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This document summarizes the flow-down of requirements from TFR-483 and SAR-419 and provides a roadmap for a verification of the requirements. The project vital safety systems include the vault shield plugs (VSP) and the cask-to-vault adapting structure (CVAS).
Abstract not provided.
As part of the regular nuclear material accounting and control process, sites that have accountable nuclear material regularly perform measurements of the material in their possession. The expectation for the type of measurements and their frequency is set out in Department of Energy (DOE) Order 474.2, and further detailed in each site’s respective material control and accountability (MC&A) plan and associated controlling documents. The measurements of interest in this report are those performed on external (between DOE sites) transfer receipts of Category III or IV accountable materials. In 2020, a revision of DOE Order 474.2 was proposed that would significantly increase the requirement that measurements be performed upon receipt of material, and further, that those measurements be quantitative verification (rather than qualitative confirmatory) measurements. A survey was performed of DOE sites to discuss the impact on their MC&A measurements program if more rigorous verification measurements requirements for Category III and IV receipts were implemented. An assessment of nondestructive assay (NDA) measurement instruments to meet those needs was also performed. Across the board, sites indicated that personnel and budget would cause the largest impact. Sites which are qualified as Category I/II indicated a reduced impact as these sites already have the measurement capabilities and some personnel but would require the acquisition of additional measurement systems. Category III/IV sites indicated a much larger impact as these sites generally have smaller MC&A programs. A few smaller sites stated a measurement program would need to be created to handle the demand. Calorimetry, quantitative gamma spectroscopy and neutron coincidence counting systems were identified as common systems that could meet the verification measurement requirements. The other, and perhaps larger issue is the need for calibration standards for the calorimetry and neutron coincidence counting systems. It is proposed that a comparative study be performed for ISOCS and ISOTOPICS, qualification of a COTS calorimetry system be performed, and development of an AWCC system that uses a neutron generator.
Safety limits are required to maintain the integrity of physical barriers designed to prevent the uncontrolled release of radioactive materials in nuclear power plants. The safety analysis establishes two critical constraints that include an analytical limit in terms of a measured or calculated variable, and a specific time after the analytical limit is reached to begin protective action. Keeping with the nuclear regulations and industry standards, satisfying these two requirements will ensure that the safety limit will not be exceeded during the design basis event, either an anticipated operational occurrence or a postulated accident. Various studies on the setpoint determination methodology for the safety-related instrumentation have been actively performed to ensure that the requirement of the analytical limit is satisfied. In particular, the protection setpoint methodology for the advanced power reactor 1400 (APP1400) and the optimized power reactor 1000 (OPR1000) has been recently developed to cover both the design basis event and the beyond design basis event. The developed setpoint methodology has also been quantitatively validated using specific computer programs and setpoint calculations. However, the safety of nuclear power plants cannot be fully guaranteed by satisfying the requirement of the analytical limit. In spite of the response time verification requirements of nuclear regulations and industry standards, it is hard to find the studies on the systematically integrated methodology regarding the response time evaluation. In cases of APR1400 and OPR1000, the response time analysis for the plant protection system is partially included in the setpoint calculation and the response time test is separately performed via the specific plant procedure. The test technique has a drawback which is the difficulty to demonstrate completeness of timing test. The analysis technique has also a demerit of resulting in extreme times that not actually possible. Thus, the establishment of the systematic response time evaluation methodology is needed to justify the conformance to the response time requirement used in the safety analysis. This paper proposes the response time evaluation methodology for APR1400 and OPR1000 using the combined analysis and test technique to confirm that the plant protection system can meet the analytical response time assumed in the safety analysis. In addition, the results of the quantitative evaluation performed for APR1400 and OPR1000 are presented in this paper. The proposed response time analysis technique consists of defining the response time requirement, determining the critical signal path for the trip parameter, allocating individual response time to each component on the signal path, and analyzing the total response time for the trip parameter, and demonstrates that the total analyzed response time does not exceed the response time requirement. The proposed response time test technique is composed of defining the response time requirement, determining the critical signal path for the trip parameter, determining the test method for each component on the signal path, performing the response time test, and demonstrates that the total test result does not exceed the response time requirement. The total response time should be tested in a single test that covers from the sensor to the final actuation device on the instrument channel. When the total channel is not tested in a single test, separate tests on groups of components or single components including the total instrument channel shall be combined to verify the total channel response. For APR1400 and OPR1000, the ramp test technique is used for the pressure and differential pressure transmitters and the step function testing technique is applied to the signal processing equipment and final actuation device. As a result, it can be demonstrated that the response time requirement is satisfied by the combined analysis and test technique. Therefore, the proposed methodology in this paper plays a crucial role in guaranteeing the safety of the nuclear power plants systematically satisfying one of two critical requirements from the safety analysis. (authors)
The Light Initiated High Explosive (LIHE) facility performs high rigor, high consequence impulse testing for the nuclear weapons (NW) community. To support the facility mission, LIHE's extensive data acquisition system (DAS) is comprised of several discrete components as well as a fully integrated system. Due to the high consequence and high rigor of the testing performed at LIHE, a measurement assurance plan (MAP) was developed in collaboration with NW system customers to meet their data quality needs and to provide assurance of the robustness of the LIHE DAS. While individual components of the DAS have been calibrated by the SNL Primary Standards Laboratory (PSL), the integrated nature of this complex system requires verification of the complete system, from end-to-end. This measurement assurance plan (MAP) report documents the results of verification and validation procedures used to ensure that the data quality meets customer requirements.
Robotic, in-pipe, nondestructive assay (NDA) is a revolutionary and important means for measuring, analyzing, and reporting radioactive holdup deposits in enrichment piping. This innovation promises vast advantages for efficiency, safety, and cost savings. Payoff, however, lies in utilization much more than innovation. Crossing that threshold from a technical capability to an everyday tool occurs only after rigorous commissioning, adoption, and infusion. This must meet the highest standard, and there can be no compromise. Certification of a new NDA technology must meet the highest standards for quality, safety, and efficacy. Prior to this work, there was no precedent of procedure and method for certification of a robotic methodology for this purpose. The best of robotics had not certified NDA. The best of NDA had not certified robotics. Hence, a great deal of discovery, resourcefulness, and collaboration contributed to this success. This paper chronicles the first-of-kind chartering, co-development, testing, documentation, and site integration that certified this trailblazing in-pipe, robotic NDA. Commissioning had to demonstrate and certify end-to-end NDA automation from robot traversal to computer analysis to report generation of contaminated process piping at the Portsmouth, Ohio gaseous diffusion facility. This was a rigorous campaign of tests, evaluations, and documentation. Highlights included multiple tests in four elevated pipes exercised the gamut of NDA and robotic capabilities. Each of the four test pipes presented a unique loading of U-235. The tests demonstrated operations in both 30- inch and 42-inch diameter pipes with varying lengths and terminated by a variety of valves and fittings. Tests in a fifth pipe at floor level demonstrated deployment and applicability for pipes at this ground height and robot recognition of another fitting type. All operations, calibration, deployment, auto-reporting, and analysis were handled by Portsmouth site personnel. These tests concluded a series of component-level acceptance tests which certified all constituent technologies and operational features. Test metrics included comparison of reported to ground truth results and assessment of data quality from quality assurance and replicate measurement comparison. Beyond the radiometry associated with any NDA method, this unique commissioning path exercised and evaluated all the robotic features that are intrinsic to automation. These include odometry, geometric modeling, autonomy, remote launch and recovery, and end-to-end data flow. Each of these required verification and certification. These evaluations were first-of-kind for the decommissioning community, requiring innovation and resourcefulness in their own right. (authors)
Residual stresses induced during forging and welding can cause detrimental failure in reservoirs due to enhanced possibility of crack propagation. Therefore, reservoirs must be designed with yield strengths in a tight range. This report summarizes an effort to verify and validate a computational tool that was developed to aid in prediction of the evolution of residual stresses throughout the manufacturing process. The application requirements are identified and summarized in the context of the Predictive Capability Maturity Model (PCMM). The phenomena of interest that the model attempts to capture are discussed and prioritized using the Phenomena Identification and Ranking Table (PIRT) to identify any gaps in our approach. The fidelity of the modeling approach is outlined and details on the implementation and boundary conditions are provided. The code verification requirements are discussed and solution verification is performed, including a mesh convergence study on the series of modeling steps (forging, machining and welding). Validation activities are summarized, including validation of the displacements, residual stresses, recrystallization, yield strength and thermal history. A sensitivity analysis and uncertainty quantification are also performed to understand how variations in the manufacturing process affect the residual stresses.
This document specifies the acceptance criteria for items fabricated by Holtec International under the scope of INL Contract # 215725, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Verification is confirmed with initials (or the need for remediation may be identified) along with the date of inspection for each requirement. Verification may be performed by independent proxy (University of South Carolina participant, under INL Contract Number 207046) or by an INL employee, remotely by camera images.
The U.S. Department of Energy (DOE) technical standard DOE-STD-3020-2015 requires verification testing at the Filter Test Facility (FTF) of high-efficiency particulate air (HEPA) filters to be installed in nuclear facility (Hazard Category 1, 2, 3, and radiological facilities) confinement ventilation systems, or to be installed in habitability systems (e.g., filters for protection of workers required to act for the control and mitigation of emergency situations). Filters in these applications perform a safety function in accident situations, are designated as important to safety, or are necessary for habitability systems. Different FTFs have operated for decades, and results of tests are available to examine the need for and benefits of continued execution of 100 percent QA verification testing. The objective of this study and resulting report was to assess the need for 100 percent QA testing at the current FTF in Maryland, based on an examination of a database of test results from 2013 to August 2020. Filters that require independent testing per DOE-STD-3020-2015 to be installed in nuclear facilities are sent to the FTF before being shipped to the facilities. The FTF examines the documentation (e.g., certificates of compliance, results of particle penetration and airflow resistance testing by the manufacturers, labels affixed to filters and packages, and other requirements specified in purchase orders by the nuclear facility operators), implements visual inspections of the physical state of the filters, and checks the dimensions of frames. Filters that meet the documentation and visual inspection requirements are subjected to particle penetration and airflow resistance testing. In case of rejections, the filters are shipped back to the manufacturers, at the cost of the manufacturers, for replacement. Filters that pass all tests are shipped to the nuclear facilities.
Arms control treaties are necessary to reduce the large stockpiles of the nuclear weapons that constitute one of the biggest dangers to the world. However, an impactful treaty hinges on effective inspection exercises to verify the participants’ compliance to the treaty terms. Such procedures would require verification of the authenticity of a warhead undergoing dismantlement. Previously proposed solutions lacked the combination of isotopic sensitivity and information security. Here we present the experimental feasibility proof of a technique that uses neutron induced nuclear resonances and is sensitive to the combination of isotopics and geometry. The information is physically encrypted to prevent the leakage of sensitive information. Our approach can significantly increase the trustworthiness of future arms control treaties while expanding their scope to include the verified dismantlement of nuclear warheads themselves.
Purpose: Radiotherapy (RT) contours delineated either manually or semiautomatically require verification before clinical usage. Manual evaluation is very time consuming. A new integrated software tool using supervised pattern contour recognition was thus developed to facilitate this process. Methods: The contouring tool was developed using an object-oriented programming language C# and application programming interfaces, e.g. visualization toolkit (VTK). The C# language served as the tool design basis. The Accord.Net scientific computing libraries were utilized for the required statistical data processing and pattern recognition, while the VTK was used to build and render 3-D mesh models from critical RT structures in real-time and 360° visualization. Principal component analysis (PCA) was used for system self-updating geometry variations of normal structures based on physician-approved RT contours as a training dataset. The inhouse design of supervised PCA-based contour recognition method was used for automatically evaluating contour normality/abnormality. The function for reporting the contour evaluation results was implemented by using C# and Windows Form Designer. Results: The software input was RT simulation images and RT structures from commercial clinical treatment planning systems. Several abilities were demonstrated: automatic assessment of RT contours, file loading/saving of various modality medical images and RT contours, and generation/visualization of 3-D images and anatomical models. Moreover, it supported the 360° rendering of the RT structures in a multi-slice view, which allows physicians to visually check and edit abnormally contoured structures. Conclusion: This new software integrates the supervised learning framework with image processing and graphical visualization modules for RT contour verification. This tool has great potential for facilitating treatment planning with the assistance of an automatic contour evaluation module in avoiding unnecessary manual verification for physicians/dosimetrists. In addition, its nature as a compact and stand-alone tool allows for future extensibility to include additional functions for physicians’ clinical needs.
At Lawrence Livermore National Laboratory in the National Ignition Facility Optics Group, we use the systems engineering approach for project management and as a design tool. Systems engineering is used in a graded approach to design and project management that is based on risk, informing how much rigor to apply. The tools and techniques from systems engineering offer a framework to organize projects with everyone speaking the same language to provide consistent and repeatable project success that satisfies the stakeholders’ needs and meets the mission. The classes have provided a framework with tools for communicating system design, requirements, verification and validation, and an operational context.
The verification of nuclear warheads for arms control faces a paradox: International inspectors must gain high confidence in the authenticity of submitted items while learning nothing about them. Conventional inspection systems featuring ''information barriers'', designed to hide measurments stored in electronic systems, are at risk of tampering and snooping. Here we show the viability of fundamentally new approach to nuclear warhead verification that incorporates a zero-knowledge protocol, designed such that sensitive information is never measured so does not need to be hidden. We interrogate submitted items with energetic neutrons, making in effect, differential measurements of neutron transmission and emission. Calculations of diversion scenarios show that a high degree of discrimination can be achieved while revealing zero information. Timely demonstration of the viability of such an approach could be critical for the nexxt round of arms-control negotiations, which will likely require verification of individual warheads, rather than whole delivery systems.
The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.
Whiteshell Laboratories, located in Manitoba, Canada, provided research facilities for the Canadian nuclear industry since the early 1960's. The facility was centered on an organically cooled, heavy water moderated nuclear reactor in the world, Whiteshell Reactor 1 (WR-1). WR-1 was safely shut down in 1985, de-fueled and drained, and has been in a safe Storage with Surveillance (Deferred Decommissioning) state until present day. In support of an Environmental Assessment for proposed In Situ Decommissioning of WR-1, detailed characterization information regarding the reactor core components (calandria, calandria tubes, pressure tubes) was necessary to verify inventory estimates and neutron activation models produced circa 1992. The Fuel Channels of WR-1 (pressure tubes) were expected to be the most activated components of the reactor core, and to account for about 60% of the reactor core inventory. The interior surface of the fuel channel is also the most readily accessible surface within the reactor core. As such the decision was made to collect a sample of one of each of 3 Fuel Channel material types (Stainless Steel, Zr-2.5Nb, Ozhenite) for comparison to modelled estimates of neutron activation of various core materials. The expected dose rates from the samples were between 5 and 100 mSv (500-10,000 mrem) on contact based on a 200 mg sample. Sampling the Fuel Channels required verification the Fuel Channel was drained of organic coolant, removal of residual organic coolant from the Fuel Channel wall, removal of material from the interior surface of the Fuel Channel at the centerline of flux (∼25' below the operator position), and extraction of that material from the core. Several options for existing tooling and technologies, including hydraulic scrapers, and horizontal drill technologies, were looked at for their applicability to WR-1. While suitable approaches to higher radiation field sample collection, they were overly complex, costly and time consuming for the configuration at WR-1. In an effort to reduce costs, and to simplify the sampling process, a specialized and simple tool was developed to mechanically scrape a small sample of the interior wall of the Fuel Channels, collect the sample into a vial, and remove the sample from the core for analysis. The tool was designed, constructed, tested and implemented in the field entirely by Canadian Nuclear Laboratories Staff. The tool required no powered components, or electronic guidance or measurements; and was designed and constructed for a fraction of the cost of other commercially available technologies for sample collection within reactor cores. The samples were successfully collected without any injuries, contamination events or equipment failures. The tool remains operational in the event additional sampling is required. This paper provides additional details of the design, construction, testing and use of the tool and the successes and lessons learned during the various stages of design and implementation. (authors)
The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. One way to address the uncertainties in the HTGR analysis tools is to assess the sensitivity of critical parameters (such as the calculated maximum fuel temperature during loss of coolant accidents) to a few important input uncertainties. The input parameters were identified by engineering judgement in the past but are today typically based on a Phenomena Identification Ranking Table (PIRT) process. The input parameters can also be derived from sensitivity studies and are then varied in the analysis to find a spread in the parameter of importance. However, there is often no easy way to compensate for these uncertainties. In engineering system design, a common approach for addressing performance uncertainties is to add compensating margins to the system, but with passive properties credited it is not so clear how to apply it in the case of modular HTGR heat removal path. Other more sophisticated uncertainty modelling approaches, including Monte Carlo analysis, have also been proposed and applied. Ideally one wishes to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies, and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Therefore some safety analysis calculations may use a mixture of these approaches for different parameters depending upon the particular requirements of the analysis problem involved. Sensitivity analysis can for example be used to provide information as part of an uncertainty analysis to determine best estimate plus uncertainty results to the required confidence level. In order to address uncertainty propagation in analysis and methods in the HTGR community the IAEA initiated a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) that officially started in 2013. Although this project focuses specifically on the peculiarities of HTGR designs and its simulation requirements, many lessons can be learned from the LWR community and the significant progress already made towards a consistent methodology uncertainty analysis. In the case of LWRs the NRC has already in 1988 amended 10 CFR 50.46 to allow best-estimate (plus uncertainties) calculations of emergency core cooling system performance. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) also established an Expert Group on "Uncertainty Analysis in Modelling" which finally led to the definition of the "Benchmark for Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of LWRs". The CRP on HTGR UAM will follow as far as possible the on-going OECD Light Water Reactor UAM benchmark activity.
Our research group comprising personnel from University of Utah/EGI (UU/EGI), University of Nevada at Reno (UNR) and Lawrence Berkeley National Laboratory (LBNL) has been applying magnetotellurics (MT), diagnostic structural affiliations, and soil gas flux and geochemistry to assist in identifying hidden, high-enthalpy geothermal systems in extensional regimes of the U.S. Great Basin. We are specifically looking for upwelling low-resistivity and geologic structures that can carry fluids from magmatic or high-grade metamorphic conditions in the deep crust upward to exploitable depths, and to verify the nature of the deep sources through soil gas geochemical compositions. The project was motivated by prior MT transect coverage of western and central Nevada centered upon the Dixie Valley geothermal system where such favorable indicators were first recognized. The high-angle structures are taken to be fluidized fault zones connecting deep magmatic/metamorphic activity with the geothermal system but the concept requires verification by testing at other systems. The project was set up with a two-phased organization. Phase I was carried out at the McGinness Hills system, central Nevada, where Ormat’s flagship power facility is located and a considerable amount of pre-existing data was available. Transect MT data also showed a strong low-resistivity upwelling originating from interpreted deep crustal magmatic underplating. Controlling structures on production as indicated by Ormat data and new mapping by UNR were favorable to dilatency being an accommodation zone between major normal faults of opposing dip. A 3D MT survey and inversion confirmed the existence of the steep low-resistivity zone dipping ESE toward the deep crust and placed N-S bounds upon the feature. Ormat personnel cooperated with PI Wannamaker in the field in providing fluid samples from the production intervals for He isotope composition. Elevated 3He was verified through analysis at LBNL confirming that there was a magmatic connection into the producing system. High CO2 soil gas flux including possibly metamorphic 13C and 14C component was measured over the area of dilatent structures. Hence, the triad of indicators posed above was confirmed in Phase I. Subsequently, Phase II of the project proceeded in the greenfield Kumiva-Blackrock Desert district of northwestern Nevada to see if a new system could be identified. Transect MT data also showed a strong low-resistivity upwelling originating from interpreted deep crustal magmatic underplating. However, given that there was no exploitable geothermal system recognized at the outset in the region, the collection of new geoscientific data was required. An areal MT survey of 131 sites over the Kumiva-Blackrock Desert district was completed, and a 3D inversion produced using the in-house, DOE-supported finite element inversion algorithm. Particular MT low resistivity upwellings that received followup study occurred under the west flank of the Seven Troughs Range, under the playa immediately west of the Blue Wing Mountains, and under northern Granite Springs Valley. Structural assessment of the project area by Co-I J. Faulds at UNR provided numerous favorable Quaternary fault settings, which were correlated to the MT upwelling structures. Extensive profiling for CO2 soil gas flux over favorable geophysical and geological structure was performed by LBNL personnel. Soil gas flux anomalies generally were not large but did show correlation with MT upwelling structure and favorably dilatent geological structures. Isotope analyses showed presence of possible inorganic/metamorphic 13C but 14C concentrations did not exceed background values. We view the initial concept of a confluence of MT low-resistivity upwelling, favorably dilatent 3D geological structure, and elevated soil gas flux including 13C component to be supported by the further evidence of this project although the indicators in the Phase II study were more diffuse. The northern Granite Springs Valley structure is receiving followup in the UNR Play Fairway Analysis, being in the southern termination of a major normal fault with silicified soil and warm temperature probe results
Analysis activities for Nonproliferation and Arms Control verification require the use of many types of data. Tabular structured data, such as Excel spreadsheets and relational databases, have traditionally been used for data mining activities, where specific queries are issued against data to look for matching results. The application of visual analytics tools to structured data enables further exploration of datasets to promote discovery of previously unknown results. This paper discusses the application of a specific visual analytics tool to datasets related to the field of Arms Control and Nonproliferation to promote the use of visual analytics more broadly in this domain. Visual analytics focuses on analytical reasoning facilitated by interactive visual interfaces (Wong and Thomas 2004). It promotes exploratory analysis of data, and complements data mining technologies where known patterns can be mined for. Also with a human in the loop, they can bring in domain knowledge and subject matter expertise. Visual analytics has not widely been applied to this domain. In this paper, we will focus on one type of data: structured data, and show the results of applying a specific visual analytics tool to answer questions in the Arms Control and Nonproliferation domain. We chose to use the T.Rex tool, a visual analytics tool developed at PNNL, which uses a variety of visual exploration patterns to discover relationships in structured datasets, including a facet view, graph view, matrix view, and timeline view. The facet view enables discovery of relationships between categorical information, such as countries and locations. The graph tool visualizes node-link relationship patterns, such as the flow of materials being shipped between parties. The matrix visualization shows highly correlated categories of information. The timeline view shows temporal patterns in data. In this paper, we will use T.Rex with two different datasets to demonstrate how interactive exploration of the data can aid an analyst with arms control and nonproliferation verification activities. Using a dataset from PIERS (PIERS 2014), we will show how container shipment imports and exports can aid an analyst in understanding the shipping patterns between two countries. We will also use T.Rex to examine a collection of research publications from the IAEA International Nuclear Information System (IAEA 2014) to discover collaborations of concern. We hope this paper will encourage the use of visual analytics structured data analytics in the field of nonproliferation and arms control verification. Our paper outlines some of the challenges that exist before broad adoption of these kinds of tools can occur and offers next steps to overcome these challenges.