Graduate Thesis Or Dissertation

MCNP 6 Simulations of Short-Lived, Low-Yield Fission Byproducts, Gamma-Ray Spectrometry, and their Comparison to Natural Uranium Measurements for the Evaluation of Available Evaluated Nuclear Data Files

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  • Natural uranium samples were irradiated in-core for 7 seconds using the pneumatic rabbit system within the TRIGA Mark II Reactor at Oregon State University, and then used to generate a gamma-ray spectrum with a high purity germanium (HPGe) detector. The irradiated sample was simulated in Monte Carlo N-Particle (MCNP), a radiation transport code developed to simulate fission, or burnup, of special nuclear material for application in the field of nuclear forensics, to replicate the experiment, and a simulated gamma ray spectrum was generated within MCNP for this sample. The results of these experiments were compared to each other in order to highlight inconsistencies between available evaluated nuclear data and expected physical data. There were limitations discovered within MCNP that restrict the viability of using MCNP for replicating gamma ray spectra produced by fission byproducts, due to lack of available data for short-lived, low-yield fission byproducts. Further work may be performed to experimentally verify and/or correct the current available data to reflect actual fission yields, in order to improve the reliability of MCNP for use in repeating characteristic analysis of special nuclear material.
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