Browsing by Subject "Safeguards"
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Item Design of a Safeguards Instrument for Plutonium Quantification in an Electrochemical Refining System(2013-06-25) Le Coq, Annabelle GThere has been a strong international interest in using pyroprocessing to close the fast nuclear reactor fuel cycle and reprocess spent fuel efficiently. To commercialize pyroprocessing, safeguards technologies are required to be developed. In this research, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been investigated as a method to safeguard the process and more precisely quantify the 239Pu content of pyroprocessing materials. This method uses a detector array with different filters to isolate the low-energy resonance in 239Pu neutron fission cross section. The relative response of the different detectors allows for the quantification of the amount of 239Pu in the pyroprocessing materials. The Monte-Carlo N-Particle (MCNP) code was used to design a prototype SINRD instrument. This instrument is composed of a neutron source pod and a SINRD detector pod. Experimental measurements were also performed to validate the MCNP model of the instrument. Based on the results from simulations and experiments, it has been concluded that the MCNP model accurately represents the physics of the experiment. In addition, different SINRD signatures were compared to identify which of them are usable to determine the fissile isotope content. Comparison of different signatures allowed for reduction in the uncertainty of the 239Pu mass estimate. Using these signatures, the SINRD instrument was shown to be able to quantify the 239Pu content of unknown pyroprocessing materials suitable for safeguards usage.Item Exploration of Ion-Exchanged Glass for Seals Applications(2012-10-19) Ghanbari, RoushanAs the nuclear industry grows around the globe, it brings with it a need for more safeguards and proliferation resistant technologies. The International Atomic Energy Agency (IAEA) depends on effective containment and surveillance (C/S) technologies and methods for maintaining continuity of knowledge over nuclear assets. Tags and seals, a subset of C/S technologies, are an area where innovation has been relatively stagnant for the past fifteen years. It is necessary to investigate technologies not previously used in this field in order to defend against emerging threats and methods of defeat. Based on a gap analysis of tags and seals currently being used by the IAEA, completed with the input of several subject matter experts, the technology selected for investigation was ion-exchanged glass. Ion-exchanged glass is relatively inexpensive, has high strength, and can be used in a variety of applications. If identical pieces of glass are exchanged under the same conditions and subjected to the same point load, the fracture patterns produced can be compared and used as a verification measure. This technology has the potential to be used in passive seal applications. Each image was categorized depending on its fracture as a "3 leaf" or "4 leaf" pattern. These two populations were separately analyzed and evaluated. Several methods used to analyze the fracture patterns involve the use of image analysis software such as ImageJ and the MATLAB Control Point Selection Tool. The statistical analysis software Minitab was used to validate the use of facture pattern analysis as verification tool. The analysis yielded a 60% verified comparison for samples demonstrating a "3 leaf" fracture pattern and a 78% verified comparison for samples with a "4 leaf" fracture pattern. This preliminary analysis provides a strong indication of the plausibility for the use of ion-exchanged glass as a verification measure for C/S measures and specifically tags and seals.Item Feasibility Study of a Portable Coupled 3He Detector with LaBr3 Gamma Scintillator for Field Identification and Quantification of Nuclear Material(2010-07-14) Strohmeyer, Daniel C.In recent years, there have been several research endeavors to increase the ability to identify and quantify special nuclear material in field measurements. These have included both gamma spectroscopy and neutron coincidence systems that are portable and work in a variety of environments. In this work, a Monte Carlo Neutral Practicle X (MCNPX) model was used to design an instrument that includes four gamma detection slabs placed within four neutron detection slabs. The combination of gamma spectroscopy and neutron coincidence counting in a single instrument allows for direct measurement of plutonium (Pu) mass without need for assumptions or operator declarations. A combined neutron-gamma instrument was designed for use in characterizing and quantifying Pu in field samples. This detector consists of a plastic scintillator containing LaBr3 nanoparticles and a polyethylene slab containing four 3He tube detectors. The system was tested via simulation with MCNPX for four Pu samples of known quality and quantity. These samples had masses ranging from 100-300 g of Pu. It was found that the designed detector system could be used to determine 240Pu-effective mass to within 3.5% accuracy and to characterize the isotopic content of the Pu to within 2% accuracy for all isotopes except for 238Pu and 242Pu. The system could determine 238Pu isotopic content to within 14% accuracy but is completely unable to determine 242Pu content. This system has the ability to Four Plutonium (Pu) samples of known quantity were modeled and tested to determine what data was available from each individual signature. Each model included a separate MCNPX deck for each individual isotope that contributes to the gamma signature in photon mode and a spontaneous fission and (alpha,n) deck for the neutron signature. The first three samples were used to create spectrums and efficiency curves for each odd isotope as well as for a Pu effective mass for the neutron signature. The data from these simulations were then used to identify the isotopics in the fourth sample to within acceptable accuracy. From this data, a total Pu mass was obtained as well as an ability to determine the ratio of (alpha,n) to spontaneous fission neutrons without additional simulations. This provides a new method to detect and identify the Pu content within a sample without producing requiring supplemental additional information since isotopics can be determined with the combined use of the gamma and neutron systems.Item A game theoretic approach to nuclear safeguards selection and optimization(2013-08) Ward, Rebecca Morgan; Schneider, Erich A.This work presents a computational tool that calculates optimally efficient safeguarding strategies at and across nuclear fuel cycle facilities for a cost-constrained inspector seeking to detect a state-facilitated diversion or misuse. The tool employs a novel methodology coupling a game theoretic solver with a probabilistic simulation model of a gas centrifuge enrichment plant and an aqueous reprocessing facility. The simulation model features a suite of defender options at both facilities, based on current IAEA practices, and an analogous menu of attacker proliferation pathway options. The simulation model informs the game theoretic solver by calculating the detection probability for a given inspector-proliferator strategy pair and weighting the detection probability by the quantity and quality of material obtained to generate a scenario payoff. Using a modified fictitious play algorithm, the game iteratively calls the simulation model until the equilibrium is reached and outputs the optimal inspection strategy, proliferation strategy, and the equilibrium scenario payoff. Two types of attackers are modeled: a breakout-willing attacker, whose behavior is driven by desire for high value material; and a risk-averse attacker, who desires high-value material but will not pursue a breakout strategy that leads to certain detection. Results are presented demonstrating the sensitivity of defender strategy to budget and attacker characteristics, for an attacker known to be targeting the enrichment or reprocessing facility alone, as well as an attacker who might target either facility. The model results indicate that the optimal defender resource allocation strategy across multiple facilities hardens both facilities equitably, such that both facilities are equally unattractive targets to the attacker.Item New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles(2010-07-14) Metcalf, Richard R.The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.Item Quantitative NDA Measurements of Advanced Reprocessing Product Materials Containing U, NP, PU, and AM(2013-04-05) Goddard, BradenThe ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (?,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.