika ng Pilipinas (SPP) for consideration, the Authors warrant that their work is original, does not infringe on existing copyrights, and is not under active consideration for publication elsewhere.
Upon acceptance of their manuscript, the Authors further agree to grant SPP the non-exclusive, worldwide, and royalty-free rights to record, edit, copy, reproduce, publish, distribute, and use all or part of the manuscript for any purpose, in any media now existing or developed in the future, either individually or as part of a collection.
All other associated economic and moral rights as granted by the Intellectual Property Code of the Philippines are maintained by the Authors.