Browsing by Subject "plutonium"
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Item Computational Nuclear Forensics Analysis of Weapons-grade Plutonium Separated from Fuel Irradiated in a Thermal Reactor(2014-04-27) Coles, Taylor MarieThe objective of this thesis work is to utilize computational models to reliably predict the intrinsic signature of the weapons-grade plutonium separated from a Pressurized Heavy Water Reactor (PHWR), specifically an Indian 220 MWe PHWR. The PHWR produced weapons-grade plutonium due to the low-burnup seen by the fuel in the computational model. The computational modeling for this project was completed using MCNPX-2.7 radiation transport code. MCNPX-2.7 was used to perform burnup calculations for the PHWR in able to determine the resulting isotopic makeup of actinides and trace elements found in the discharged fuel. The discharged fuel of interest was a single bundle of natural uranium fuel which had undergone a burnup of about 1 GWd/tU. During the PHWR core burnup simulation, certain fuel channels were reshuffled and replaced with a number of new or "fresh" fuel bundles to simulate the process of refueling the reactor; however, it was later determined that utilizing a computational model of a single bundle with reflective boundary conditions on all sides was sufficient in producing the necessary data. That single bundle was burnt to the desired burnup and the final fuel composition of that bundle was used in the isotopic analysis. The specific fission products and actinides selected for this analysis were chosen based upon five parameters; the amount of production, half-life, activity, probability of detection, and the Plutonium Uranium Extraction Process (PUREX) decontamination factor. An uncertainty analysis associated with Monte Carlo methodology was completed using the computational model to predict the mean and standard deviation of the amount of production from the PHWR. Ratios of the selected isotopes concentrations and activities per 1 Kg of total plutonium with a decontamination factor of 106 were calculated for the PHWR. The intrinsic signature of the PHWR was also compared to that from a Fast Breeder Reactor (FBR), and a ratio of the PHWR results to the FBR results was completed to determine if noticeable differences could be seen between the two reactor types, hence, proving the existence of identifyable intrinsic physical signatures in separated weapons-grade plutonium produced by differing reactor types. Ultimately, if smuggled weapons-grade plutonium is intercepted, an analysis of isotopic signatures would be able to attribute the material back to a source reactor. The future work would include experimental data collected after single fuel pellets of natural uranium fuel have been irradiated to the desired burnup in the Oak Ridge National Laboratory- High Flux Isotope Reactor (ORNL-HFIR), and then separated using the PUREX process to experimentally determine the intrinsic signature of the fuel. The experimental data is not yet available.Item Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel(2011-10-21) Lafleur, AdrienneThe development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: 1) SINRD provides absolute measurements of burnup independent of the operator?s declaration. 2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3? from LWR spent LEU and MOX fuel. 3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. 4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. 5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.