Title: Gamma Source Terms and Burnup Calculations for Spent Fuel
Mentor: Holly Trellue
Description: This project will primarily involve generating passive gamma ray source terms in the Monte Carlo code MCNP for a range of spent nuclear fuel assemblies and/or rods. The MCNP input files will then be used by others to simulate performance of particular Non Destructive Assay detectors of interest. In addition, irradiation simulations of different reactor types and/or fuel pins may be performed using the code Monteburns, which links MCNP to an isotope depletion/generation code such as ORIGEN-S. The results will give insight about both the isotopic compositions present in irradiated material and the gamma rays emitted from it. The work will require experience with MCNP and advanced programming skills.
Duration: 6 months if possible; shorter if needed
Number of students: 1