Atomic and micro-structure features of nanoporous aluminosilicate glasses from reactive molecular dynamics simulations
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Materials Science↗
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Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes
Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes PE for high-temperature processing. After removal of the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using an Fe-based catalytic process. The PE-derived graphite anode in a lithium-ion coin cell showed a specific capacity of 345 mAh/g at 0.05C with an initial Coulombic efficiency of 87% and reversible capacity retention of ~100% at different current rates. It also showed a specific capacity of up to 313 mAh/g at 0.5 discharge/charge cycles per hour (0.5C) and Coulombic efficiency of 99.9% after 250 cycles, indicating excellent electrochemical performance as an anode material for lithium-ion batteries. This method illustrates that there are opportunities for upcycling large quantities of PE waste to produce graphite powders suitable for use in LIBs.
25 ENERGY STORAGE↗
Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes
Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes the PE for high-temperature processing. After removing the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using a Fe-based low-temperature (<1500 °C) catalytic process. The PE-derived graphite powder showed excellent electrochemical performance as an anode material for lithium-ion batteries (LIBs) with a capacity of up to 302 mAh/g at 0.5 discharge/charge cycles per hour (0.5 C) and capacity retention of 100% after 415 cycles. This method illustrates there are opportunities for upcycling large quantities of PE waste to produce graphite powders.
Huynh, Ngoc Tien [NETL Site Support Contractor, Na↗
Teaching Radioactive Waste Management in an Undergraduate Engineering Program - 13269
The University of Ontario Institute of Technology is Ontario's newest university and the only one in Canada that offers an accredited Bachelor of Nuclear Engineering (Honours) degree. The nuclear engineering program consists of 48 full-semester courses, including one on radioactive waste management. This is a design course that challenges young engineers to develop a fundamental understanding of how to manage the storage and disposal of various types and forms of radioactive waste, and to recognize the social consequences of their practices and decisions. Students are tasked with developing a major project based on an environmental assessment of a simple conceptual design for a waste disposal facility. They use collaborative learning and self-directed exploration to gain the requisite knowledge of the waste management system. The project constitutes 70% of their mark, but is broken down into several small components that include, an environmental assessment comprehensive study report, a technical review, a facility design, and a public defense of their proposal. Many aspects of the project mirror industry team project situations, including the various levels of participation. The success of the students is correlated with their engagement in the project, the highest final examination scores achieved by students with the strongest effort in the project. (authors)
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Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes
Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes the PE for high-temperature processing. After removing the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using a Fe-based low-temperature (<1500 °C) catalytic process. The PE-derived graphite powder showed excellent electrochemical performance as an anode material for lithium-ion batteries (LIBs) with a capacity of up to 302 mAh/g at 0.5 discharge/charge cycles per hour (0.5 C) and capacity retention of 100% after 415 cycles.
Huynh, Ngoc Tien↗
Variability of K d Values in Cementitious Materials and Sediments
Measured distribution coefficients (K d values) for environmental contaminants provide input data for performance assessments (PA) that evaluate physical and chemical phenomena for release of radionuclides from wasteforms, degradation of engineered components and subsequent transport of radionuclides through environmental media. Research efforts at SRNL to study the effects of formulation and curing variability on the physiochemical properties of the saltstone wasteform produced at the Saltstone Disposal Facility (SDF) are ongoing and provide information for the PA and Saltstone Operations. Furthermore, the range and distribution of plutonium K d values in soils is not known. Knowledge of these parameters is needed to provide guidance for stochastic modeling in the PA. Under the current SRS liquid waste processing system, supernate from F & H Tank Farm tanks is processed to remove actinides and fission products, resulting in a low-curie Decontaminated Salt Solution (DSS). At the Saltstone Production Facility (SPF), DSS is mixed with premix, comprised of blast furnace slag (BFS), Class F fly ash (FA), and portland cement (OPC) to form a grout mixture. The fresh grout is subsequently placed in SDF vaults where it cures through hydration reactions to produce saltstone, a hardened monolithic waste form. Variation in saltstone composition and cure conditions of grout can affect the saltstone's physiochemical properties. Variations in properties may originate from variables in DSS, premix, and water to premix ratio, grout mixing, placing, and curing conditions including time and temperature (Harbour et al. 2007; Harbour et al. 2009). There are no previous studies reported in the literature regarding the range and distribution of K d values in cementitious materials. Presently, the Savannah River Site (SRS) estimate ranges and distributions of K d values based on measurements of K d values made in sandy SRS sediments (Kaplan 2010). The actual cementitious material K d values and solubility values differ from the sandy sediments. The K d value range and distribution currently used in the PA are estimated to range between 0.25*K d and 1.75*K d , where the minimum and maximum values of the ranges reflect the 95% confidence level for the mean K d value (Kaplan 2010). The objective of the research with cementitious materials was to measure the range and distribution of a monovalent (Cs) and I - (anion), divalent (Sr), and trivalent (Eu) ions for a variety of laboratory-prepared saltstone surrogate samples to establish a K d range other than that which is presently used in the PA. It has been observed in laboratory samples that cure temperature profiles can affect properties such as heat of hydration, permeability, porosity, compressive strength, and set time (Harbour et al. 2009). The intent was to identify a range and distribution that could be used by stochastic modelers for the PA. Furthermore, the intent was to replace the arbitrarily selected distributions based on geological sandy sediments and to base it on actual cementitious materials. The scope of this study did not include understanding saltstone sorption mechanisms responsible for increasing or decreasing sorption. Similar to the work with cementitious materials, the purpose of the Pu sediment K d dataset was not to attempt to understand through statistics how to better understand Pu sorption to sediments or to lower Pu K d variance. The sediment Pu K d data is included in this study because it is a key risk driver for the PAs on the SRS, and there is presently no direct studies of Pu variability in SRS soils. Instead the distribution of Pu sediment K d values was assumed to be similar to other cations, as presented by Kaplan (2010).
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MSW Variability Mapping and Conversion to Biofuel
MSW (Municipal Solid Waste) is a form of biomass which consists of categorized components of waste/trash. The general categories are paper, yard trash, construction & debris, appliances, tires, glass, metals, aluminum & steel cans, plastics, organics, inorganics, and HHW (Household Hazardous Waste). This project focuses on the factors within a region or population that contribute to variability in the composition of MSW and in turn MSW’s convertibility to biofuel. A list of contributors was determined (Social Vulnerability Index, Access to Public Transportation, Racial Distribution, GDP, Personal Income) and then JMP was used to perform a Multivariate analysis to determine correlations and a Partial Least-Squares regression to determine Variable Importance Plots for each MSW category. In addition to data analysis, the convertibility of MSW to biofuel was studied via microwave pyrolysis system in order to separate and characterize the various gaseous and bio-oil products.
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Characterization of unsaturated hydraulic conductivity in fractured media using the multistep outflow method - 15461
The multistep outflow method is routinely used to characterize the unsaturated hydraulic conductivity of soils. The technique involves placing a soil sample in a pressure plate apparatus, subjecting the sample to multiple gas pressures in discrete steps through time, and measuring the transient volume of pore fluid extracted. Unsaturated hydraulic property values are estimated through inverse modeling of the experimental conditions. In this study the multistep outflow concept was applied to micro-cracked cementitious materials to assess the efficacy of the technique for these materials. The cementitious materials tested were salt-waste simulant grout samples artificially damaged through oven-drying. Compared to typical soils, fractured media exhibit higher saturated conductivity and lower air-entry pressure, and the volume of fluid extractable from fractures is much lower than soil porosity. To accommodate these material differences the standard test apparatus was modified to incorporate a higher conductivity ceramic pressure plate, a high-precision digital balance for logging outflow mass, a low volume (diameter) effluent line, multiple in-line high-precision gas regulators, and a high-precision low-range pressure gauge. Testing to date indicates that the modified apparatus can provide a viable means to measure the unsaturated hydraulic properties of micro-fractured cementitious materials. However the accuracy/uniqueness of inverse modeling is limited by the inherent characteristics of fractured media: high saturated conductivity, low air-entry pressure, and strong non-linearities. Hydraulic property results in the form of van Genuchten / Mualem curves are presented for three fractured grout specimens. DOE Performance Assessments often involve cementitious barriers and/or waste forms that are predicted or assumed to degrade over time due to various mechanisms such as carbonation-influenced reinforcing steel corrosion, external sulfate attack, differential settlement, and seismic activity. Physical degradation typically takes the form of small-scale cracking / fracturing, and the affected materials reside in unsaturated hydrogeologic zones. In these cases, unsaturated hydraulic properties are needed for fractured cementitious materials to simulate moisture movement and contaminant transport within and around the facility. The outflow extraction method, as implemented in the present study, provides a suitable method for estimating these material properties. (authors)
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Synthesizing Highly Crystalline Graphite Powder for Lithium-Ion Battery Anodes from Bulk Polyethylene Waste
Upcycling plastic waste into graphite can potentially be used to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char, then be converted into a highly crystalline bulk graphite powder using a Fe-based catalytic process. The PE-derived graphite demonstrates excellent electrochemical performance as an anode material for lithium-ion batteries.
Gao, Yuan↗
RADIOLOGICAL IMPLICATIONS OF THE PHYSIOCHEMICAL FORMS OF STRONTIUM RELEASED DURING POSTULATED NUCLEAR WASTE FACILITY ACCIDENT CONDITIONS
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Brown, M.↗
San Francisco Biofuel Program: Brown Grease to Biodiesel Demonstration
Municipal wastewater treatment facilities have typically been limited to the role of accepting wastewater, treating it to required levels, and disposing of its treatment residuals. However, a new view is emerging which includes wastewater treatment facilities as regional resource recovery centers. This view is a direct result of increasingly stringent regulations, concerns over energy use, carbon footprint, and worldwide depletion of fossil fuel resources. Resources in wastewater include chemical and thermal energy, as well as nutrients, and water. A waste stream such as residual grease, which concentrates in the drainage from restaurants (referred to as Trap Waste), is a good example of a resource with an energy content that can be recovered for beneficial reuse. If left in wastewater, grease accumulates inside of the wastewater collection system and can lead to increased corrosion and pipe blockages that can cause wastewater overflows. Also, grease in wastewater that arrives at the treatment facility can impair the operation of preliminary treatment equipment and is only partly removed in the primary treatment process. In addition, residual grease increases the demand in treatment materials such as oxygen in the secondary treatment process. When disposed of in landfills, grease is likely to undergo anaerobic decay prior to landfill capping, resulting in the atmospheric release of methane, a greenhouse gas (GHG). This research project was therefore conceptualized and implemented by the San Francisco Public Utilities Commission (SFPUC) to test the feasibility of energy recovery from Trap Waste in the form of Biodiesel or Methane gas.
09 BIOMASS FUELS↗
Selection and properties of alternative forming fluids for TRISO fuel kernel production
Current Very High Temperature Reactor (VHTR) designs incorporate TRi-structural ISOtropic (TRISO) fuel, which consists of a spherical fissile fuel kernel surrounded by layers of pyrolytic carbon and silicon carbide. An internal sol-gel process forms the fuel kernel using wet chemistry to produce uranium oxyhydroxide gel spheres by dropping a cold precursor solution into a hot column of trichloroethylene (TCE). Over time, gelation byproducts inhibit complete gelation, and the TCE must be purified or discarded. The resulting TCE waste stream contains both radioactive and hazardous materials and is thus considered a mixed hazardous waste. Changing the forming fluid to a non-hazardous alternative could greatly improve the economics of TRISO fuel kernel production. Selection criteria for a replacement forming fluid narrowed a list of ~10,800 chemicals to yield ten potential replacement forming fluids: 1-bromododecane, 1- bromotetradecane, 1-bromoundecane, 1-chlorooctadecane, 1-chlorotetradecane, 1-iododecane, 1-iodododecane, 1-iodohexadecane, 1-iodooctadecane, and squalane. The density, viscosity, and surface tension for each potential replacement forming fluid were measured as a function of temperature between 25 °C and 80 °C. Calculated settling velocities and heat transfer rates give an overall column height approximation. 1-bromotetradecane, 1-chlorooctadecane, and 1-iodododecane show the greatest promise as replacements, and future tests will verify their ability to form satisfactory fuel kernels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Practical Approach to a Closed Nuclear Fuel Cycle and Sustained Nuclear Energy - 12383
Recent systems analysis studies at Oak Ridge National Laboratory (ORNL) have shown that sufficient information is available from previous research and development (R and D), industrial experience, and current studies to make rational decisions on a practical approach to a closed nuclear fuel cycle in the United States. These studies show that a near-term decision is needed to recycle used nuclear fuel (UNF) in the United States, to encourage public recognition that a practical solution to disposal of nuclear energy wastes, primarily UNF, is achievable, and to ensure a focus on essential near-term actions and future R and D. Recognition of the importance of time factors is essential, including the multi-decade time period required to implement industrial-scale fuel recycle at the capacity needed, and the effects of radioactive decay on proliferation resistance, recycling complexity, radioactive emissions, and high-level-waste storage, disposal form development, and eventual emplacement in a geologic repository. Analysis of time factors led to identification of the benefits of processing older fuel and an 'optimum decay storage time'. Further benefits of focused R and D can ensure more complete recycling of UNF components and minimize wastes requiring disposal. Analysis of recycling costs and nonproliferation requirements, which are often cited as reasons for delaying a decision to recycle, shows that (1) the differences in costs of nuclear energy with open or closed fuel cycles are insignificant and (2) nonproliferation requirements can be met by a combination of 'safeguards-by-design' co-location of back-end fuel cycle facilities, and applied engineered safeguards and monitoring. The study shows why different methods of separating and recycling used fuel components do not have a significant effect on nonproliferation requirements and can be selected on other bases, such as process efficiency, maturity, and cost-effectiveness. Finally, the study concludes that continued storage of UNF without a decision to recycle is not a solution to the problem of nuclear waste disposal, but can be a deterrent to public confidence in nuclear energy. In summary, our studies have shown, in contrast to findings of the more prominent studies, that today we do have sufficient knowledge to make informed choices for the values and essential methods of UNF recycling, based on previous research, industrial experience, and current analyses. We have shown the significant importance of time factors, including the benefits of an optimum decay storage time on deploying effective nonproliferation safeguards, enabling reduced recycling complexity and environmental emissions, and optimizing waste management and disposal. Together with the multi-decade time required to implement industrial-scale UNF recycle at the capacity needed to match generation rate, our conclusion is that a near-term decision to recycle as many UNF components as possible is vitally needed. Further indecision and procrastination can lead to a loss of public confidence and favorable perception of nuclear energy. With no near-term decision, the path forward for UNF disposal will remain uncertain, with many diverse technologies being considered and no possible focus on a practical solution to the problem. However, a near-term decision to recycle UNF fuel and to take advantage of processing UNF and surface storing HLW, together with development and incorporation of more-complete recycling of UNF components, can provide the focus needed for a practical solution to the problem of nuclear waste disposal. (authors)
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Accelerated Leach Testing of Glass (ALTGLASS): II. Mineralization of hydrogels by leachate strong bases
The durability of high level nuclear waste glasses must be predicted on geological time scales. Waste glass surfaces form hydrogels when in contact with water for varying test durations. As the glass hydrogels age, some exhibit an undesirable resumption of dissolution at long times while others exhibit near steady-state dissolution, that is, nonresumption of dissolution. Resumption of dissolution is associated with the formation of zeolitic phases while nonresumption of dissolution is associated with the formation of clay minerals. Hydrogels with a stoichiometry close to that of imogolite, (Al 2 O 3 ·Si(OH) 4 ), with ferrihydrite (Fe 2 O 3 ·0.5H 2 O), have been shown to be associated with waste glasses that resume dissolution. Aluminosilicate hydrogels with a stoichiometry of allophane-hisingerite ((Al,Fe) 2 O 3 ·1.3-2Si(OH) 4 ) have been shown to be associated with waste glasses that exhibit near steady-state dissolution at long times. These phases are all amorphous and poorly crystalline and are also found on natural weathered basalt glasses. Interaction of these hydrogels with excess alkali and OH – (strong base) in the leachates, causes the Al 2 O 3 · n SiO 2 (where n=1-2) hydrogels to mineralize to zeolites. Excess alkali in the leachate is generated by alkali in the glass. As a result, preliminary rate-determining leach layer forming exchange reactions are hypothesized based on these findings.
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Preliminary Analysis of Advanced Reactors Storage, Transportation, and Disposal
Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP[1]) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes the efforts by the Spent Fuel & Waste Science and Technology (SFWST) in evaluating characteristics and packaging options for advanced reactor spent nuclear fuel forms. The fuel forms were categorized into three types: (1) tristructural isotropic (TRISO), (2) metallic, and (3) fuel salt. This work emphasized TRISO and metallic SNF and waste streams because of the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations. Preliminary information for the spent-fuel salts discharged from molten-salt reactors (MSRs) is also examined to provide a baseline for future efforts. All calculations and assumptions used publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. The paper also includes a literature review and analysis on the storage, transportation, and disposal evaluations and experiences from reactors with similar fuel forms. Advanced-reactor vendors cite past experiences with Fort St. Vrain for TRISO and the Experimental Breeder Reactor II (EBR-II) for metallics that have major influences on fuel design. Finally, the paper includes preliminary concepts of operation for advanced-reactor SNF. This encompasses storage, transportation, potential treatment, and disposal activities from both a per-canister and systems-integration perspective.
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Spectroscopic Properties of Tc(I) Tricarbonyl Species Relevant to the Hanford Tank Waste
Technetium-99 (Tc) exists predominately in soluble forms in the liquid supernatant and salt cake fractions of the nuclear tank waste stored at the U.S. DOE Hanford Site. In the strongly alkaline environments prevalent in the tank waste, its dominant chemical form is pertechnetate (TcO4-, oxidation state +7). However, attempts to remove Tc from the Hanford tank waste using ion-exchange processes specific to TcO 4 - only met with limited success, particularly processing tank waste samples containing elevated concentrations of organic complexants. This suggests that a significant fraction of the soluble Tc can be present as non-pertechnetate low-valent Tc (oxidation state < +7) (non-pertechnetate). The chemical identities of these non-pertechnetate species are poorly understood. Previous analysis of the SY-101 and SY-103 tank waste samples provided strong evidence that non-pertechnetate can be comprised of [Tc(CO) 3 ] + complexes containing Tc in oxidation state +1 (Lukens et al. 2004). During the last two years, our team has expanded this work and demonstrated that high-ionic-strength solutions typifying tank waste supernatants promote oxidative stability of the [Tc(CO) 3 ] + species (Rapko et al. 2013; Levitskaia et al. 2014). It also was observed that high-ionic-strength alkaline matrices stabilize Tc(VI) and potentially Tc(IV) oxidation states, particularly in presence organic chelators, suggesting that the relevant Tc compounds can serve as important redox intermediates facilitating the reduction of Tc(VII) to Tc(I). Designing strategies for effective Tc processing, including separation and immobilization, necessitates understanding the molecular structure of these non-pertechnetate species and their identification in the actual tank waste samples. To-date, only limited information exists regarding the nature and characterization of the Tc(I), Tc(IV), and Tc(VI) species. One objective of this project is to identify the form of non-pertechnetate in the Hanford waste. To do this, we are developing a spectral library of reference non-pertechnetate compounds that can be compared against actual waste samples. The emphasis of the fiscal year 2015 work was Tc(I) tricarbonyl [Tc(CO) 3 ] + compounds. The key findings are summarized below.
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Effect of Silica Particle Size of Nuclear Waste-to-Glass Conversion - 17319
The process for converting nuclear waste-to-glass in an electric melter occurs in the cold cap, a crust of reacting solids floating on the glass pool. As the melter feed (a mixture of the nuclear waste and glass forming and modifying additives) heats up in the cold cap, glass-forming reactions ensue, causing the feed matrix to connect, trapping reaction gases to create a foam layer. The foam layer reduces the rate of melting by separating the reacting feed from the melt pool. The size of the silica particle additives in the melter feed affects melt viscosity and, hence, foam stability. To investigate this effect, seven nuclear waste simulant feeds of a high-level waste were batched as slurries and prepared with dissimilar ranges of silica particle size. Each slurry feed was charged into a laboratory-scale melter (LSM) to produce a cold cap and the propensity of feeds to foam was determined by pressing dried feeds into pellets and monitoring the change of pellet volume in response to heating. Two of these slurries were designed to have dissimilar glass viscosities at 1150°C. In the low temperature region of the cold cap, before the melter feed connects, the feeds without fine silica particles behaved similar to the high viscosity feed as their volume contracted while the feed with silica particles no larger than 5 µm reacted like the low viscosity feed. However, the feed volume similarities reversed as the feed connected and expanded through the foam region of the cold cap.
Cold Cap↗
Effect of Silica Particle Size on Nuclear Waste-to-Glass Conversion - 17319
The process for converting nuclear waste-to-glass in an electric melter occurs in the cold cap, a crust of reacting solids floating on the glass pool. As the melter feed (a mixture of the nuclear waste and glass forming and modifying additives) heats up in the cold cap, glass-forming reactions ensue, causing the feed matrix to connect, trapping reaction gases to create a foam layer. The foam layer reduces the rate of melting by separating the reacting feed from the melt pool. The size of the silica particle additives in the melter feed affects melt viscosity and, hence, foam stability. To investigate this effect, seven nuclear waste simulant feeds of a high-level waste were batched as slurries and prepared with dissimilar ranges of silica particle size. Each slurry feed was charged into a laboratory-scale melter (LSM) to produce a cold cap and the propensity of feeds to foam was determined by pressing dried feeds into pellets and monitoring the change of pellet volume in response to heating. Two of these slurries were designed to have dissimilar glass viscosities at 1150 deg. C. In the low temperature region of the cold cap, before the melter feed connects, the feeds without fine silica particles behaved similar to the high viscosity feed as their volume contracted while the feed with silica particles no larger than 5 μm reacted like the low viscosity feed. However, the feed volume similarities reversed as the feed connected and expanded through the foam region of the cold cap. (authors)
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