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.

The 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 3σ 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.

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. / text

The effects of global warming raise demands for reduced CO2 emissions, whereas at the same time the world’s need for energy increases. With the aim to resolve some of the difficulties facing today’s nuclear power, striving for safety, sustainability and waste minimization, a new generation of nuclear energy systems is being pursued: Generation IV. New reactor concepts and new nuclear facilities should be at least as resistant to diversion of nuclear material for weapons production, as were the previous ones. However, the emerging generation of nuclear power will give rise to new challenges to the international safeguards community, due to new and increased flows of nuclear material in the nuclear fuel cycle. Before a wide implementation of Generation IV nuclear power facilities takes place, there lies still an opportunity to formulate safeguards requirements for the next generation of nuclear energy systems. In this context, this thesis constitutes one contribution to the global efforts to make future nuclear energy systems increasingly resistant to nuclear material diversion attempts. This thesis comprises three papers, all of which concern safeguards and proliferation resistance in Generation IV nuclear energy systems and especially recycling facilities: In Paper I, proliferation resistances of three fuel cycles, comprising different reprocessing techniques, are investigated. The results highlight the importance of making group actinide extraction techniques commercial, due to the inherently less vulnerable isotopic and radiological properties of the materials in such processes. Paper II covers the schematic design and safeguards instrumentation of a Generation IV recycling facility. The identification of the safeguards needs of planned facilities can act as a guide towards the development of new instrumentation suitable for Generation IV nuclear energy systems. Finally, Paper III describes a mode of procedure for assessing proliferation resistance of a recycling facility for fast reactor fuel. The assessments may be used, as in this case, as an aid to maintain or increase the inherent proliferation resistance when performing facility design changes and upgrades.

Nuclear safeguards are intrinsic and extrinsic features of a facility which reduce probability of the successful acquisition of special nuclear material (SNM) by hostile actors. Future bulk handling facilities in the United States will include both domestic and international safeguards as part of a voluntary agreement with the International Atomic Energy Agency. A new framework for safeguards, the Safeguards Envelope Methodology, is presented. A safeguards envelope is a set of operational and safeguards parameters that define a range, or “envelope,” of operating conditions that increases confidence as to the location and assay of nuclear material without increasing costs from security or safety. Facilities operating within safeguards envelopes developed by this methodology will operate with a higher confidence, a lower false alarm rate, and reduced safeguards impact on the operator. Creating a safeguards envelope requires bringing together security, safety, and safeguards best practices. This methodology is applied to an example facility, the Idaho Chemical Processing Plant. An example diversion scenario in the front-end of this nuclear reprocessing facility, using actual operating data, shows that the diversion could have been detected more easily by changing operational parameters, and these changed operational parameters would not sacrifice the operational efficiency of the facility, introduce security vulnerabilities, or create a safety hazard.

The PGET device is currently being developed for partial-defect verication purposes on nuclear fuel assemblies. It Comprises CdTe detector elements in a heavy tungsten-alloy collimator, for which collimator slit openings define the field-of-view. This study aims at calculating the spatial response function of this device for further deployment in tomographic reconstruction algorithms. In this work, the detector response for 2 dierent sources (662 keV from Cesium-137 and 1274 keV from Europium-154) was simulated using the MCNPX software package. In the simulations, energy windows used in measurements with the PGET device were deployed. The results show the expected characteristics with strong response for a source position directly in front of the collimator slit opening and decreasing response as the source is moved into the penumbra and umbra region. The uncertainty of the simulated response function was less than 3.5 % for both sources. Separate simulations were made to quantify contributions from septal penetration and scattering from the collimator material into the detector for the energy windows covering the full -energy peak. These contributions were found to be around3% for the source of Cesium-137 and 6% for the source of Europium-154.

Today's schools face a mounting number of court cases resulting from conflicts between parents of children with special needs and educators tasked with meeting those needs (Osborne, 2009). Principals have the enormous responsibility to ensure appropriate services to educate students with disabilities and, as special education leaders, require a skill set that includes knowledge of current laws, litigation, student learning needs, and how to support parents' decision making rights and responsibilities. A gap is evident between what principals know about special education leadership and case law and what principals are doing in the field. The purpose of this study was to identify effective actions and behaviors that support Virginia principals' leadership in special education decision making. More specifically, the study examined what can be done to minimize special education disputes between parents and schools and identify principals' skill sets to minimize special education disputes. Two concurrent Delphi studies were conducted with 16 member panels; stakeholders with familial responsibilities to children with disabilities and professional experts with responsibility to special education compliance participated. A final round exchanged findings between the panels. The study identified a list of best practices for Virginia school principals to support special education leadership and decision making. / Ed. D.

The Digital Cherenkov Viewing Device (DCVD) is an instrument used by authority inspectors to assess irradiated nuclear fuel assemblies in wet storage for the purpose of nuclear safeguards. Originally developed to verify the presence of fuel assemblies with long cooling times and low burnup, the DCVD accuracy is sufficient for partial defect verification, where one verifies that part of an assembly has not been diverted. Much of the recent research regarding the DCVD has been focused on improving its partial defect detection capabilities. The partial-defect analysis procedure currently used relies on comparisons between a predicted Cherenkov light intensity and the intensity measured with the DCVD. Enhanced prediction capabilities may thus lead to enhanced verification capabilities. Since the currently used prediction model is based on rudimentary correlations between the Cherenkov light intensity and the burnup and cooling time of the fuel assembly, there are reasons to develop alternative models taking more details into account to more accurately predict the Cherenkov light intensity. This work aims at increasing our understanding of the physical processes leading to the Cherenkov light production in irradiated nuclear fuel assemblies in water. This has been investigated through simulations, which in the future are planned to be complemented with measurements. The simulations performed reveal that the Cherenkov light production depends on fuel rod dimensions, source distribution in the rod and initial decay energy in a complex way, and that all these factors should be modelled to accurately predict the light intensity. The simulations also reveal that for long-cooled fuel, Y-90 beta-decays may contribute noticeably to the Cherenkov light intensity, a contribution which has not been considered before. A prediction model has been developed in this work taking fuel irradiation history, fuel geometry and Y-90 beta-decay into account. These predictions are more detailed than the predictions based on the currently used prediction model. The predictions with the new model can be done quickly enough that the method can be used in the field. The new model has been used during one verification campaign, and showed superior performance to the currently used prediction model. Using the currently used model for this verification, the difference between measured and predicted intensity had a standard deviation of 15.4% of the measured value, and using the new model this was reduced to 8.4%.

Europeanization of Criminal Law This thesis deals with Europeanization of Criminal Law. The topic itself is broad, selected issues of Europeanization of Criminal Procedural Law are therefore emphasized. In the introduction, certain problems which accompany the Europeanization in a Criminal Law field are presented. The biggest obstacle is a close connection of Criminal Law with state sovereignty, which is something that states are reluctant to restrict in favour of European Union. Another hindrance to Europeanization is a difference among national criminal regulations, which make an achievement of a compromise regarding the harmonization harder. The first chapter concentrates on the terms Europeanization of Criminal Law, European Criminal Law and Criminal Law of the European Union. Their definition and differentiation is provided. The second chapter discusses in brief the evolution of Europeanization of Criminal Law prior to the adoption of Schengen treaties. The informal cooperation in criminal matters took place in this era. Unlike the one in the chapter three, which already addresses the formal cooperation in criminal matters. It describes gradual development from Schengen cooperation, through the cooperation under Maastricht, Amsterdam and Nice Treaty, up to the cooperation on the basis of...

Two important issues related to nuclear materials safeguards are the continuous monitoring of nuclear processing facilities to verify that undeclared uranium is not processed or enriched and to verify that declared uranium is accounted for. The International Atomic Energy Agency (IAEA) is tasked with ensuring special nuclear facilities are operating as declared and that proper material safeguards have been followed. Traditional safeguards measures have relied on IAEA personnel inspecting each facility and verifying material with authenticated instrumentation. In newer facilities most plant instrumentation data are collected electronically and stored in a central computer. Facilities collect this information for a variety of reasons, most notably for process optimization and monitoring. The field of process monitoring has grown significantly over the past decades, and techniques have been developed to detect and identify changes and to improve reliability and safety. Several of these techniques can also be applied to international and domestic safeguards. This dissertation introduces a safeguards monitoring system developed for both a simulated Uranium blend down facility, and a water-processing facility at the Oak Ridge National Laboratory. For the simulated facility, a safeguards monitoring system is developed using an Auto-Associative Kernel Regression model, and the effects of incorporating facility specific radiation sensors and preprocessing the data are examined. The best safeguards model was able to detect diversions as small as 1.1%. For the ORNL facility, a load cell monitoring system was developed. This monitoring system provides an inspector with an efficient way to identify undeclared activity and to identify atypical facility operation, included diversions as small as 0.1 kg. The system also provides a foundation for an on-line safeguards monitoring approach where inspectors remotely facility data to draw safeguards conclusion, possibly reducing the needed frequency and duration of a traditional inspection.