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DOE OSTI2020

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Geo-polymer solidification technology approved by Czech / Slovak nuclear authority to immobilise NPP resins and sludge waste - 15555

This paper presents the features and performance record of the geo-polymer solidification technology, SIAL{sup R}, licensed for use by both the Slovakian (UJD SR) and Czech Nuclear (SUJB) regulators. An overview of its performance at Bohunice in Slovakia, where it has been used to successfully treat resins and sludge wasteforms will be provided along with references to other plants where it has been used. More recently, geo-polymers have been noted as an immobilization technology which shows potential for resins produced by treatment of radioactive contaminated water resulting from the clean-up of the Fukushima Daiichi Plant. The Nuclear Power Plant A-1 located in Jaslovske Bohunice was completed in 1972 and operated for 5 years until two accidents in 1976 and 1977. After the second accident in February 1977 (INES level 4) the Nuclear Power Plant was permanently shut down for decommissioning. Fuel assemblies and fuel cladding damaged in the accidents led to significant Sr-90, Cs-137 and transuranic contamination. As a consequence of the long-term corrosion of the barrier's material, highly contaminated sludge accumulated which could not be effectively immobilized using conventional methods such as cementation or bitumen treatment due to adverse physical-chemical properties and high specific activity of the waste (Cs-137). This challenge led to AMEC developing and licensing the geo-polymer, SIAL{sup R}. Today, this geo-polymer is one of the most proven and widely used for on-site solidification of radioactive materials such as sludge, resins, sorbents and organic liquids. There are numerous benefits to using this geo-polymer for waste immobilisation. It can be used at room temperature and can incorporate up to approximately four times as much waste (sludge, resin) as a Portland cement matrix equivalent depending on the waste being treated. In addition, it is characterized by excellent mechanical and physical properties, compared with the earlier generation techniques. This includes higher mechanical strength, lower leachability, low volatility, posing a low fire risk and excellent physical stability in the presence of frost and water (no distortion or cracking). Dedicated facilities to receive and encapsulate waste from different locations can be created but this geo-polymer can also provide efficient and practical on-site treatment of radioactive waste streams. This is achievable because the equipment used to encapsulate waste using SIAL{sup R} is also modular, flexible and versatile. At Bohunice, the facility for encapsulation was designed to occupy existing areas of the plant in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles (ROVs) and for the waste to be immediately treated on site without the need for transportation saving both time and resource. To achieve this, some of the rooms and areas of the building were assessed and modified to ensure that the waste package flow was suitable whilst ensuring that appropriate safety measures were in place such as adequate shielding to reduce or entirely eliminate dose uptake of workers. To date, SIAL{sup R} has been used to successfully immobilize approximately 500 tons of waste which includes sludge and resin from Bohunice, Slovakia and approximately 250 m{sup 3} of spent ion exchange resins from tanks on site at the Dukovany nuclear plant, Czech Republic. It now has a track record of over 15 years which includes on-going research and development. The International missions WANO and OSART evaluated the SIAL{sup R} matrix technology at NPP Dukovany as an example of good practice. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Theory of Carbon Dioxide Absorption by Hanford Waste - 18208

Hanford nuclear waste has a very high pH, and Hanford tanks are ventilated with air. Air has CO{sub 2(g)} in it, which is an acid, so it has some affinity for the waste. The absorption of CO{sub 2} by the waste is important because CO{sub 2} absorption consumes hydroxide at the liquid-air interface, and the hydroxide is important for inhibiting tank corrosion. NaOH is periodically added to some tanks in order to mitigate the surface acidification. The NaOH added then becomes part of the waste and must be treated as waste, increasing the cost of the mission. The Hobbs Equation, developed at the Savannah River site, is used at Hanford to predict CO{sub 2} absorption from the tank ventilation system. In this study, the Hobbs Equation functional form is theoretically evaluated from thermodynamic calculations of CO{sub 2} absorption, and diffusion rates in the waste surface. This study theoretically justifies the empirical equation reported by Hobbs. Thermodynamic calculations performed using the Environmental Simulation Program (ESP) software on a typical double-shell tank waste composition indicate that essentially 100 % of CO{sub 2} that contacts the waste will react with the waste whenever the pH is above 10. In the 1990's, a substantial amount of data was collected on the concentration of CO{sub 2} in the off-gas of ventilated double-shell tanks at Hanford. This historic data was mined in order to determine the quantity of CO{sub 2} absorbed by the waste. The results of this data mining activity determined that tanks below approximately pH 10 (i.e., tank 241-AN-107) will show limited CO{sub 2} absorption, consistent with both the Hobbs model and the theory developed here. Two other tank with higher pH showed contrasting CO{sub 2} absorption. One showed substantial CO{sub 2} absorption, and the other showed minimal CO{sub 2} absorption despite the high waste pH in one tank, presumably because there was minimal contact between the waste surface and most of the air flow. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 20326

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering scale. Veolia Nuclear Solutions designed, installed, and commissioned in September 2018, at Perma-Fix Northwest in Richland Washington, a 10-metric ton full-scale GeoMelt unit (GeoMelt{sup R} Richland) for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt{sup R} Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

GHG emission factors developed for the collection, transport and landfilling of municipal waste in South African municipalities

Highlights: ► An average GHG emission factor for the collection and transport of municipal solid waste in South Africa is calculated. ► A range of GHG emission factors for different types of landfills (including dumps) in South Africa are calculated. ► These factors are compared internationally and their implications for South Africa and developing countries are discussed . ► Areas for new research are highlighted. - Abstract: Greenhouse gas (GHG) emission factors are used with increased frequency for the accounting and reporting of GHG from waste management. However, these factors have been calculated for developed countries of the Northern Hemisphere and are lacking for developing countries. This paper shows how such factors have been developed for the collection, transport and landfilling of municipal waste in South Africa. As such it presents a model on how international results and methodology can be adapted and used to calculate country-specific GHG emission factors from waste. For the collection and transport of municipal waste in South Africa, the average diesel consumption is around 5 dm{sup 3} (litres) per tonne of wet waste and the associated GHG emissions are about 15 kg CO{sub 2} equivalents (CO{sub 2} e). Depending on the type of landfill, the GHG emissions from the landfilling of waste have been calculated to range from −145 to 1016 kg CO{sub 2} e per tonne of wet waste, when taking into account carbon storage, and from 441 to 2532 kg CO{sub 2} e per tonne of wet waste, when carbon storage is left out. The highest emission factor per unit of wet waste is for landfill sites without landfill gas collection and these are the dominant waste disposal facilities in South Africa. However, cash strapped municipalities in Africa and the developing world will not be able to significantly upgrade these sites and reduce their GHG burdens if there is no equivalent replacement of the Clean Development Mechanism (CDM) resulting from the Kyoto agreement. Other low cost avenues need to be investigated to suit local conditions, in particular landfill covers which enhance methane oxidation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Life cycle assessment of electronic waste treatment

Highlights: • Life cycle assessment of electronic waste recycling is quantified. • Key factors for reducing the overall environmental impact are indentified. • End-life disposal processes provide significant environmental benefits. • Efficiently reduce the improper disposal amount of e-waste is highly needed. • E-waste incineration can generate significant environmental burden. - Abstract: Life cycle assessment was conducted to estimate the environmental impact of electronic waste (e-waste) treatment. E-waste recycling with an end-life disposal scenario is environmentally beneficial because of the low environmental burden generated from human toxicity, terrestrial ecotoxicity, freshwater ecotoxicity, and marine ecotoxicity categories. Landfill and incineration technologies have a lower and higher environmental burden than the e-waste recycling with an end-life disposal scenario, respectively. The key factors in reducing the overall environmental impact of e-waste recycling are optimizing energy consumption efficiency, reducing wastewater and solid waste effluent, increasing proper e-waste treatment amount, avoiding e-waste disposal to landfill and incineration sites, and clearly defining the duties of all stakeholders (e.g., manufacturers, retailers, recycling companies, and consumers)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Converting campus waste into renewable energy – A case study for the University of Cincinnati

Highlights: • A case study to show the benefits of waste-to-energy projects at a university. • Evaluated the technical and economic feasibilities as well as GHG reduction. • A tool for other universities/communities to evaluate waste-to-energy projects. - Abstract: This paper evaluates the implementation of three waste-to-energy projects at the University of Cincinnati: waste cooking oil-to-biodiesel, waste paper-to-fuel pellets and food waste-to-biogas, respectively. The implementation of these waste-to-energy (WTE) projects would lead to the improvement of campus sustainability by minimizing waste management efforts and reducing GHG emissions via the displacement of fossil fuel usage. Technical and economic aspects of their implementation were assessed and the corresponding GHG reduction was estimated. Results showed that on-site implementation of these projects would: (1) divert 3682 L (974 gallons) of waste cooking oil to 3712 L (982 gallons) of biodiesel; (2) produce 138 tonnes of fuel pellets from 133 tonnes of waste paper (with the addition of 20.75 tonnes of plastics) to replace121 tonnes of coal; and (3) produce biogas that would be enough to replace 12,767 m{sup 3} natural gas every year from 146 tonnes of food waste. The economic analysis determined that the payback periods for the three projects would be 16 months for the biodiesel, 155 months for the fuel pellet, and 74 months for the biogas projects. The reduction of GHG emission from the implementation of the three WTE projects was determined to be 9.37 (biodiesel), 260.49 (fuel pellets), and 11.36 (biogas) tonnes of CO{sub 2}-eq per year, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

System Performance Testing of the Pulse-Echo Ultrasonic Instrument for Critical Velocity Determination during Hanford Tank Waste Transfer Operations - 13584

The delivery of Hanford double-shell tank waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is governed by specific Waste Acceptance Criteria that are identified in ICD 19 - Interface Control Document for Waste Feed. Waste must be certified as acceptable before it can be delivered to the WTP. The fluid transfer velocity at which solid particulate deposition occurs in waste slurry transport piping (critical velocity) is a key waste acceptance parameter that must be accurately characterized to determine if the waste is acceptable for transfer to the WTP. Washington River Protection Solutions and the Pacific Northwest National Laboratory have been evaluating the ultrasonic PulseEcho instrument since 2010 for its ability to detect particle settling and determine critical velocity in a horizontal slurry transport pipeline for slurries containing particles with a mean particle diameter of =14 micrometers (μm). In 2012 the PulseEcho instrument was further evaluated under WRPS' System Performance test campaign to identify critical velocities for slurries that are expected to be encountered during Hanford tank waste retrieval operations or bounding for tank waste feed. This three-year evaluation has demonstrated the ability of the ultrasonic PulseEcho instrument to detect the onset of critical velocity for a broad range of physical and rheological slurry properties that are likely encountered during the waste feed transfer operations between the Hanford tank farms and the WTP. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Conceptual Design for a Short-Lived Low- and Intermediate-Level Waste Repository at the Al Tuwaitha Nuclear Center, Baghdad, Iraq - 17189

The European Commission under their Instrument for Nuclear Safety Cooperation (INSC) has initiated a project supporting Iraq's efforts for the safe disposal of short lived low- and intermediate-level radioactive waste (SL-LILW) arising from the decommissioning of Iraqi nuclear installations. The INSC project includes both the development of an initial conceptual level basic design and preliminary safety assessment (PSAR) report and the subsequent detailed engineering design and supporting intermediate safety assessment report (ISAR). The project is being executed by a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany; JAVYS a.s. of Slovakia and MCM Environmental Services Ltd of the United Kingdom. The current paper focuses on the development of the conceptual level basic design. In previous decades Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were located. The facilities suffered significant damage from bombing and subsequent looting during the first Gulf War in 1991. Currently, the Ministry of Science and Technology (MoST) is the radioactive waste management operator in Iraq and also the license holder for the ten Iraqi nuclear sites, all of them heavily damaged. Because of the ongoing security concerns in Iraq MoST has selected the Al Tuwaitha Nuclear Center as the preferred and best suited location for development of an SL-LILW repository. As an initial step in development of the conceptual design various repository design options for the site were examined and assessed. The assessment considered similar facilities that have been implemented internationally. At the conclusion of the assessment and in conjunction with the Iraqi Beneficiaries from MoST, a near-surface vault design was selected for the facility. The design of the facility relies on experience gained from existing low-level waste repositories at Centre de l'Aube in France and the El Cabril solid radioactive waste disposal facility in Spain as well as on more recent experience gained in the design of the Bulgaria National Disposal Facility (NDF) under construction near the Kozloduy Nuclear Power Plant. Similar to these reference designs, the concept for the Al Tuwaitha Disposal Facility (ATDF) will rely on a series of reinforced concrete vaults into which containerised waste will be placed. Each vault will consist of two chambers designed specifically to hold the waste. Central drainage points in each waste chamber will lead to an infiltration water collection system designed to detect any failures in the waste containment. Each disposal vault will be sealed after filling by installation of a reinforced concrete closure slab. Once all of the waste has been disposed of the site will be decommissioned and a multilayer protective cover will be installed over the sealed disposal vaults. The multilayer cover will act to further isolate the waste from the environment by limiting the access of water to the waste. Due in part to the presence of shallow groundwater at the site and to avoid potential damage to the facility by flooding from the nearby Tigris River the concept for the Al Tuwaitha Disposal Facility requires construction of the disposal vaults on a raised platform several meters above ground level. Construction of this platform will be integrated into the conceptual safety design and materials will be selected which further enhance the isolation function of the disposal system by hindering or delaying the release of radionuclides to the environment. The infiltration water collection system, which will be constructed in a specifically designed accessible gallery embedded and fully integrated into the platform, will provide access to inspection points located beneath each disposal chamber. The ATDF when completed will be a further demonstration of the adaptability of the disposal vault concept for use in isolating SL-LILW from the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Recycling process for recovery of gallium from GaN an e-waste of LED industry through ball milling, annealing and leaching

Waste dust generated during manufacturing of LED contains significant amounts of gallium and indium, needs suitable treatment and can be an important resource for recovery. The LED industry waste dust contains primarily gallium as GaN. Leaching followed by purification technology is the green and clean technology. To develop treatment and recycling technology of these GaN bearing e-waste, leaching is the primary stage. In our current investigation possible process for treatment and quantitative leaching of gallium and indium from the GaN bearing e-waste or waste of LED industry dust has been developed. To recycle the waste and quantitative leaching of gallium, two different process flow sheets have been proposed. In one, process first the GaN of the waste the LED industry dust was leached at the optimum condition. Subsequently, the leach residue was mixed with Na{sub 2}CO{sub 3}, ball milled followed by annealing, again leached to recover gallium. In the second process, the waste LED industry dust was mixed with Na{sub 2}CO{sub 3}, after ball milling and annealing, followed acidic leaching. Without pretreatment, the gallium leaching was only 4.91 w/w % using 4 M HCl, 100 °C and pulp density of 20 g/L. After mechano-chemical processing, both these processes achieved 73.68 w/w % of gallium leaching at their optimum condition. The developed process can treat and recycle any e-waste containing GaN through ball milling, annealing and leaching. - Highlights: • Simplest process for treatment of GaN an LED industry waste developed. • The process developed recovers gallium from waste LED waste dust. • Thermal analysis and phase properties of GaN to Ga{sub 2}O{sub 3} and GaN to NaGaO{sub 2} revealed. • Solid-state chemistry involved in this process reported. • Quantitative leaching of the GaN was achieved.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2019

Is Radioactive Tank Waste a Bingham Fluid?

Slurries and sludges across the United States Department of Energy (DOE) complex rank among the most rheologically interesting. Their composition is heterogeneous, spanning a very broad range of particle sizes, densities, and interparticle forces. All exhibit shear thinning, some have yield stresses, and many are thixotropic. Despite the variety, these complex fluids are often represented using the historic Bingham fluid model, which fits higher shear rate data to a simple straight line. The intercept provides a yield stress, which has been a key design parameter in construction of large-scale waste processing facilities. However, many radioactive wastes are simply not Bingham fluids, which extrapolate but poorly across low to intermediate shear rates that are characteristic of typical processing conditions. Indeed, processing shear rates as high as 200 1/s, which has been a typical minimum shear rate used in fitting, are seldom encountered in nuclear waste processing. Therefore, more realistic rheological models are necessary to accurately predict waste processing performance. PNNL recently re-evaluated the rheology of reconstituted Hanford REDOX sludge waste against a wide variety of rheological models including the Bingham, Cross, Cross with yield stress, Carreau, biviscous, Herschel-Bulkley (which includes a power law dependence), Casson, and Gay models. They found that all of the models provided a closer fit than the Bingham model and that the biviscous model and Cross with yield stress model were convincing. However, reconstituted Hanford REDOX sludge waste is but one type of DOE waste and a direct contrast and comparison of these three models against undiluted, unmixed tank waste (actual not simulant) has not been performed previously. Therefore, the purpose of this paper is to evaluate the rheology of actual tank waste with these more accurate rheological models. In this paper, we evaluate select rheological data for slurry samples from Hanford’s AZ-101, AZ-102, and SY-101 waste tanks. In each of these cases, we find that Cross’ model with yield stress and the biviscous model significantly outperform the Bingham fluid model. Furthermore, the AZ-101 data also shows that the shear stress peak that startup significantly exceeds the Bingham yield stress, which is commonly observed in the incipient moments of rheological measurements on simulants. In contrast, Cross’ model may empirically accommodate an initial spike in shear stress at modest shear rates. These are important observations because computational and analytical fluid dynamics simulations rely on rheological constitutive models for accurately and conservatively predict waste processing performance. These findings suggest the need for better rheological modeling of and validation against radioactive waste.

Rheology, Radioactive Waste↗

DOE OSTI2013

Review Of Rheology Models For Hanford Waste Blending

The area of rheological property prediction was identified as a technology need in the Hanford Tank Waste – waste feed acceptance initiative area during a series of technical meetings among the national laboratories, Department of Energy-Office of River Protection, and Hanford site contractors. Meacham et al. delivered a technical report in June 2012, RPP-RPT-51652 “One System Evaluation of Waste Transferred to the Waste Treatment Plant” that included estimating of single shell tank waste Bingham plastic rheological model constants along with a discussion of the issues inherent in predicting the rheological properties of blended wastes. This report was selected as the basis for moving forward during the technical meetings. The report does not provide an equation for predicting rheological properties of blended waste slurries. The attached technical report gives an independent review of the provided Hanford rheological data, Hanford rheological models for single tank wastes, and Hanford rheology after blending provided in the Meacham report. The attached report also compares Hanford to SRS waste rheology and discusses some SRS rheological model equations for single tank wastes, as well as discussing SRS experience with the blending of waste sludges with aqueous material, other waste sludges, and frit slurries.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Data sharing report characterization of the surveillance and maintenance project miscellaneous process inventory waste items Oak Ridge National Laboratory, Oak Ridge, TN

The U.S. Department of Energy (DOE) Oak Ridge Office of Environmental Management (EM-OR) requested Oak Ridge Associated Universities (ORAU), working under the Oak Ridge Institute for Science and Education (ORISE) contract, to provide technical and independent waste management planning support under the American Recovery and Reinvestment Act (ARRA). Specifically, DOE EM-OR requested ORAU to plan and implement a sampling and analysis campaign to target certain items associated with URS|CH2M Oak Ridge, LLC (UCOR) surveillance and maintenance (S&M) process inventory waste. Eight populations of historical and reoccurring S&M waste at the Oak Ridge National Laboratory (ORNL) have been identified in the Waste Handling Plan for Surveillance and Maintenance Activities at the Oak Ridge National Laboratory, DOE/OR/01-2565&D2 (WHP) (DOE 2012) for evaluation and processing for final disposal. This waste was generated during processing, surveillance, and maintenance activities associated with the facilities identified in the process knowledge (PK) provided in Appendix A. A list of items for sampling and analysis were generated from a subset of materials identified in the WHP populations (POPs) 4, 5, 6, 7, and 8, plus a small number of items not explicitly addressed by the WHP. Specifically, UCOR S&M project personnel identified 62 miscellaneous waste items that would require some level of evaluation to identify the appropriate pathway for disposal. These items are highly diverse, relative to origin; composition; physical description; contamination level; data requirements; and the presumed treatment, storage, and disposal facility (TSDF). Because of this diversity, ORAU developed a structured approach to address item-specific data requirements necessary for acceptance in a presumed TSDF that includes the Environmental Management Waste Management Facility (EMWMF)—using the approved Waste Lot (WL) 108.1 profile—the Y-12 Sanitary Landfill (SLF) if appropriate; EnergySolutions Clive; and the Nevada National Security Site (NNSS) (ORAU 2013b). Finally, the evaluation of these wastes was more suited to a judgmental sampling approach rather than a statistical design, meaning data were collected for each individual item, thereby providing information for item-byitem disposition decisions. ORAU prepared a sampling and analysis plan (SAP) that outlined data collection strategies, methodologies, and analytical guidelines and requirements necessary for characterizing targeted items (ORAU 2013b). The SAP described an approach to collect samples that allowed evaluation as to whether or not the waste would be eligible for disposal at the EMWMF. If the waste was determined not to be eligible for EMWMF disposal, then there would be adequate information collected that would allow the waste to be profiled for one of the alternate TSDFs listed above.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Vitrification of Hanford Tank Waste 241-AP-107 in a Continuous Laboratory-Scale Melter

Low-activity waste (LAW) stored in underground tanks on the Hanford Site in Washington State is planned to be filtered for solids removal and processed through ion exchange columns for cesium removal. These pretreatment steps will allow the waste to be transferred to the Hanford Tank Waste Treatment and Immobilization Plant’s LAW Facility for immobilization into glass. The liquid waste will be combined with glass-forming chemicals (GFCs) to form a waste feed slurry that can be fed to electric melters for vitrification. The process of continuously converting the aqueous feed slurry into a melt is dynamic and includes multiple reactions, degassing, and dissolution processes that depend on heat from the melt below. In this conversion process, waste components are partitioned into one of two streams: glass and off-gas. Washington River Protection Solutions has requested processing information and chemical information associated with these waste products for actual waste from Hanford tank 241-AP-107 (referred to herein as AP-107). To acquire this type of information, a small-scale melter system was designed that would not require high volumes of input waste or the large resource commitment of a full-scale melter system, while also providing dynamic information that would be difficult to determine from batch reactions in a crucible system. A continuous laboratory-scale melter (CLSM) has been designed to operate with a continuous feeding process while periodically pouring glass product and collecting off-gas. The CLSM vessel has been sized to collect the relevant process and chemical information from obtainable volumes of AP-107 waste samples. A total 8.6 L of actual AP-107 tank waste were received after filtering for solids removal and ion exchange for cesium removal. This volume of AP-107 waste was mixed with GFCs to form an estimated 11.8 L of melter feed slurry. A mass of 7.01 kg of glass product were poured from the CLSM vessel during 10.07 hours of charging the AP-107 melter feed, indicating that about 15.0 kg of melter feed was vitrified into glass. The CLSM system was also designed with the capability to fully divert the flow of off-gas produced in the CLSM vessel to a sampling line that could capture the volatile species of interest, such as technetium-99 ( 99 Tc). This novel sampling system avoided the difficulties of slipstream sampling and could be activated once the feeding reached a steady state within the CLSM vessel. Off-gas product samples captured via high-efficiency particulate air (HEPA) filters using this method, as well as selected glass product samples were sent to an analytical lab for chemical analysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

FY2019 Performance Assessment Annual Review for the E-Area Low-Level Waste Facility

The Savannah River Site (SRS) E-Area Low-Level Waste Facility (ELLWF) consists of six types of disposal units described in the Performance Assessment (PA) (WSRC, 2008): Low Activity Waste Vault (LAWV), Intermediate Level Vault (ILV), Trenches [Slit Trenches (STs), Engineered Trenches (ETs), and Component-in-Grout (CIG) Trenches], and Naval Reactor Component Disposal Areas (NRCDAs). The ELLWF is a part of the Solid Waste Management Facility (SWMF). SWMF is managed and operated by the SRS Management and Operations prime contractor, Savannah River Nuclear Solutions (SRNS). Within SRNS, the Solid Waste Management (SWM) organization is responsible for operating the SWMF, and the Savannah River National Laboratory (SRNL) is the technical agency responsible for preparing and maintaining the PA. SWMF operations have been performed at SRS since 1952. The mission of the SWMF is to provide storage, processing, disposal, and shipment of radioactive, hazardous, and mixed waste. The SWMF is committed to treat, store, and dispose of these waste products in a manner that protects the environment and the health and safety of the facility worker, the co-located worker, and the offsite general public. Wastes handled in the SWMF include low level waste, transuranic waste, hazardous waste, Toxic Substances Control Act waste, and mixed waste (containing both hazardous and radioactive constituents).

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DOE OSTI2020

FY2019 Performance Assessment Annual Review for the E-Area Low-Level Waste Facility

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DOE OSTI2019

Cesium Ion Exchange Testing Using a Three-Column System with Crystalline Silicotitanate and Hanford Tank Waste 241-AP-107

Washington River Protection Solutions (WRPS) is working to support early production of immobilized low-activity waste by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) facility. This goal incorporates the design of a Tank-Side Cesium Removal (TSCR) system, which in effect filters tank waste supernate and processes it through a cesium ion exchange medium to remove 137Cs. The 137Cs-depleted product is sent to the WTP for vitrification. Rovira et al. (2018) reported the successful Cs removal from Hanford tank 241-AP-107 waste (hereafter called AP-107) using crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC (product IONSIVTM R9140-B, Batch 2081000057) in a lead-lag column system. Since that testing, WRPS has changed the column design to a three-column system. WRPS requested a repeated study with AP-107 tank waste using a different batch of IONSIVTM R9140-B, Lot 2002009604, in a three-column format. A small-scale test platform was constructed to demonstrate solids filtration, Cs removal, and LAW vitrification and installed at Pacific Northwest National Laboratory. This report describes the small-scale ion exchange testing system component of the test platform and the processing of approximately 8.55 L of filtered AP-107 supernate using CST, Lot 8056262-999, that passed through a 25-mesh filter. This report also describes the Cs ion exchange batch contact testing to determine the Cs distribution coefficient and the maximum Cs loading capacity in the AP-107 tank waste matrix. Batch contact testing helps to evaluate CST performance on tank waste supernate and is often used prior to processing tank waste through ion exchange columns to assess some performance metrics (e.g., 50% Cs breakthrough). Batch contact tests were performed with the filtered AP-107 tank waste at four Cs concentrations at a phase ratio of 200 (liquid volume to exchanger mass). The distribution coefficient (K d ) at the equilibrium condition of 9.2 µg Cs/mL (AP-107 feed condition) was determined to be 806 mL AP 107/g CST. With a CST bed density of 1.01 g/mL (<25 mesh CST), this K d corresponded to a predicted 50% Cs breakthrough of 814 bed volumes (BVs). The Cs load capacity at the equilibrium feed condition was determined to be 8.3 mg Cs/g dry CST. The three-column format testing was prototypic to the intended TSCR operations, albeit on a small scale with 10-mL CST beds. Flowrate was adjusted to match the CST contact time expected for the full-scale operation, i.e., matched bed volumes per hour (BV/h) flow rate. The feed was processed downflow through the lead column, then through the middle column, and then through the polish column. Loading continued at an average of 1.88 BV/h (1.83 to 2.04 BV/h range) until the entire available AP-107 feed was processed. The Cs-decontaminated product will be provided for vitrification testing. The lead column only reached 45% Cs breakthrough after processing 855 BVs of feed. The 50% Cs breakthrough was extrapolated to occur at ~900 BVs. This extrapolated 50% Cs breakthrough value differed from the batch contact estimate (814 BVs) by 13%. The waste acceptance criteria (WAC) or contract limit for the WTP LAW vitrification facility is 3.18E-5 Ci 137 Cs per mole Na. For the AP-107 tank waste, only 0.127% of the influent 137 Cs concentration may be delivered to the WTP; this required a Cs decontamination factor of 787. The Cs effluent from the middle column reached the WAC after processing 590 BVs. Cs breakthrough from the polish column began at 590 BVs, reaching 6.3E-3 µCi/mL or 4.3E-3% Cs breakthrough after processing 844 BVs of feed.

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DOE OSTI2022

Follow-on Report of Analysis of Approaches to Supplemental Treatment of Low-Activity Waste at the Hanford Nuclear Reservation (Vol. I)

The Hanford Site, in southeast Washington State, is preparing to disposition approximately 56,000,000 gallons (56 Mgal) of radioactive and chemically hazardous wastes currently stored in underground tanks at the site. Tank wastes will be divided into a high-activity fraction and a low-activity fraction for subsequent treatment and disposition. A waste processing and treatment facility, the Waste Treatment and Immobilization Plant (WTP), will include the high-level waste (HLW) vitrification facility (WTP HLW Vitrification Facility) for immobilizing the high-activity fraction and a low-activity waste (LAW) vitrification facility (WTP LAW Vitrification Facility) for immobilizing the low-activity fraction. Both facilities will use vitrification technology to immobilize the Hanford tank wastes in a glass waste form. The volume of LAW to be treated and disposed of following waste retrieval and WTP operations will exceed the planned processing capacity of the WTP LAW Vitrification Facility. ORP-11242,-River Protection Project System Plan, estimates a shortfall in LAW treatment capacity of approximately 56 Mgal, approximately 50% of the projected LAW volume. To maintain the planned tank waste processing mission schedule, the U.S. Department of Energy (DOE) will require additional LAW treatment capacity (termed “supplemental LAW”) external to the WTP process. LAW must be solidified by a treatment technology before the waste can be permanently disposed of in an approved DOE on-site disposal facility or a commercial (state or U.S. Nuclear Regulatory Commission [NRC-licensed]) off-site mixed low-level waste disposal facility. A decision on the approach to supplemental LAW treatment, processing, and disposal has not yet been made

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