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DOE OSTI2025

Evaluation of Technologies to Mitigate the Presence of Gaseous Elemental Mercury in Waste Disposal Containers

A study was conducted to evaluate sorbent technologies that can mitigate the presence of elemental mercury (Hg⁰) in waste containers for mercury-contaminated debris (MCD). Decontamination and demolition (D&D) activities at the Y-12 National Security Complex (Y-12) and other U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR) facilities generate MCD requiring offsite disposal. The debris is packaged in appropriate waste containers and may be temporarily stored onsite prior to transport for treatment and/or disposal. During transportation of loads that had no visible liquid Hg at the point of origin, temperature changes can cause Hg⁰ to evaporate, condense, and form droplets on container walls. Furthermore, vibration during transportation could cause beads of Hg to be released from the debris, container walls, and ceiling, resulting in pools of liquid Hg⁰ on the container floor. Waste acceptance criteria (WAC) limitations for commercial disposal facilities, the Nevada National Security Site (NNSS), and ORR mixed low-level waste landfills prohibit the presence of any free liquids in containers identified as a solid waste form. Potential solutions to mitigate the presence of residual liquids that could be formed through vapor condensation include the use of sorbents or similar materials to capture and stabilize volatile Hg⁰ vapors and thus ensure compliance with landfill WAC requirements. This report summarizes data from small-scale laboratory experiments conducted to evaluate sorbent materials for Hg⁰ vapor suppression and sorption of liquid Hg⁰ that could form under relevant transportation and disposal conditions. A series of experiments was conducted to evaluate commercial sorbent materials and their effectiveness for Hg⁰ sorption across a temperature range from 19.4°C to 60°C. The impact of residual moisture on sorption was investigated under relevant conditions, and leaching tests were performed to assess the stability of Hg⁰ captured sorbent materials. The results provide estimates for sorbent quantities needed for a given Hg⁰ mass loading based on experimental results exposing sorbents to gaseous and liquid Hg⁰ at various mass ratios. Overall, sorbents that were most effective for Hg⁰ vapor suppression were brominated activated carbons, mackinawite-based sorbents coated on vermiculite, and sulfur-modified granular activated carbon. Elevated temperatures and moisture conditions did not result in significant increases of Hg⁰ headspace concentrations, and the materials also demonstrated high sorption capacities for the sorption of liquid Hg⁰.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Off-Gas Design and Testing from Advanced Chlorination Process

Chlorination of Zr alloy cladding with sulfur-chloride reagents is being considered as a means of separating the main metal from alloying elements, imbedded U, actinides, and fission and activation products to purify the Zr. This would allow the Zr to be disposed of as low-level nuclear waste or possibly be recycled. A multistage process was developed that involves several steps and waste streams. Recycling of sulfur-chloride reagents S 2 Cl 2 and SOCl 2 is planned to minimize the waste stream. The generation of off-gas and entrainment of volatiles is anticipated to be minimized. However, given the complex nature of spent nuclear fuel (SNF), some volatile components will be generated, including the solvents themselves; Cl 2 used for the regeneration of S 2 Cl 2 ; volatile fission product chlorides, such as Sb and 3 H; and chlorinated alloying components, such as Sn, Nb, Fe, and Mo. This report discusses the chemical phenomena that might give rise to volatile species and their planned capture. The volatility of the transition metals will depend on their oxidation state and the compounds that are formed, such as Zr tritide. Loading radionuclides in the process off-gas will depend on the inputs to the process. For instance, processing separated Zr to reduce waste will have a much lower loading of radionuclides than if alternative chlorination were used for decladding SNF. Much of the volatile process streams will be part of the solvent recycling, which may involve online scrubbing. Generally, a caustic scrubber should be included in the off-gas to remove acidic vapors (e.g., Cl 2 , SOCl 2 ). The decontamination procedures for removing these species could involve pretreatment (i.e., heating) options or reduction by contact with H 2 . This report discusses these options along with a framework for testing these alternatives in future experiments. Technologies for off-gas capture will leverage those developed in the Materials Recovery and Waste Forms program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

200-IA-1 Operable Unit Ecological Risk Screen

The purpose of this environmental calculation file (ECF) is to document the assumptions, inputs, methodology, equations, and results of a screening level ecological risk assessment (SLERA) conducted for eight of 12 representative waste sites within the newly formed 200 IA-1 Operable Unit (OU); the remaining four waste sites lacked soil data within the shallow vadose zone and were not further evaluated. The SLERA was conducted to evaluate ecological receptor exposure and potential risk using soil data for radionuclides and nonradionuclides occurring within the shallow vadose zone (0 to 15 ft) soil of the OU. The representative Hanford ecological receptors evaluated are: plants, soil invertebrates, California quail (Callipepla californica), Meadowlark (Sturnella neglecta), Killdeer (Charadrius vociferus), Red-tailed hawk (Buteo jamaicensis), Great Basin pocket mouse (Perognathus parvus), Deer mouse (Peromyscus maniculatus), Grasshopper mouse (Onychomys leucogaster), and badger (Taxidea taxus). During the SLERA, detected soil sample results from the shallow vadose zone were used to calculate Hazard Quotients (HQs) for nonradionuclides and Sum of Fractions (SOF) for radionuclides. The results of this SLERA support the preparation of DOE/RL-2020-51, 200-IA-1 Operable Unit Focused Feasibility Study.

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DOE OSTI2015

A Methodology for defining intermediate level waste package characteristics that ensure criticality safety in a geological disposal facility - 14599

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. The basis for ensuring criticality safety of intermediate level waste (ILW) packages is to specify controls on the fissile material content of individual packages through the application of deterministic criticality safety assessments (CSAs). The approach is defined and controlled through RWM's Disposability Assessment process, which aims to ensure that: criticality is prevented during waste package transport to a GDF; the risk of criticality is tolerable and as low as reasonably practicable (ALARP) during GDF operations; and the likelihood and consequences of criticality arising from the evolution of waste packages after GDF closure are low. The most restrictive phase determines the package fissile material limit and, until now, in many cases the GDF post-closure phase has provided the most restrictive limit. Currently, the GDF post-closure limits are derived deterministically using highly stylised and conservative assumptions about fissile material accumulations occurring on very long timescales into the future. RWM, waste producers and regulators do not currently consider this approach to be yielding an appropriate balance of risk. Therefore, RWM is modifying the process for setting fissile material limits on ILW packages by making use of the outcomes of its recent research projects to establish the likelihood of post-closure criticality in a GDF and the consequences of hypothetical post-closure criticality events (presented in companion papers). This paper presents the revised methodology, which is based on the requirement that criticality in a GDF can be demonstrated to be of low and tolerable consequences with regard the performance of a GDF. A probabilistic 'low-likelihood package envelope' is proposed that establishes the packaging and disposal facility conditions under which post-closure criticality is considered to be of low likelihood and low consequence. The bounding conditions for establishing the extent of the package envelope have been evaluated through application of the methodologies and models developed in RWM's Likelihood of Criticality research project. It is expected that the vast majority of ILW packages would meet the package envelope requirements. The methodology also allows the use of other options, such as deterministic CSAs, for waste packages that fall outside the low-likelihood package envelope. (authors)

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

Hanford WTP LAW Melter Startup and Tuning Feed Material Development - 17065

The U.S. Department of Energy's (DOE) Hanford Site in southeast Washington state is home to 56 million gallons of chemical and radioactive waste stored in underground tanks, the result of more than four decades of plutonium production. To address this challenge, the U.S. Department of Energy contracted Bechtel National, Inc., to design and build the world's largest radioactive waste treatment plant. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) will use vitrification to immobilize most of Hanford's waste. Vitrification involves blending the waste with glass-forming materials, and heating it to high temperature to form molten glass, then pouring it into stainless steel containers. In this glass form, its radioactivity will decay over hundreds to thousands of years. In advance of full operation of WTP, a sequential process of commissioning will be undertaken to ensure the plant operates safely and within design requirements. The first phase of WTP operations will focus on the Low-Activity Waste (LAW) Vitrification Facility and its supporting facilities and systems. Commissioning of the LAW Facility will employ a progressive approach to address the complexity of system operation and introduction of hazards. This approach starts with introduction of low hazard and simple testing configurations, and incrementally increasing the complexity and hazardous makeup of testing feed materials, culminating in fully integrated treatment of radioactive tank waste during hot commissioning. Testing with simulants that are chemically similar to radioactive tank waste introduces significant chemical hazards into treatment systems in the form of nitrogen oxides (NOx) and anhydrous ammonia. A method of testing to allow tuning and demonstration of equipment and to develop operator proficiency prior to introducing these chemical hazards was deemed necessary. To that end, the development and testing of a new 'tuning feed' material is in progress. The tuning feed is important because its use will enable the first-time integration of all LAW Facility process systems. The tuning feed must include key physical and chemical characteristics that correspond to the commissioning simulant and low activity tank waste, without the inherently hazardous materials or toxic off-gas constituents that result from tank waste treatment. Tuning feed is being developed to provide representative melter feed rheology, glass chemistry, cold cap formation, and off-gas generation rate in order to predict system response to actual tank waste to the extent practical. The tuning feed must also be compatible with process equipment and other process streams that will be encountered during the cold commissioning testing program. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

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

Validation of Non-Invasive Waste Assay System (Gamma Box Counter) Performance at AECL Whiteshell Laboratories - 13136

Low-level radioactive waste (LLW) in solid form, resulting from decommissioning and operations activities at AECL's Whiteshell Laboratories (WL), is packaged in B-25 and B-1000 standard waste containers and characterized before it is shipped to an on-site interim storage facility, pending AECL decisions on long term management of its LLW. Assay of the waste packages before shipment contributes to an inventory of the interim storage facility and provides data to support acceptance at a future repository. A key characterization step is a gamma spectrometric measurement carried out under standard conditions using an automated, multi-detector Waste Assay System (WAS), purchased from Antech Corporation. A combination of ORTEC gamma acquisition software and custom software is used in this system to incorporate multiple measurements from two collimated high-resolution detectors. The software corrects the intensities of the gamma spectral lines for geometry and attenuation, and generates a table of calculated activities or limits of detection for a user-defined list of radioisotopes that may potentially be present. Validation of WAS performance was a prerequisite to routine operation. Documentation of the validation process provides assurance of the quality of the results produced, which may be needed one or two decades after they were generated. Aspects of the validation included setting up a quality control routine, measurements of standard point sources in reproducible positions, study of the gamma background, optimization of user-selectable software parameters, investigation of the effect of non-uniform distribution of materials and radionuclides, and comparison of results with measurements made using other gamma detector systems designed to assay bulk materials. The following key components of the validation process have been established. A daily quality control routine has been instituted, to verify stability of the gamma detector operation and the background levels. A test box containing plywood (specific gravity 0.5) with holes for reproducible positioning of known point sources has been constructed and used in several configurations. Gamma spectra have been processed multiple times with user-adjustable parameters set to a variety of values to gain agreement on the most appropriate settings. A similar optimization process was followed with the custom software provided by the instrument manufacturer to reach agreement on the most stable settings for reliable results. Measurements made with the Canberra In-Situ Object Counting System (ISOCS) instrument were compared with the WAS results as part of an estimate of instrument precision and accuracy. The validation report will conclude with recommendations for the configuration, appropriate use and limitations of the Waste Assay System. (authors)

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DOE OSTI2017

Functions and Requirements of the Radioactive Waste Test Platform

The purpose of this document is to provide the function and requirements for the Radioactive Waste Test Platform (hereafter called “test platform”). Historically, testing has been performed at the bench scale with actual waste samples (Fiskum et al. 2006a, b, 2008, 2009; Lumetta et al. 2009; Edwards et al. 2009; Smith et al. 2000, 2001). These tests form the basis for waste treatment operations at the Hanford Site. However, since 2008, there has been limited capability to perform bench-scale tests with actual waste samples. Prior work has shown that simulants can provide useful information about the performance of unit operations; however, there have been examples in the past where simulants did not adequately capture the expected performance of the tank waste materials. Therefore, as the Hanford Site moves towards initiating treatment of tank waste, it is important to have a platform available that can evaluate process options utilizing tank waste samples. This platform can be utilized to evaluate processing options and to address troubleshooting for unexpected system performance. The initial operations for the test platform will be to contribute towards maturing the Direct Feed Low-Activity Waste (DFLAW) feed qualification program and to demonstrate an option for disposition of the melter offgas stream. This equipment will be the ideal platform to evaluate additional process options including filtration operating conditions associated with precipitation, methods to extend resin lifetime associated with multi-cycle degradation, and evaluation of other approaches to minimize the impact of recycle streams from the effluent management system (including Tc) on melter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2003

Partitioning of Inorganic Elements Consumed by Humans Between the Various Fractions of Human Wastes: A Review and Analysis of Existing Literature

The nutritional requirements of humans and astronauts are well defined and show consistency, but the same cannot be said of human wastes. Nutrients taken up by humans can be considered to fall into two major categories - organic and inorganic fractions. Carbon, hydrogen, oxygen, nitrogen and sulfur are elements that are associated with the organic fraction. These elements are taken up in large amounts by humans and when metabolized released in wastes often in gaseous forms or as water. On the other hand, a large number of the elements are simply exchanged and can be accounted for in the liquid and solid wastes of humans. These elements fall into three major categories - cationic macroelements (e.g. Ca, K, Na, Mg and Si), anionic macroelements (e.g P, S and Cl), 17 essential microelements, (e.g. Fe, Mn, Cr, Co, Cu, Zn, Se and Sr). When provided in the recommended concentrations to an adult human, these elements should not normally accumulate in humans, but will be excreted in the different human wastes. Knowledge of the partitioning of these elements between the different human waste fractions is fundamental to understanding (a) how these elements can be recovered for reuse in space habitats, and (b) to developing the processors for waste management. The current literature is exhaustive but sometimes also conflicting. We have used the existing knowledge of nutrition and waste from medical literature and NASA documentation to develop a consensus to typify and chemically characterize the various human wastes. The partitioning of these elements has been developed into a functional model.

Wignarajah, K.↗
DOE OSTI2012

Treatability Variance for Containerised Liquids in Mixed Debris Waste - 12101

The TRU Waste Processing Center (TWPC) is a Department of Energy facility whose mission is to receive and process for appropriate disposal legacy Contact Handled (CH) and Remote Handled (RH) waste, including debris waste stored at various DOE Oak Ridge facilities. Acceptable Knowledge (AK) prepared for the waste characterizes the waste as mixed waste, meaning it is both radioactive and regulated under the Resource Conservation and Recovery Act (RCRA). The AK also indicates that a number of the debris waste packages contain small amounts of containerised liquids. The documentation indicates liquid wastes generated in routine lab operations were typically collected for potential recovery of valuable isotopes. However, during activities associated with decontamination and decommissioning (D and D), some containers with small amounts of liquids were placed into the waste containers with debris waste. Many of these containers now hold from 2.5 milliliters (ml) to 237 ml of liquid; a few contain larger volumes. At least some of these containers were likely empty at the time of generation, but documentation of this condition is lacking. Since WIPP compliant AK is developed on a waste stream basis, rather than an individual container basis, and includes every potential RCRA hazardous constituent within the waste stream, it is insufficient for the purpose of characterizing individual containers of liquid. Debris waste is defined in 40 CFR 268.2(g) as 'solid material exceeding a 60 mm particle size that is intended for disposal and that is: a manufactured object; or plant or animal matter; or natural geologic material'. The definition further states that intact containers of hazardous waste that are not ruptured and that retain at least 75% of their original volume are not debris. The prescribed treatment is removal of intact containers from the debris waste, and treatment of their contents to meet specific Land Disposal Restrictions (LDR) standards. This is true for containers with incidental amounts of liquids, even if the liquid is less than 50% of the total waste volume. Under the proposed variance, all free or containerised liquids (up to 3.8 liters(L)) found in the debris would be treated and returned in solid form to the debris waste stream from which they originated. The waste would then be macro-encapsulated. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Oxidative Stability of Tc(I) Tricarbonyl Species Relevant to the Hanford Tank Waste

Technetium (Tc), which exists predominately in the liquid supernatant and salt cake fractions of the nuclear tank waste stored at the U.S. DOE Hanford Site, is one of the most difficult contaminants to dispose of and/or remediate. In the strongly alkaline environments prevalent in the tank waste, its dominant chemical form is pertechnetate (TcO 4 - , oxidation state +7). However, based on experimentation to-date, a significant fraction of the soluble Tc cannot be effectively separated from the wastes and may be present as a non-pertechnetate species. The presence of a non pertechnetate species significantly complicates disposition of low-activity waste (LAW), and the development of methods to either convert them to pertechnetate or to separate the non-pertechnetate species directly is needed. The challenge is the uncertainty regarding the nature and stability of the alkaline-soluble, low-valence, non pertechnetate species in the liquid tank waste. One objective of the Tc management project is to address this knowledge gap. This fiscal year (FY) 2015 report summarizes experimental work exploring the oxidative stability of model low-valence Tc(I) tricarbonyl species, derived from the [Tc(CO) 3 ] + moiety. These compounds are of interest due to their implied presence in several Hanford tank waste supernatants. Work in part was initiated in FY 2014, and a series of samples containing non-pertechnetate Tc generated ex situ or in situ in pseudo-Hanford tank supernatant simulant solutions was prepared and monitored for oxidation to Tc(VII) (Levitskaia et al. 2014). This experimentation continued in FY 2015, and new series of samples containing Tc(I) as [Tc(CO) 3 ] + •Ligand was tested. The monitoring method used for these studies was a combination of 99 Tc NMR and EPR spectroscopies.

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

Defense waste processing facility (DWPF) liquids model: revisions for processing higher TIO 2 containing glasses

Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. This report documents the development of revised TiO 2 , Na 2 O, Li 2 O and Fe 2 O 3 coefficients in the SWPF liquidus model and revised coefficients (a, b, c, and d).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

BLENDING ANALYSIS FOR RADIOACTIVE SALT WASTE PROCESSING FACILITY

Savannah River National Laboratory (SRNL) evaluated methods to mix and blend the contents of the blend tanks to ensure the contents are properly blended before they are transferred from the blend tank such as Tank 21 and Tank 24 to the Salt Waste Processing Facility (SWPF) feed tank. The tank contents consist of three forms: dissolved salt solution, other waste salt solutions, and sludge containing settled solids. This paper focuses on developing the computational model and estimating the operation time of submersible slurry pump when the tank contents are adequately blended prior to their transfer to the SWPF facility. A three-dimensional computational fluid dynamics approach was taken by using the full scale configuration of SRS Type-IV tank, Tank 21H. Major solid obstructions such as the tank wall boundary, the transfer pump column, and three slurry pump housings including one active and two inactive pumps were included in the mixing performance model. Basic flow pattern results predicted by the computational model were benchmarked against the SRNL test results and literature data. Tank 21 is a waste tank that is used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work scope described here consists of two modeling areas. They are the steady state flow pattern calculations before the addition of acid solution for tank blending operation and the transient mixing analysis during miscible liquid blending operation. The transient blending calculations were performed by using the 95% homogeneity criterion for the entire liquid domain of the tank. The initial conditions for the entire modeling domain were based on the steady-state flow pattern results with zero second phase concentration. The performance model was also benchmarked against the SRNL test results and literature data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Multi-modal chemical characterization of highly viscous submicrometer organic particles

Distinguishing highly viscous organic particles within complex mixtures of atmospheric aerosol and accurate descriptions of their composition, size distributions, and mixing states are challenges at the forefront of aerosol measurement science and technology. Here, we present results obtained from complementary single-particle measurement techniques employed for the in-depth characterization of highly viscous particles. In this study, we demonstrate advantages and synergy of this multi-modal particle characterization approach based on the analysis of individual viscous particles formed in the air-discharged waste produced by a common sewer pipe rehabilitation technology. Using oil immersion flow microscopy, we investigate particle size distributions and morphology of colloidal components present in field-collected aqueous waste condensates. We compare these results with corresponding measurements of viscous particles formed in drying droplets of the aerosolized discharged waste. The colloidal components and viscous particles were found to be approximately 10 µm and 0.5 µm, respectively. The aerosolized viscous particles exhibited a spherical morphology, while the colloidal particles appeared noticeably fractal, resembling fragments of a cured composite material. Chemical imaging of the viscous particles collected on substrates was performed using scanning electron microscopy and soft X-ray spectro-microscopy techniques. Through these methods, comprehensive description of these particles emerged, confirming their high solid-like viscosity, wide-ranging sizes, diverse carbon speciation with high degrees of oxygenation, and high organic volume fractions. The aerosolized viscous particles were further characterized using high-throughput single particle mass spectrometry. This technique provides real-time measurements of composition, size, and morphological metrics for large numbers of individual particles, enabling the identification of their distinct mass spectrometric signatures.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Sludge Processing Options for early HLW Treatment at Hanford

The U.S. Department of Energy’s (DOE) Hanford Site has 177 underground storage tanks that contain wastes from past nuclear fuel reprocessing and waste-management operations. Over 20% of this waste is in the form of an insoluble sludge that will require solids concentration and washing prior to vitrification for long-term disposal. An assessment of potential flowsheet operations to support feed preparation activities prior to high level waste (HLW) vitrification has been conducted to better evaluate pretreatment processing options. Settling studies assessing the baseline approach of a settle-decant method were explored as well as a crossflow filtration system to be used alternatively for concentrating and washing HLW sludge. Significant variations in behavior of settling rates and sludge characteristics give reason to evaluate alternative pretreatment options for the HLW. Non-radioactive sludge containing iron oxide, boehmite, and gibbsite were evaluated via gravity settling and crossflow filtration to determine the behavior of these compounds in various tank waste matrices. Understanding the predictive capabilities of HLW solids settling as well as sludge concentration via crossflow filtration can help provide technical guidance during flowsheet planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Concentration of HLLW from Future SNF Recycling for Efficient Immobilization in a CCIM

Sponsored by the Department of Energy Nuclear Energy’s Fuel Cycle Research and Development Program, the Cold Crucible Induction Melter is being developed as the next generation of melter technology for High Level Liquid Waste’s efficient immobilization in highly durable glass ceramic and ceramic forms. Concentration of the radioactive High Level Liquid Waste generated from the proposed future recycling of spent nuclear fuel, after the fuel’s dissolution in nitric acid, is necessary to take advantage of the inherent attributes of Cold Crucible Induction Melting technology. Based on a provided range of commercial spent nuclear fuel fission product composition data and its expected High Level Liquid Waste raffinate composition data as provided in oxide form, an analysis was completed to concentrate the waste. The analysis involved using nitric acid vapor liquid equilibrium data over a range of boiling temperatures and performing spreadsheet calculations to concentrate the High Level Liquid Waste through evaporation. The calculation results will provide a concentrated nonradioactive surrogate High Level Liquid Waste melter feed recipe for testing in Idaho National Laboratory’s Cold Crucible Induction Melter Pilot Plant. This testing will provide a quantifiable verification of the relatively high feed rates of Cold Crucible Induction Melters compared to those achievable with the current ceramic lined Joule Heated Melters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Numerical modeling of Joule heated ceramic melter

Computational fluid dynamics and heat transfer models of waste glass melters are being developed to support the vitrification campaign at the Waste Treatment and Immobilization Plant (WTP) as part of the United States nuclear waste stewardship and environmental protection programs. The WTP mission aims to achieve the stable and durable processing of vitrified waste in the safest, most efficient, and economical manner possible. Hanford’s tank waste is far more complex and varied than the waste treated by other vitrification facilities and both startup and long-term operational challenges are likely to be even greater than prior experience has shown. There are thousands of different waste compositions, which makes processing the waste a complex and challenging task. Models of test melters, pilot-scale, and full-scale melters have been developed to provide understanding and insight into the various physical and chemical phenomena occurring during vitrification. So the results of these modeling activities will be used to support the WTP operations to process legacy nuclear waste into a stable form for disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

In situ formation of sintered cordierite–mullite nano–micro composites by utilizing of waste silica fume

Highlights: ► We succeeded to obtain in situ formed sintered cordierite–mullite nano–macro composites from waste and pure materials at 1400 °C. ► Their sinterability was greatly dependent on both firing temperature and composition. ► XRD patterns showed that the optimum temperature required for formation of sintered cordierite–mullite nano–macro composites was achieved at 1400 °C. ► The batch containing 70 wt.% cordierite and 30 wt.% mullite exhibited the best properties. ► Microstructures of the densified composites were composed of nano–macro cordierite–mullite structures. -- Abstract: This study aims at in situ formation of sintered cordierite–mullite nano–macro composites having high technological properties using waste silica fume, calcined ball clay, calcined alumina, and magnesia as starting materials. The starting materials were mixed in different ratios to obtain different cordierite–mullite composite batches in which the cordierite contents ranged from 50 to 100 wt.%. The batches were uni-axially pressed at 100 MPa and sintered at 1350, 1400 and 1450 °C to select the optimum temperature required for cordierite–mullite nano–macro composites formation. The formed phases were identified by X-ray diffraction (XRD) pattern. The sintering parameters in terms of bulk density (BD) and apparent porosity (AP) were determined. The microstructure of composites has been investigated by scanning electron microscope (SEM). Cold crushing strength (CCS) of the sintered batches was evaluated. The result revealed that the cordierite–mullite nano–macro composites were in-situ formed at 1400 °C. The batch containing 70 wt.% cordierite showed good physical and mechanical properties.

36 MATERIALS SCIENCE↗