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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Community Engagement - A Citizen-Centric Approach to Seeking a Social License - 20017

The long regulatory road has finally been completed-after years of effort a Record of Decision has been signed and a Permit to Construct has been granted or a Permit to Operate has been obtained. But what about the other road that should be taken in parallel-community engagement? Working with the public is not a sequential activity, rather it is a consequential one and needs to be integrated with overall project planning. Unless you have effectively, meaningfully, and patiently engaged with the project's communities of impact, you may find your project trying to obtain a social license or a social license to operate (SLO). The SLO has its origins in the mining industry and its roots in the business model of corporate social responsibility and sustainability. These latter practices are well known to the US DOE and its contractor community. Aspects of the SLO are emerging as individuals and communities are becoming more informed and have increased expectations for being able to influence and shape decisions. When local community issues are not sought out, listened to, or addressed early, and questions are left unanswered, they can become agenda items for larger unaffiliated groups, and project loss (through delays and/or cancellation) can occur. Issues can transition from resolvable to intractable, a type of SLO face-off. Social media campaigns, serial negative media coverage, and protest signs at project sites opposing regulatory decisions already made are no longer anomalies. These types of incidents demonstrate the increasingly delicate relationship between approved regulatory/technical decisions and public acceptance of those decisions. A SLO is not a requirement. However, the building-blocks of a SLO - working with members of affected communities to build understanding, potentially to obtain and maintain community acceptance or even gain approval or support - are evidence of leadership by project sponsors. Actions taken, or not taken, by project sponsors truly have the ability to influence an outcome. While projects are not assured of success or failure, engaging the community in an empowered process is an investment of time and resources toward success for the project and thus for the communities of impact. Authors Note: Extensive literature searches were performed in developing this paper. The majority of the published literature on SLO as a 'movement' was found from sources in Canada, the European Union (Germany and France in particular), South Asia, South Korea, and many Latin American countries. The literature addressed resource development (mining), infrastructure, and energy projects. Articles written by US sources often spoke to origins and structural theory, specific project issues such as Not in My Back Yard (NIMBY) and to protests/demonstrations related to those projects, rather than systemic opposition/conflict phenomena. For these reasons many international SLO experiences and sources are cited in the discussion. The authors recognize these non-US experiences with SLO as having a 'forecast' value to US projects, especially those related to nuclear projects of any type, waste treatment and disposal, energy development (especially fossil fuels as well as alternative sources), mining, and infrastructure. The body of future SLO experiences in the US will shape the future responses to it. (authors)

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Development of New Antifoam for High Level Waste Processing in DWPF

During processing in the Defense Waste Processing Facility (DWPF), the evaporation of sludge leads to large amounts of foaming. This foaming can cause a foam-over, which contaminates other parts of the facility, most importantly the condensate. Currently, the antifoam being used is not sufficient at more extreme pH values and relatively high concentrations are required to minimize foaming. This means that a foam-over is more likely in parts of the process where the pH moves away from the neutral point and antifoam is needed in large quantities to prevent such occurrences. Adding this antifoam can also double processing time. The current antifoam also degrades into flammable byproducts, which creates a hazard in the processing. There are two main ideas proposed in order to solve the problems with the current antifoam: A new antifoam; A mechanical solution (sprays, agitator, etc). The focus of this summer was to develop a new antifoam that works across a broader range of pH values, can be added in lower concentrations, and does not degrade into flammable byproducts. This solution was pursued by obtaining different superspreaders and antifoams from companies across the country in order to perform a series of tests on them; these tests include superspreader testing, Foamscan testing, boiling testing, and Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) process testing. To begin testing, four different superspreaders were tested on their ability to spread. The four superspreaders include: Y-17112, Y-17309, Antifoam 747 (the old antifoam), and Silwet L77 (the main component in Antifoam 747). These superspreaders were tested across a range of pH values and concentrations from 100 ppm to 5000 ppm. After completion of the tests, the results indicated that Y-17112 spread across a wider pH range than the others, even at lower concentrations. This is ideal for developing a new antifoam, as the pH throughout processing in DWPF varies and the lower the necessary concentration, the less antifoam that must be added per cycle. In order to test whether or not the superspreaders work as an antifoam, they have to be tested on real foam. In order to test this, a foamy simulant similar to the foaminess of the sludge in DWPF needed to be developed. This required testing many different sludges and designing our own simulant with different concentrations of insoluble solids. Eventually, a simulant similar enough to what is seen in DWPF was identified. Each superspreader will be tested on the simulant at different pH values and at different concentrations. The simulant will be allowed to foam and small amounts of the superspreader will be added while the Foamscan software tracks the height of the foam column; this allows us to know if the superspreader decreased the foam or not. Next, boiling testing in the Foamscan will be completed. This is testing identical to the Foamscan testing at room temperature, only at boiling conditions similar to DWPF. After boiling testing, the superspreader that is deemed most successful will be tested in a real, scaled-down SRAT/SME cycle. This will ensure it works with all the real chemicals that are added in DWPF and that it will remain successful in all the conditions that it may face. After the SRAT/SME testing, a decision will be made on whether or not the superspreader is successful and it may or may not be chosen as the new antifoam for DWPF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Antarctica Radiological Source Removals (Final Report)

On 18 May 2015, the research vessel Akademik Fedorov of the Russian Antarctic Expedition (RAE) delivered four radioisotope thermoelectric generators (RTGs) from Antarctica to the port of St. Petersburg, Russia. The RTGs had been previously used in the work program of the former Soviet Antarctic Expedition (SAE). From the mid-1970s, RTGs have been used to provide autonomous power at geophysical and meteorological stations located in remote regions of Antarctica. The research was completed in the early 1990s, but not all the generators had been removed when the stations were decommissioned. The remaining four unsecured RTGs, with their strontium-90 ( 90 Sr) radiological material and a combined total activity of 80,000 curies (Ci), still posed a potential risk as they could be used for nefarious purposes. Therefore, a plan of action for the removal of the RTGs from Antarctica was developed. This was done within the framework of the international program for the disposal of Russian RTGs. Financial support and technical oversight was provided by the U.S. Department of Energy (DOE), while logistical support was provided by the Russian Antarctic Expedition (RAE) of the Arctic and Antarctic Research Institute (AARI) and by subsidiaries of the Russian Federation’s Rosatom State Corporation. Recovering and repatriating the RTGs from the continent of Antarctica to the Russian Federation proved to be a very difficult task, including satisfying all the legal aspects within a very tight timeframe. This involved indemnification issues, adhering to new laws enacted after the Soviet Union era, and obtaining all the necessary permits and certifications for transport of the RTGs by air, sea, and land. It was an international, collaborative effort involving the cooperation of multiple organizations to complete the legal work and preparation of necessary documents. In addition, AARI needed to create a separate structure within its organization to manage the process, including the provision of special medical screenings, training, and certifications of its institute personnel. This was done so AARI could fulfill its obligation for the safe handling of radioactive materials while in the field. Recovery of the RTGs required overcoming several logistical hurdles. The Akademik Fedorov, which transported the RTGs back to the Russian Federation, required retrofits to accommodate the RTGs for safe and secure transport of radioactive material. The United States Project Team (USPT) inspected the vessel while it was in dry dock in Turku, Finland, being retrofitted with their design. Special sleds had to be designed to transport RTGs, as well as equipment for identifying crevasses in the glaciers, over the Antarctic packed snow and ice. Provision of adequate fuel supplies for the long trek to the interior of Antarctica to recover one of the RTGs also had to be addressed. With the successful recovery, repatriation, and disassembly of the RTGs, the Radioactive Heat Sources (RHSs) were extracted and dispositioned, thus eliminating any risk of these radioactive sources being used for terrorist purposes. Successful completion of the project also fulfilled the requirements of the Protocol on Environmental Protection to the Antarctic Treaty concerning the need to remove unused radioactive materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Multi-discipline Waste Acceptance Process at the Nevada National Security Site - 13573

The Nevada National Security Site low-level radioactive waste disposal facility acceptance process requires multiple disciplines to ensure the protection of workers, the public, and the environment. These disciplines, which include waste acceptance, nuclear criticality, safety, permitting, operations, and performance assessment, combine into the overall waste acceptance process to assess low-level radioactive waste streams for disposal at the Area 5 Radioactive Waste Management Site. Four waste streams recently highlighted the integration of these disciplines: the Oak Ridge Radioisotope Thermoelectric Generators and Consolidated Edison Uranium Solidification Project material, West Valley Melter, and classified waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Technical basis for the reduction of the maximum temperature TGA-MS analysis of oxide samples from the 3013 destructive examination program

Thermogravimetric analysis with mass spectroscopy of the evolved gas (TGA-MS) is used to quantify the moisture content of materials in the 3013 destructive examination (3013 DE) surveillance program. Salts frequently present in the 3013 DE materials volatilize in the TGA and condense in the gas lines just outside the TGA furnace. The buildup of condensate can restrict the flow of purge gas and affect both the TGA operations and the mass spectrometer calibration. Removal of the condensed salts requires frequent maintenance and subsequent calibration runs to keep the moisture measurements by mass spectroscopy within acceptable limits, creating delays in processing samples. In this report, the feasibility of determining the total moisture from TGA-MS measurements at a lower temperature is investigated. A temperature of the TGA-MS analysis which reduces the complications caused by the condensation of volatile materials is determined. Analysis shows that an excellent prediction of the presently measured total moisture value can be made using only the data generated up to 700 °C and there is a sound physical basis for this estimate. It is recommended that the maximum temperature of the TGA-MS determination of total moisture for the 3013 DE program be reduced from 1000 °C to 700 °C. It is also suggested that cumulative moisture measurements at 550 °C and 700°C be substituted for the measured value of total moisture in the 3013 DE database. Using these raw values, any of predictions of the total moisture discussed in this report can be made.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A New Group Crystallization Approach in Used Nuclear Fuel Recycling: A Look at Fission Product Behaviors

Nuclear energy has the potential to play a large role in meeting future needs of the U.S. sustainable energy strategy, as a clean, carbon-free power source. However, to take advantage of the inherent benefits, many hurdles must be overcome in order for nuclear power to achieve these expectations. One of the largest challenges faced, is the management of used nuclear fuel (UNF). Currently, no strategy has been identified to deal with the growing UNF problem, and so the UNF remains stored on site at nuclear power plants either in cooling pools or in dry cast storage. The amount of stored UNF continues to grow, and with it, so does the concern surrounding its long-term radiotoxicity. This material could be reprocessed to recover the actinides, maximizing the energy potential of the UNF and minimizing the waste generation. Not only do U and Pu, the major energy producing actinides, need to be recovered, but recovering Np and Am would make waste generated from reprocessing significantly safer for long-term storage in a geological repository. There are a number of methods currently in use to reprocess UNF; throughout Europe, the Purex process is widely implemented, while the Japanese have developed the Next process. The Next process removes a bulk of the U by crystallization prior to a Purex-like separation scheme. In the Next process, it is undesirable to have any other species removed from solution in the crystallization step, but rather the goal is to obtain a pure uranyl nitrate hexahydrate (UNH) product. Alternatively, a group separation of U-Am by a crystallization process would provide an elegant solution. Such a process has become attainable with the renewed interest in accessing Am(VI) in nitric acid solutions. In order to synthesize and stabilize Am(VI), a highly oxidizing system is required, which has only been demonstrated on a limited basis. By oxidizing Np, Pu, and Am to their hexavalent state, they can be co-crystallized with UNH, and thereby removed from solution. Recent results have demonstrated that all the hexavalent actinide (An) species are removed from solution in near proportion, by a simple reduction of temperature in 6-8 M HNO{sub 3}. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Improving PCM Waste Management Through New Approaches to Screening for LLW Treatment and Disposal - 20145

A major step forward in the management of Plutonium Contaminated Material (PCM) Waste has been taken at the Sellafield site with the introduction of new capabilities to enable wastes previously conservatively designated as PCM Intermediate Level Waste (ILW) to be re-classified for treatment as Low Level Waste (LLW). The Sellafield site, located in the North West of England, within the United Kingdom (UK), has operated plutonium processing plants and waste facilities for almost 70 years. During this time management of the raw PCM waste produced has presented a variety of challenges. The waste is currently processed by high force waste compaction in the Waste Treatment Complex (WTC). The WTC converts 200 litre drums of raw PCM waste into a compacted, grout encapsulated monolith form within stainless steel containment, ready for emplacement within the UK's Geological Disposal Facility (GDF) when it becomes available. To help progress some of the wastes waiting for treatment, recent improvements have been made with the introduction of new waste assay instrumentation which enables the identification of a 'candidate' waste potentially suitable for disposal via the LLW treatment route. Opening a new LLW route for previously designated PCM has been a major advance in PCM waste management policy. It has required a range of technical assessment arrangements to be developed, tested and commissioned and involved the engagement of a wide range of stakeholders. In particular, very detailed technical analysis methods have been developed to ensure that there is a robust and conservative approach to the final sentencing of waste. Following completion of commissioning tests, the first 86 m{sup 3} of previously designated PCM waste was successfully treated through the LLW route and consigned to LLW disposal in June 2019. The direct benefit of sentencing waste to the correct waste route (LLW) is the significant cost savings across the complete range of waste management activities. This work is also strategically important for demonstrating to a range of stakeholders including workforce and regulators, that waste previously designated as PCM can be re-classified in a systematic and justifiable way without increasing risk to personnel or the wider environment. By ensuring that waste is being managed appropriately it will be possible to carry out more remediation work safely, in less time, and at a lower cost. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Investigation into Reductive Separation of Tc-99 from Low Activity Waste Off-Gas Condensate by Zero Valent Iron Materials - 20311

This study is investigating the potential for reductive separation of technetium-99 (Tc) from off-gas condensate generated from vitrification of legacy tank waste at the Hanford Tank Waste Treatment and Immobilization Plant. Management of Tc is a high-priority activity for the U.S. Department of Energy (DOE) Office of Environmental Management (EM) complex, which must be addressed based on knowledge gaps and technology development needs outlined in the DOE-EM Technetium Management Program Plan [1]. Technetium is one of the main risk drivers in the permanent disposal of low activity waste (LAW) by vitrification at the Hanford Site due to its high volatility and only fractional incorporation into borosilicate glass. Volatilized Tc will be captured by an off-gas treatment system, and the current plan is to continuously recycle generated condensate back to the LAW glass melter feed. Off-gas recycling is effective at increasing Tc loading in the LAW glass, but it also disproportionately increases concentrations of sulfate, halides, and other volatile compounds impeding overall LAW processing leading to significantly greater waste glass volume requirements. Removal of Tc from secondary liquid waste offers an alternative disposal route to continuous off-gas recycling which may result in significant reduction of the volumes of LAW waste glass and substantial cost savings. The objective of this research funded by the DOE-EM Minority Serving Institutions Partnership Program (MSIPP) is to determine the suitability of zero valent iron (ZVI) as a separation technology to remove Tc from the off-gas condensate stream to minimize its recycle. While ZVI has been proposed for various remediation applications with Tc, it has not been studied for its suitability for removing Tc from the LAW off-gas condensate. The goal is to identify the optimal conditions for which Tc{sup (VII)} will be reduced to Tc{sup (IV)} species and incorporated into an iron matrix. This paper presents a programmatic overview of our evaluation of commercial ZVI materials for their efficiency towards Tc{sup (VII)} reductive removal from simplified off-gas simulant solutions under aerobic conditions. Tc concentrations in solutions were monitored for up to 30 days to determine efficiency and kinetics of reductive removal of Tc. The best performing candidate materials were selected for further experiments evaluating the effects of (1) variable ionic strength, (2) the presence of chromium and corrosive anions, and (3) ZVI aging on Tc{sup (VII)} reductive removal in batch experiments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Demonstration Testing of a Thermal Desorption Unit to Receive and Treat Waste with Unlimited Concentration of PCBs - 13437

For the last nine years, EnergySolutions and TD*X Associates LP have teamed up to provide the most comprehensive organic removal treatment process in the radioactive waste industry. The high performance thermal desorption unit (HP-TDU) located at the EnergySolutions Clive facility in Utah has successfully processed over 1,850 tons of organically contaminated radioactive mixed waste. Products from the HP-TDU system include a radioactively contaminated dry solid material that can be disposed in the on-site landfill and an organic condensate with high thermal energy content that is generally below background radiation and capable of free-release to a non-radioactive incinerator. Over the years, Permits and approvals have been obtained through the state of Utah, United States Environmental Protection Agency (USEPA) Region 8, and USEPA headquarters that enable the treatment of several waste categories including volatile and semi-volatile organic compounds, combustion-coded (CMBST) compounds, volatile metals, and polychlorinated biphenyls (PCBs). The unit has recently successfully completed Demonstration Testing for PCB concentrations up to 660,000 ppm (parts per million). Solid processed material from this Demonstration Testing was less than two ppm PCBs in three separate treatment runs; reprocessing or additional treatment was not needed to meet this limit. Through post-demonstration permitting, the system is unlimited in scope as approval has been given to receive and solidify up to pure PCBs down to this processing limit concentration to complete treatment of mixed waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A Model from the Mind to the Operating Facility - 16448

The reprocessing of Spent Magnox fuel has been undertaken at the Sellafield site in the United Kingdom (UK) for more than 50 years and is now entering the final few years of its operating programme. The age of this facility has resulted in additional operational challenges which have led to the need for a sustained programme of continuous improvement. It has therefore been necessary to assess what potential plant improvements are key to delivering a successful end to the operating programme. In 2003 the UK National Nuclear Laboratory (NNL) (then Nexia Solutions) worked with Sellafield Ltd to produce a visual discrete event simulation (DES) model. The DES approach models plant equipment (such as a storage pond crane or fuel element processing equipment), and the key processes that these items undertake. Process times, availability and process logic are all combined to provide a prediction of the plant performance. Through the use of 'what-if' scenarios, changes to any model parameter can be undertaken. The benefits of such an approach have been in the ability for Sellafield Ltd to understand which changes to make and which do not add value. This then allows the operating facility to focus the improvement programme on quantitatively underpinned benefits, thus increasing the potential to complete the reprocessing programme ahead of schedule, which is a key aim for Sellafield Ltd. Maximising the benefit of this model has required a close working relationship between NNL, Sellafield Ltd Technical and the Sellafield Ltd Operations teams to ensure that the model is using the correct information and is addressing the real issues at hand. Over the past 12 years many ideas have been assessed and three examples have been selected to demonstrate the range of use of the model, and the different approaches used to identify issues and opportunities, agree an approach and provide solutions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Building flexibility into the design of a pilot plant for the immobilisation of Pu containing residues and wastes

NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100 kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated waste-forms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable waste-forms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL's Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO{sub 2} as a surrogate for PuO{sub 2}. This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of immobilising the identified waste and residues inventory within 3 years. During that period it will also be used to demonstrate the potential of this technology to deliver the immobilisation of a proportion of the UK plutonium stockpile that may not be suitable for use as MOx fuel should that decision be taken. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Radioisotope Analysis of Wastewater from Livermore Site Retention Tanks by Gel Laboratory Gross Alpha, Gross Beta and Tritium Sampling Method

Lawrence Livermore National Laboratory discharged approximately 4.4% of the City of Livermore’s total wastewater in 2022 (LLNL’s Annual Site Environmental Report, Chapter 5, 2022). This volume includes wastewater from Sandia National Laboratories (SNL) and some process wastewater from Site 300. Due to the high volume and constituents of the discharge, LLNL works alongside the City of Livermore under permit #1250, requiring wastewater generated to be monitored and sampled in accordance with permit limits. Process wastewater, from buildings with the highest risk to sewer, is collected by wastewater retention tanks throughout the Livermore Site and sampled prior to discharge. Domestic wastewater directly discharges to sanitary sewer. To maintain permit requirements and ensure proper wastewater discharge practices, an internal wastewater audit was conducted during the summer of 2023. Current wastewater practices, regulatory knowledge and risk management across various Livermore Site buildings were evaluated. Workspaces connected to a wastewater retention tank and sanitary sewer drains were major focus areas. Data collected from walk-throughs prompted further evaluation as many practices were reported to be done based on historical usage. A table of concerns was created to showcase reasons for auditing and proceeding action. An analysis of current and historical retention tank usage throughout LLNL Livermore Site buildings with radioisotope results over a 5-year period from 2019 to 2024, was done to assess building trends and any significant changes throughout the 5-year period. Analytes evaluated were Gross Alpha, Gross Beta and Tritium (GABT) of eleven buildings at the Livermore Site, posing the highest risk to sanitary sewer for radioisotopes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Utilizing a Novel Acoustic Backscatter Array to Characterize Waste Consolidation and Settling in a Horizontal Flow Clarifier - 16051

As part of a collaboration between the University of Leeds, Sellafield Ltd and MMI Engineering Ltd, bespoke acoustic instrumentation and numerical simulation methods have been established to optimize nuclear legacy waste transfers. Specifically, a new ultrasonic measurement system has been developed, allowing suspended particle concentration and consolidation to be measured remotely. This technology has been combined with computational fluid dynamics (CFD) modelling of settling particulate systems, permitting full characterization, simulation and prediction of sludge separation processes. Presented here are experimental and numerical results on the transport and settling behavior of non-active nuclear-analogue suspensions in two flow geometries: gravitational settling of glass dispersions in a 250-liter column; and transport and sedimentation of flocculated calcite in a close-to-full-scale clarifier (9 m{sup 3} total volume and 4.8 m{sup 2} length). In addition, details of the design, construction and calibration of a custom-built Ultrasonic Array Research Platform (UARP) are described. Settling column results suggested that particle concentration changes could be simulated with good accuracy by a CFD model (developed using the open source OpenFOAM framework) when compared to measured samples. Large-scale clarifier trials were used to assess the effectiveness of the UARP controller with an attached in situ acoustic backscatter array (ABA). The ABA consisted of 8 transducers angled down at 20 deg. to the vertical, pulsing at 2-2.5 MHz in cycles (covering 1.8 m total depth). In comparison to a range of sampled concentrations across the length of the clarifier, results confirmed that the UARP-ABA system was highly successful at characterizing these complex flows over several hours of operation, allowing visualization of depth-wise bulk density changes and bed build-up over time. Collectively, this work highlights the capability of these tools to monitor and optimize a number of current waste transfer operations at Sellafield, with further potential applications for solid-liquid separations across waste, water treatment and minerals engineering industries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Materials Identification System for 200 liter/55 gallon Waste Drums using Computer Tomographic Techniques - 17162

In 2014 a thermal runaway event within a drum in-situ at the Waste Isolation Pilot Plant (WIPP) facility near Carlsbad, NM, resulted in the drum being breached and activity released. This emphasized the need for a better determination of drum contents and identification of material combinations and in-drum conditions. Enhanced characterization is required to identify and quantify the presence of absorbent materials, determine whether they are organic or inorganic, locate their position in the drum with respect to other materials, and quantify how much liquid is being held in the absorbent material. As part of the waste characterization process to ensure compliance with the WIPP waste acceptance criteria, drums are currently inspected by Real Time Radioscopic (RTR) x-ray to identify prohibited items such as free liquids and un-punctured aerosol cans, but that technique cannot satisfy these newly identified requirements, and a new approach to non-destructive examination was required. This approach complements the current RTR inspection with Cone-Beam Area Detector Computed Tomography (CT) scanning of the whole drum, followed by more-quantitative CT scanning of specific areas of concern, using a highly collimated Linear Detector Array. The density of representative waste drums required x-ray energies of up to 9 MeV, which meant that sorting of materials by their photo-electric absorption is not viable as this occurs in the 80 - 180 keV range. However, a notable characteristic of higher energy imaging in this context is the general correspondence between material density and total attenuation in the 1.5-4.0 MeV range due to Compton Scatter. Development of the imaging chain, comprising the x-ray source and the detectors was critical to achieving the required system performance. Theoretical analysis of the energy spectra of the 9 MeV beam, and filtration of the beam to remove the lower energies which cause beam hardening artifacts in the image was required. Theoretical evaluation of the required efficiency and image acquisition performance of area detectors and linear array detectors, was carried out to complete the selection of equipment. A system of density calibration rods was designed and, together with the selected equipment, was configured and used to acquire cone-beam CT data and line-scan CT data by scanning a 55 gallon drum containing surrogate waste. From the acquired data, CT reconstruction was performed and the basis of algorithms to identify materials were developed. Following this, a series of blind tests were done using a range of surrogate materials the contents of which were unknown to the system operators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Development of a Testbed for Pipeline Flushing - 20425

Pipelines which carry high-level waste within the DOE complex should be properly flushed in the cases where stationary or moving beds of solid sediment occur, or lines are prone to hydrogen gas buildup. Present guidelines establish a minimum for the amount of water (flush volume) and flush velocity values used for post-transfer flushing operations to achieve satisfactory cleaning of pipelines. However, further studies are needed to find optimal operating velocity modes/values in order to minimizing the flush volume and consequent downstream waste. These efforts are significantly helpful to DOE waste remediation sites by preserving tank storage, preventing additional waste processing, and minimizing dilution and changes in waste chemistry. An experimental test loop was recently developed for study of non-Newtonian slurry flushing at the Florida International University (FIU). This loop was designed to create sediment beds of various materials and bed heights and investigate parameters that effect the efficiency of flushing operations. The objective is to find flush velocity values/modes which lead to satisfactory cleaning of transport lines with a minimum amount of water usage. A 165 ft experimental test loop made of 3-inch carbon steel pipes was constructed and used in initial testing campaigns. This pipe loop was equipped with real-time monitoring instruments to record variation of pressure, density, mass flow rate, and temperature during testing. Post-flush in-situ no-flow evaluations were performed using ultrasonic, endoscopy, and visual inspection. Additional post-flush evaluations were conducted using filtration and density monitoring of residuals within a designated circulations loop. These provisions enable comparison of pipeline cleanness before the start and after termination of flushing operations for efficiency evaluation. This paper presents efforts associated with flushing of kaolin-water mixtures at different concentrations and with various flush volume and flush velocity values/modes. In initial testing, efforts focused on creating repeatable sediment beds inside the pipeline in both fully-flooded (no pre-flush drainage) and gravity-drained (pre-flush drained) systems. Bed characterization (i.e., measurement of sediment height and solids concentration) was conducted to ensure about consistency of initial conditions between tests. Preliminary results for flushing study of 10% vol. kaolin-water mixtures starting from a fully-flooded initial condition showed that flushing with 1-line volume (65 gallons) could remove up to 72 percent of solids from the pipeline. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

The Impact of Ethylenediaminetetraacetic Acid (EDTA) on the Sorption of Nd(III), Th{sup (IV)}, and U(VI) onto Dolomite in WIPP-Relevant Brines, GWB and ERDA-6 - 20356

Storage of legacy, transuranic waste at the Waste Isolation Pilot Plant (WIPP) requires accurate risk models for performance assessments. These models must predict the long-term fate and transport of contaminants in the WIPP and surrounding environment. Laboratory support is necessary for the development of an accurate model through better understanding of contaminant behavior in the WIPP. The WIPP is a deep geologic repository that is characterized by high ionic strength due to evaporite deposits in the Salado formation (porewaters up to 7.4 M). High concentrations of sodium chloride (NaCl), calcium chloride (CaCl{sub 2}), and magnesium chloride (MgCl{sub 2}) are particularly of concern due to their potential impact on solubility and ultimately transport of contaminants of concern. Additionally, ligands capable of forming strong complexes with metals are found in significant concentrations in the WIPP environment due to their disposal from waste processing and formation during degradation of repository components (e.g. cellulose degradation by calcium hydroxide in cement). Consequently, there is a need to investigate the fate and transport of contaminants in the WIPP environment with a focus on the effects of high ionic strength and strongly coordinating ligands. Contaminants of high concern include the actinides (U(VI), Pu(IV), and Am(III)). Laboratory support of the WIPP modeling efforts was provided by this research. A series of batch experiments were utilized to study the impact of ionic strength, contaminant concentration, and the presence of ethylenediaminetetraacetic acid (EDTA) on the sorption of Nd(III), Th(IV), and U(VI) onto dolomite, a carbonate mineral found within the Culebra formation. The Culebra formation is the most transmissive layer above the WIPP, making transport of contaminants the most likely route in previous risk assessments. Two brines specifically designed to imitate conditions within the WIPP environment were used in these experiments. The U.S. Energy Research and Development Administration Well 6 (ERDA-6, 95%) brine simulates a low Mg environment, while the generic weep brine (GWB, 95%) simulates the WIPP environment with a high Mg concentration. Mg is added to the repository in the form of MgO as an engineered barrier and is naturally present in some formations. It is expected to scavenge carbonate from solutions decreasing formation of the highly soluble and competitive uranium-carbonate complexes. EDTA is used during manipulations of nuclear materials due to its ability to form strong, stable complexes with the actinides and may be found in high concentrations (up to 0.08 mM) within the WIPP as waste canisters degrade. Further, it may provide additional avenues for unwanted transport of contaminants. Nd(III) and Th(IV) were used as stable analogues of Am(III) and Pu(IV), respectively, to represent the most common oxidation states of actinides in the WIPP environment in addition to U(VI). The potential for colloid formation was also investigated with low and high concentrations of contaminants (10 and 1000 μg/L) using three size-dependent particle separation steps. The results of this research provide a better understanding of potential behavior of actinides in the presence of EDTA and within WIPP-relevant brines. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Panel Session 70: Nuclear Power Plant Waste Management, LLRW Processing Issues

This panel focused on senior utility managers discussing issues with commercial LLW processors. LLW processors had an opportunity to update utilities/NPPs on new or different services or technologies being offered to assist in their radioactive waste management goals, changes to their facility license, revisions to their acceptance guidelines, and emerging compliance issues. Panelists with presentations: UniTech - 'Expanding to Reduce Costs' LLW Processor Update (Gregg Johnstone); Veolia Nuclear Solutions Federal Services - Alaron Nuclear Services (Dave Garber); EnergySolutions Tennessee Processing Facilities (Fred Schulz); Complex Waste Management Issues Services and Treatment Facilities (George M. Taylor)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Effect of Cr(III) Adsorption on the Dissolution of Boehmite Nanoparticles in Caustic Solution

Incorporation of relatively minor impurity metals onto metal (oxy)hydroxides can strongly impact solubility. In complex highly alkaline multicomponent tank waste such as that at the Hanford Nuclear Reservation, tests indicate that the surface area-normalized dissolution rate of boehmite ($γ$-AlOOH) nanomaterials is at least an order of magnitude lower than that predicted for the pure phase. In this work we examine the dissolution kinetics of boehmite coated by adsorbed Cr(III), which adheres at saturation coverages as sparse chemisorbed monolayer clusters. Using 40 nm boehmite nanoplates as a model system, temperature-dependent dissolution rates of pure versus Cr(III)-adsorbed boehmite showed that the initial rate for the latter is consistently several times lower, with an apparent activation energy 16 kJ·mol -1 higher. Although the surface coverage is only around 50%, solution analytics coupled to multimethod solids characterization reveals a phyicochemical armoring effect by adsorbed Cr(III) that substantially reduces the number of dissolution-active sites on particle surfaces. Such findings could help improve kinetics models of boehmite and/or metal ion adsorbed boehmite nanomaterials, ultimately providing a stronger foundation for the development of more robust complex radioactive liquid waste processing strategies.

54 ENVIRONMENTAL SCIENCES↗