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Exploration and Analysis of Legacy Data at the Hanford Site Using PHOENIX - 19044

Radioactive contaminants pose risks to human health and the environment. These risks are influenced by factors such as disposal/release and transport mechanisms, location in the environment, potential exposure receptors, and the presence of comingled contaminants. Understanding the behavior, transport, and distribution of radionuclides in the environment is essential to facilitating risk-informed decision-making related to remedial design, action, and monitoring. These decisions are based on significant amounts of complex data that must be accessed, integrated, and analyzed to effectively evaluate data and formulate appropriate conclusions and recommendations. Often, data volume, complexity, and legacy access issues pose severe challenges for analysts, regulators, and stakeholders interested in evaluating scientific data to support decision-making. To address these challenges at the U.S. Department of Energy Hanford Site, Pacific Northwest National Laboratory (PNNL) has developed PHOENIX (PNNL-Hanford Online Environmental Information exchange) to facilitate risk-informed decision-making by streamlining investigative and analytical processes through the integration and visualization of previously isolated datasets created under distinct regulatory or operational regimes, each serving a specific purpose. With the large volume of data dispersed across many Hanford environmental databases, finding the correct sources of relevant data, accessing them, and relating them in a meaningful way can prove challenging and time consuming. By analyzing Hanford databases, performing substantial optimization using modern data processing methods and developing extensive queries on the refactored data, PHOENIX enables users to access, integrate, display, visualize, and retrieve data that would have been historically unfeasible to examine. The data and information accessible through PHOENIX help inform remediation-related assessments and evaluations, predict and quantify the risks that radioactive and chemical contaminants pose to human health and the environment, and communicate this information to decision makers responsible for risk-based remediation actions. (authors)

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The O{sub 2}-enriched air gasification of coal, plastics and wood in a fluidized bed reactor

Highlights: Black-Right-Pointing-Pointer The effect of the O{sub 2} in the gasification stream of a BFB gasifier has been studied. Black-Right-Pointing-Pointer Main advantage of the O{sub 2}-enriched air is the increasing of the bed temperature. Black-Right-Pointing-Pointer No remarkable effects on tar reduction. Decreasing of recognized PAHs. Black-Right-Pointing-Pointer Gasification reactions completed inside the dense bed and splashing zone. Black-Right-Pointing-Pointer Polycondensation reactions occur mainly in the freeboard region. - Abstract: The effect of oxygen-enriched air during fluidized bed co-gasification of a mixture of coal, plastics and wood has been investigated. The main components of the obtained syngas were measured by means of on-line analyzers and a gas chromatograph while those of the condensate phase were off-line analysed by means of a gas chromatography-mass spectrometer (GC-MS). The characterization of condensate phase as well as that of the water used as scrubbing medium completed the performed diagnostics. The experimental results were further elaborated in order to provide material and substances flow analyses inside the plant boundaries. These analyses allowed to obtain the main substance distribution between solid, gaseous and condensate phases and to estimate the conversion efficiency of carbon and hydrogen but also to easily visualise the waste streams produced by the process. The process performance was then evaluated on the basis of parameters related to the conversion efficiency of fuels into valuable products (i.e. by considering tar and particulate as process losses) as well as those related to the energy recovery.

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Implementation of Control Measures for Radioactive Waste Packages with Respect to the Materials Composition - 12365

In addition to the radiological characterization and control measures the materials composition has to be described and respective control measures need to be implemented. The approach to verify the materials composition depends on the status of the waste: - During conditioning of raw waste the control of the materials composition has to be taken into account. - For already conditioned waste a retrospective qualification of the process might be possible. - If retrospective process qualification is not possible, legacy waste can be qualified by spot checking according to the materials composition requirements The integration of the control of the material composition in the quality control system for radioactive waste is discussed and examples of control measures are given. With the materials-list and the packaging-list the Federal Office for Radiation Protection (BfS) provides an appropriate tool to describe the materials composition of radioactive waste packages. The control measures with respect to the materials composition integrate well in the established quality control framework for radioactive waste. The system is flexible enough to deal with waste products of different qualities: raw waste, qualified conditioned waste or legacy waste. Control measures to verify the materials composition can be accomplished with minimal radiation exposure and without undue burden on the waste producers and conditioners. (authors)

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Validation of the Modelling of Corrosion Driven Expansion in Grouted Waste-forms - 16125

One of the current preferred solutions for the disposal of Intermediate Level radioactive waste in the UK is grout encapsulation inside steel boxes for long term storage. These boxes are then to be entombed in a Geological Disposal Facility once one is available. However, both prior to and during geological disposal, waste contents can react with the water present in the encapsulant grout and begin to corrode. This corrosion product occupies a larger volume (by several hundred percent) than the original material and the corrosion occurs over the course of decades and centuries, long after the encapsulant grout has set and hardened. The expanding waste then creates stresses in the encapsulant causing both it and the surrounding box to deform and, in extreme circumstances, and after long times, for the structural integrity of the box to fail. This behaviour needs to be quantified before the box can be approved for use as any excessive deformation could lead to a loss of containment or, since the boxes are stacked on top of each other, to a significant accident in the stores. Due to the long timescales involved, physical testing of the phenomenon in full scale is impossible. Instead, a combination of chemical modelling and finite element analysis is used to predict the extent of the corrosion in the waste and its effect of the integrity of the box. There are a number of difficulties in modelling the corrosion-driven expansion and subsequent deformation of the boxes. Although the corrosion occurs over a very long time scale, the event is highly non-linear and dynamic. All of the material models, including the steel and grout are non-linear. The deformation is in the order of 10s of centimetres, well beyond the static elastic range. Finally, the components must be allowed to move relative to each other which involve a lot of part to part contacts. This means that simple static modelling, which would solve quite quickly, cannot be used. Instead a dynamic analysis code, LS-DYNA, has been used to expand the product over time to simulate the corrosion-driven expansion. This software is generally used for modelling car crashes or scenarios where waste containers have been dropped. The analyses therefore include a significant inertial component. This paper will detail the NNL's approach to modelling this long-term, quasi-static phenomenon in the dynamic, impact oriented software code LS-DYNA. It will cover the problems inherent in this approach, the errors that it creates as well as the tests and solutions that NNL worked on in order to overcome them and be confident that the solution could be validated. This work will ultimately be used to determine the maximum inventory which can be stored in each box based on the waste content and the potential extent of expansion as well as any modification that could be made to the box design or the processing streams that may improve the storage solution. (authors)

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Overview and History of DOE's Hanford Site - 12502

Hanford's DOE offices are responsible for one of the largest nuclear cleanup efforts in the world, cleaning up the legacy of nearly five decades of nuclear weapons production. Nowhere in the DOE Complex is cleanup more challenging than at the Hanford Site in southeastern Washington. Hanford cleanup entails remediation of hundreds of large complex hazardous waste sites; disposition of nine production reactors and the preservation of one as a National Historic Landmark; demolition of hundreds of contaminated facilities including five enormous process canyons; remediation of billions of gallons of contaminated groundwater; disposition of millions of tons of low-level, mixed low-level, and transuranic waste; disposition of significant quantities of special nuclear material; storage and ultimate disposition of irradiated nuclear fuel; remediation of contamination deep in the soil that could impact groundwater; decontamination and decommissioning of hundreds of buildings and structures; and treatment of 56 million gallons of radioactive waste in 177 large underground tanks through the construction of a first-of-its-kind Waste Treatment Plant. Cleanup of the Hanford Site is a complex and challenging undertaking. The DOE Richland Operations Office has a vision and a strategy for completing Hanford's cleanup including the transition to post-cleanup activities. Information on the strategy is outlined in the Hanford Site Completion Framework. The framework describes three major components of cleanup - River Corridor, Central Plateau, and Tank Waste. It provides the context for individual cleanup actions by describing the key challenges and approaches for the decisions needed to complete cleanup. The U.S. Department of Energy (DOE), as regulated by the U.S. Environmental Protection Agency (EPA) and Washington State Department of Ecology (Ecology), is implementing a strategy to achieve final cleanup decisions for the River Corridor portion of the Hanford Site. The DOE Richland Operations Office (RL) and DOE Office of River Protection (ORP) have prepared this document to describe the strategy and to begin developing the approach for making cleanup decisions for the remainder of the Hanford Site. DOE's intent is that the Completion Framework document will facilitate dialogue among the Tri-Parties and with Hanford's diverse interest groups, including Tribal Nations, State of Oregon, Hanford Advisory Board, Natural Resource Trustees, and the public. Future cleanup decisions will be enhanced by an improved understanding of the challenges facing cleanup and a common understanding of the goals and approaches for cleanup completion. The overarching goals for cleanup are sevenfold. - Goal 1: Protect the Columbia River. - Goal 2: Restore groundwater to its beneficial use to protect human health, the environment, and the Columbia River. - Goal 3: Clean up River Corridor waste sites and facilities to: Protect groundwater and the Columbia River. Shrink the active cleanup footprint to the Central Plateau, and support anticipated future uses of the land. - Goal 4: Clean up Central Plateau waste sites, tank farms, and facilities to: Protect groundwater. Minimize the footprint of areas requiring long-term waste management activities. Support anticipated future uses of the land. - Goal 5: Safely manage and transfer legacy materials scheduled for off-site disposition including special nuclear material (including plutonium), spent nuclear fuel, transuranic waste, and immobilized high-level waste. - Goal 6: Consolidate waste treatment, storage, and disposal operations on the Central Plateau. - Goal 7: Develop and implement institutional controls and long-term stewardship activities that protect human health, the environment, and Hanford's unique cultural, historical and ecological resources after cleanup activities are completed. These goals embody more than 20 years of dialogue among the Tri-Party Agencies, Tribal Nations, State of Oregon, stakeholders, and the public. They carry forward key values captured in forums such as the Hanford Future Site Uses Working Group, Tank Waste Task Force, Hanford Summits, and Hanford Advisory Board Exposure Scenario Workshops, as well as more than 200 advice letters issued by the Hanford Advisory Board (http://www.hanford.gov/page.cfm/hab). These goals help guide all aspects of Hanford Site cleanup. Cleanup activities at various areas of the site support the achievement of one or more of these goals. These goals help set priorities to apply resources and sequence cleanup efforts for the greatest benefit. These goals reflect DOE's recognition that the Columbia River is a critical resource for the people and ecology of the Pacific Northwest. The 50-mile stretch of the river known as the Hanford Reach is home to the last free-flowing section of the river in the U.S. As one of the largest rivers in North America, its waters support a multitude of uses that are vital to the economic and environmental well being of the region and it is particularly important in sustaining the culture of Native Americans. Cleanup actions must protect this river. (authors)

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Mitigation of Radioactive Waste by Sampling and Analyzing the Affected Area and Efficient Use of Different Detectors - 17139

The paper describes the ways by which the mitigation of the radioactive waste during events like decommissioning of nuclear plants, mines, other nuclear-related facilities and also in event of nuclear disaster is done, so that the rehabilitation process for flora and fauna can begin in a natural manner. There are many parameters that have to be kept in mind during the planning stage of environmental remediation (ER), as there are many hindrances and snags that might occur during environmental remediation and rescue operation. The paper describes the problems faced during ER operations in the developing countries and needs of international committees. The paper emphasizes proper site characterization and suggestions for it, and suggests simultaneous detection of various activity of radionuclides dispersed and the need for proper knowledge and handling in ER operations. It describes how the radioactivity can be an interference in detection and the importance of low-level activity. The paper describes why site characterization should be quick and prior to other important key points during rescue operations. Also, it gives a description of various detectors used widely and their related significance in ER. The paper describes the strategy on how after site surveying the input can be used to map an area, as to provide easy operation and working conditions for personnel, and how it can be helpful in the evaluation of other parameters. The paper also gives suggestions on the use of bio-indicators in above described events and their benefits. (authors)

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Seismic Testing of Glovebox Feedthrough Connectors and Vacuum Pumps

In the design of tritium processing systems, the glovebox stripper system processes the glovebox atmosphere to recover tritium released inside the glovebox confinement. The glovebox stripper system must also be capable of processing the additional in-leakage caused by a Design Basis Accident (DBA) earthquake, in order to maintain the glovebox pressure negative relative to the room. However, it is unknown how much additional leakage occurs from electrical bulkhead feedthrough connectors after a DBA event. As there are dozens of these connectors on a tritium glovebox system, the potential increase in in-leakage can be significant. One vital part of the glovebox stripper system are the vacuum pumps that maintain the negative atmosphere of the glovebox following a DBA earthquake. The pumps must be able to operate following an earthquake. One candidate pump is the Senior Aerospace MB-602 Metal Bellows (Met-Bel) pump. A second alternative is the hermetically sealed Edwards Vacuum nXDS15iC Scroll Pump. The Edwards scroll pump is potentially a less expensive alternative to the Met-Bel pumps, but it is unknown if this scroll pumps will survive a DBA accident and remain operational. In order to determine the electrical feedthrough in-leakage and the candidate pumps’ survivability, a seismic test was performed to simulate a DBA earthquake. This report summarizes the results of seismic test, including helium leak rates and functionality tests for the electrical feedthroughs and vacuum pumps. All components had pre- and post-test leak rates less than the threshold valve of 1x10 -6 scc per sec.

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Use of Ductile Iron for Radioactive Waste Packaging and Shielding in the United States - 17278

Ductile Cast Iron (DCI) has been used for packaging radioactive waste in Europe and the United Kingdom for many years. The ductile iron provides many favorable properties when compared with steel and offers many advantages including: - Monolithic structure increasing overall strength; - No welding required; no tempering or maintenance/inspection of the weld area is required; - Fast, serial production of similar containers reducing manufacturing costs; - Lower cost compared to steel or steel/lead composites; - Good corrosion resistance even when uncoated; - Can be coated or plated to provide suitable surface for decontamination; - Flexibility in design due to the manufacturing process permits a variety of shapes and very large size container possibilities; - Hazardous materials not needed or used for shielding. The flexibility in design will help generators find viable and more cost effective packaging options for radioactive waste. The package is conformed to the waste instead of conforming waste to the package. Commercial utilities often replace radioactive components which are too large for conventional packaging and have dose rates too high for flexible wraps while plant schedule and physical configuration won't allow for component segmentation. This is one situation where the use of ductile iron can be used to provide a packaging solution. By assessing the radiological characterization and component size and weight, ductile iron can be cast into certified containers to easily package, ship and dispose of the waste in a cost effective manner. (authors)

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Drying Spent Magnox Fuel for Interim Dry Storage

Spent Magnox fuel is usually stored under water for shielding and cooling following use in-reactor, but the storage water can react with Magnox clad fuel and cause corrosion and forms a sludge which generates a secondary waste product. One option to prevent unwanted corrosion and avoid expensive waste management is to dry the wet fuel. This project aims to test a drying process which will convert wet and corroded Magnox fuel to a stable form to demonstrate viability for conversion from wet to dry storage. Samples of corroded Magnox supplied by National Nuclear Laboratory. Corrosion product retrieved by various methods including dry scraping, ultrasonic treatment to form sludge and sludge drying to isolate suspension. TGA undertaken on Sample 1 and Sample 2, with brucite and hydromagnesite for comparison. Both samples showed ∼10% mass loss up to 250 deg. C attributable to water held in sample. Mass loss from dehydroxylation which liberates further water in both samples. TGA data plotted alongside chemically pure brucite and hydromagnesite shows expected similarities/differences between Sample 1(mostly brucite) and Sample 2 (brucite/hydromagnesite mixture). Following treatment, dried samples observed to slowly gain mass in ambient conditions. Presumed to be hygroscopic absorption of air water vapour as effect is reversible, with similar effect observed for hydromagnesite and brucite. Etched Magnox vacuum dried for 3.5 h at 90 deg. C to test for reaction with residual water - some discoloration but no noticeable corrosion observed. Drying etched Magnox metal to observe surface effects/corrosion during drying process.

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Preliminary Safety Analysis of the Gorleben Site: Safety Concept and Application to Scenario Development Based on a Site-Specific Features, Events and Processes (FEP) Database - 13304

Based upon the German safety criteria, released in 2010 by the Federal Ministry of the Environment (BMU), a safety concept and a safety assessment concept for the disposal of heat-generating high-level waste have both been developed in the framework of the preliminary safety case for the Gorleben site (Project VSG). The main objective of the disposal is to contain the radioactive waste inside a defined rock zone, which is called containment-providing rock zone. The radionuclides shall remain essentially at the emplacement site, and at the most, a small defined quantity of material shall be able to leave this rock zone. This shall be accomplished by the geological barrier and a technical barrier system, which is required to seal the inevitable penetration of the geological barrier by the construction of the mine. The safe containment has to be demonstrated for probable and less probable evolutions of the site, while evolutions with very low probability (less than 1 % over the demonstration period of 1 million years) need not to be considered. Owing to the uncertainty in predicting the real evolution of the site, plausible scenarios have been derived in a systematic manner. Therefore, a comprehensive site-specific features, events and processes (FEP) data base for the Gorleben site has been developed. The safety concept was directly taken into account, e.g. by identification of FEP with direct influence on the barriers that provide the containment. No effort was spared to identify the interactions of the FEP, their probabilities of occurrence, and their characteristics (values). The information stored in the data base provided the basis for the development of scenarios. The scenario development methodology is based on FEP related to an impairment of the functionality of a subset of barriers, called initial barriers. By taking these FEP into account in their probable characteristics the reference scenario is derived. Thus, the reference scenario describes a comprehensive set of probable future evolutions of the repository site. By stepwise consideration of less probable FEP or less probable characteristics of FEP within this process, alternative scenarios are also developed, which are characterized by a lower probability of occurrence. An important methodological aspect is that some assumptions had to be made for the scenario development. They allow, on the one hand, to deal systematically with incomplete knowledge regarding the geological situation below ground owing to restricted site investigations, and, on the other hand, to structure the scenario development process. The consideration of alternative assumptions may result in additional alternative scenarios. (authors)

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Keep It Smart and Simple - Henriette - The Advanced Designed Drum Gripper for Remote Handling Tasks - 19327

A power manipulator has many applications in the handling, treatment and packaging of radioactive waste. In high radiation environments, where humans cannot work safely, a power manipulator is essential to perform difficult and heavy tasks. For more than 40 years, Waelischmiller has provided power manipulators to the nuclear market. Specifically in a high radiation nuclear environment, electrical driven manipulators are preferred. The basic design of a power manipulator is divided into the following parts: Bridge, carriage, telescopic mast and manipulator arm. All parts are customizable in length, width and capacity. The standard power manipulator version is equipped with a manipulator arm. For the nuclear waste disposal industry, there is a special need for a drum gripper. Drums are used for packing low, medium or high activated material and for storage. After packing, the drums must be handled remotely. For both tasks, a power manipulator with a drum gripper is needed. For this project KTE (Kerntechnische Entsorgung Karlsruhe GmbH) required a total refurbishment of equipment in a cell used for handling, sorting and packaging radioactive waste drums. The power manipulator was obsolete and provided low availability and through put. This cell was an essential asset in their decommissioning process and was impacting KTE's ability to meet their volume objectives. Our challenge was to develop a new system with few details from the original cell construction. The new manipulator must meet all capabilities of the old system and in addition have the ability to achieve 180 degree tilting of the drums. An additional challenge was the requirement to use the existing rails in the cell which were very close to the roof structure which required a special design. Our project, named internally Henriette, is the first of its kind. The drum gripper is based on the power manipulator A1000 series. The biggest challenges in this project were to adapt the standard components of the power manipulator to the non-standard type of gripper to have 180 deg. tilting of the drum and a handling capacity of 1200 kg in the whole working range. The new manipulator fits to the old rails and is mountable in a hot cell with restricted access and can operate with the other A1000 manipulator which is located in the cell. The result is a system with no external wiring, maintenance-free and easy to decontaminate. Other features include a 5 t auxiliary hoist on additional and parallel carriages, emergency drives for bridge and carriage, emergency drive for telescope turn movement and emergency lifting unit for telescope. A tilt and pan holder for camera on the telescope and external mechanical emergency drives on the gripper were also essential. This paper will address the complexity, challenges and engineering ingenuity required to replace very old and obsolete equipment with a state-of-the-art system within very limited physical space and within a high radiation environment. (authors)

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Tank 21 and tank 24 blend and feed study: blending times, settling times, and transfers

The Salt Disposition Integration (SDI) portfolio of projects provides the infrastructure within existing Liquid Waste facilities to support the startup and long term operation of the Salt Waste Processing Facility (SWPF). Within SDI, the Blend and Feed Project will equip existing waste tanks in the Tank Farms to serve as Blend Tanks where salt solutions of up to 1.2 million gallons will be blended in 1.3 million gallon tanks and qualified for use as feedstock for SWPF. In particular, Tanks 21 and 24 are planned to be used for blending and transferring to the SDI feed tank. These tanks were evaluated here to determine blending times, to determine a range of settling times for disturbed sludge, and to determine that the SWPF Waste Acceptance Criteria that less than 1200 mg/liter of solids will be entrained in salt solutions during transfers from the Tank 21 and Tank 24 will be met. Overall conclusions for Tank 21 and Tank 24 operations include: (1) Experimental correction factors were applied to CFD (computational fluid dynamics) models to establish blending times between approximately two and five hours. As shown in Phase 2 research, blending times may be as much as ten times greater, or more, if lighter fluids are added to heavier fluids (i.e., water added to salt solution). As the densities of two salt solutions converge this effect may be minimized, but additional confirmatory research was not performed. (2) At the current sludge levels and the presently planned operating heights of the transfer pumps, solids entrainment will be less than 1200 mg/liter, assuming a conservative, slow settling sludge simulant. (3) Based on theoretical calculations, particles in the density range of 2.5 to 5.0 g/mL must be greater than 2-4 {micro}m in diameter to ensure they settle adequately in 30-60 days to meet the SWPF feed criterion (<1200 mg/l). (4) Experimental tests with sludge batch 6 simulant and field turbidity data from a recent Tank 21 mixing evolution suggest the solid particles have higher density and/or larger size than indicated by previous analysis of SRS sludge and sludge simulants. (5) Tank 21 waste characterization, laboratory settling tests, and additional field turbidity measurements during mixing evolutions are recommended to better understand potential risk for extended (> 60 days) settling times in Tank 21.

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A holistic life cycle analysis of waste management scenarios at increasing source segregation intensity: The case of an Italian urban area

Highlights: • Waste management scenarios starting from different SS intensity were analyzed by LCA. • Several experimental data were utilized for the inventory. • Collection activities influences marginally global impact. • Maximum global environmental gain was achieved by SRF for coke substitution. - Abstract: Life cycle analysis of several waste management scenarios for an Italian urban area was performed on the basis of different source segregation collection (SS) intensities from 0% up to 52%. Source segregated waste was recycled and or/recovered by composting. Residual waste management options were by landfilling, incineration with energy recovery or solid recovered fuel (SRF) production to substitute for coal. The increase in fuel and materials consumption due to increase in SS had negligible influence on the environmental impact of the system. Recycling operations such as incineration and SRF were always advantageous for impact reduction. There was lower impact for an SS of 52% even though the difference with the SS intensity of 35% was quite limited, about 15%. In all the configurations analyzed, the best environmental performance was achieved for the management system producing SRF by the biodrying process.

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Formation and growth of cerium (III) oxalate nanocrystals by liquid-cell transmission electron microscopy

Using rare earth elements as analogs for actinide series elements enables studying the chemistry of these radioactive materials without the inherent risks. A fundamental understanding of the nucleation and growth of actinide oxalates is crucial for processing nuclear waste. Using in-situ liquid cell TEM (LC-TEM), we show particle attachment processes occurring during the nucleation and growth of cerium oxalate. The early stages of the observed non-classical growth mechanisms have important implications for particle evolution and predictions of final particle morphology. In this work, we observed particle formation via monomer-by-monomer addition or the particle aggregation pathway depending on the selected precursor concentration. At the highest precursor concentration at a high dose rate, we observed particle alignment in a branch structure is due to the interfacial instability. It was demonstrated that in-situ LC-TEM is an invaluable tool in studying the cerium oxalate formation and ultimately a path to understanding actinide chemistry.

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Analysis of Harrell Monosodium Titanate Lot #050411

Monosodium titanate (MST) for use in the Actinide Removal Process (ARP) must be qualified and verified in advance. A single qualification sample for each batch of material is sent to SRNL for analysis, as well as a statistical sampling of verification samples. The Harrell Industries Lot #050411 qualification and 12 verification samples met all the requirements in the specification indicating the material is acceptable for use in the process.

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Analysis of Harrell Monosodium Titanate Lot #052511

Monosodium titanate (MST) for use in the Actinide Removal Process (ARP) must be qualified and verified in advance. A single qualification sample for each batch of material is sent to SRNL for analysis, as well as a statistical sampling of verification samples. The Harrell Industries Lot No.052511 qualification and 14 verification samples met all the requirements in the specification indicating the material is acceptable for use in the process.

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Analysis of Harrell Monosodium Titanate Lot #071311

Monosodium titanate (MST) for use in the Actinide Removal Process (ARP) must be qualified and verified in advance. A single qualification sample for each batch of material is sent to SRNL for analysis, as well as a statistical sampling of verification samples. The Harrell Industries Lot No.071311 qualification and 12 verification samples met all the requirements in the specification indicating the material is acceptable for use in the process.

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Small-Scale Experiments-Sandia Instrumented Thermal Ignition (SITI) testing summary

A number of cook-off experiments were performed to provide understanding of potential thermal ignition of drum 68660 in the February 14, 2014 radiation release at WIPP. Testing was begun according to the plan provided in Appendix B, and deviated significantly based on initial findings and other information developed during the testing. Two general types of experiments were performed: ones with nitric acid neutralized to varying degrees with Kolorsafe neutralizer, and ones with no added free liquid. Results indicate that reactivity is greater in the dry mixture, and that Fe nitrate and Ca nitrate play significant roles in ignition behavior whereas Pb nitrate, Cr nitrate, and oxalic acid do not. Within mixtures with liquid, very little exothermic behavior is observed with Swheat and water, but adding neutralized acid and nitrate salts results in significant reactivity and ignition. This behavior is suppressed by liquid water, and ignition occurs after the water has fully vaporized, although it is not clear if ignition occurs quickly because of the relatively high wall temperature at the end of the vaporization process.

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