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DOE OSTI2021

A History of Hanford Tank Waste, Implications for Waste Treatment and Disposal

More than 40 years of plutonium processing have left almost 56 million gallons of mixed radioactive waste sequestered in 177 underground tanks on the Hanford Site. Three different processing technologies were employed for plutonium purification in addition to uranium scavenging and fission product removal from the tank waste. All of these chemical processes have contributed to a complex waste stream that varies from tank to tank that presents downstream processing challenges to render the waste into a safe form for long-term storage. The current disposition pathway for Hanford tank waste is vitrification. To maximize waste loading and minimize the number of high-level waste canisters stored in a geologic repository, pretreatment of the waste is required. Both pretreatment and vitrification operations are impacted by the waste composition.

Separations, filtration, ion exchange↗

DOE OSTI2024

New Tank Mapping Method Improves Waste Removal Process

Savannah River Mission Completion is the Liquid Waste (LW) contractor at the Savannah River Site (SRS). The LW mission is tasked with treating and disposing of legacy nuclear waste. There are multiple facilities involved in this work, including the Concentration, Storage, and Transfer Facilities (CSTF), the Defense Waste Processing Facility (DWPF), the Salt Waste Processing Facility (SWPF), and the Saltstone Production Facility (SPF). The CSTF includes 43 underground waste tanks used to store and support processing of radioactive liquid waste. Waste removal activities, such as salt dissolution campaigns and sludge agitation, are conducted within the CSTF waste tanks to convert the waste into a form that allows for downstream processing at other LW facilities. While performing these waste removal campaigns, camera inspections are performed to assess the quantity and distribution of the remaining waste within the waste tank (i.e. saltcake or sludge). Understanding the quantity and distribution of the salt/sludge within the waste tanks allows for improved waste removal strategies (e.g. mixing pump operation) and refined safety controls. Typically, several camera inspections are performed during a waste removal transfer to verify the elevation of the visible salt/sludge mounds against the known elevation of the liquid surface. The camera inspection footage must then be interpreted by a trained engineer who will develop a 2-D map that depicts the waste distribution at various elevations within the waste tank. This tank mapping is then used in conjunction with conservative assumptions to evaluate the volume of saltcake or sludge that is present within the waste tank.

Mini, Melany↗

DOE OSTI2018

The Retrieval project Asse II - Lessons Learned in Radiological Protection - 18263

From 1967 to 1978 in total 125,787 drums and packages with low- and intermediate level waste were emplaced in the Asse II mine. The disposal operation ended in 1978 due to a revision of the Atomic Energy Act. For the disposal of radioactive waste, a plan approval procedure would have been mandatory due to the Atomic Energy Act, but the mining law was the legal basis for operating the Asse II mine, further on. A concept for a closure of the mine was not existing at that time. On January 1, 2009, the Federal Office for Radiation Protection (Bundesamt fuer Strahlenschutz BfS) became the new license-holder and operator of the Asse II mine under the Atomic Energy Act. The Act on the Organizational Restructuring in the Field of Disposal in Germany became effective on July 30, 2016. The act provides that a federally owned company, the Bundesgesellschaft fuer Endlagerung (BGE mbH, Federal Company for Radioactive Waste Disposal) is formed to carry out the operative tasks of searching for a High Level Waste (HLW) repository, the construction and operation of repositories (Konrad, Morsleben) and of the Asse II mine. Since April 25, 2017 BGE is in charge of the operative tasks that have so far been carried out by BfS. Up to now, the long-term safety of the Asse II mine has not been demonstrated in the case that all the waste will remain inside the mine. Further problems arise from a lack of host rock stability, the incomplete backfilling of the mining areas that have not been used for the emplacement of radioactive waste, and a continuous brine inflow into the mine of approx.12 m{sup 3} per day. For these reasons, according to chap. 57b of the Atomic Energy Act, the Asse II mine is to be closed without delay after retrieving the emplaced nuclear waste. Besides the recent organizational changes in Germany and the special problems giving rise to retrieve the waste, in this paper the operational experiences since 2009 on the way from mining law to atomic law will be described. This applies to the radiation protection regime, the release and clearance procedure, and some measures to boost efficiency. (authors)

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DOE OSTI2019

Oxidation of Organics in the Defense Waste Processing Facility Condensate

Defense Waste Processing Facility (DWPF): The state of South Carolina currently holds 35 million gallons of radioactive waste in 51 underground tanks. DWPF was built in an effort to convert the waste into a more stable form. It turns the sludge waste from the tanks into glass, which will keep the waste stable for 10,000 years. Currently, DWPF uses formic acid in its flowsheet and there is one major problem associated with its use: it decomposes into hydrogen, a flammable gas. Glycolic acid, however, has been found as a possible alternative that performs similarly to formic acid but does not directly form hydrogen. One of the main issues with the glycolic acid flowsheet is the potential for there to be glycolate and other organics in DWPF's Recycle Collection Tank (RCT). These organics could lead to hydrogen generation later in the downstream evaporators and must be destroyed in the RCT in order for glycolic acid to be a suitable replacement for formic acid. Oxidants can be added to the RCT to destroy the organics. Two oxidizers, permanganate and Fenton's reagent (iron catalyzed hydrogen peroxide), are possible candidates. Find optimum processing conditions for the oxidation of organic species in a typical RCT simulant. A reactor served as a small-scale version of the RCT. During each run, an RCT simulant with 8 added organics, including glycolate, was created in the reactor. Either permanganate or peroxide was slowly added for each run through a syringe pump. Each experiment lasts 6 hours and a total of 23 runs are scheduled. Variables of Interest: Oxidizer (Permanganate or Peroxide); Temperature (15 deg. C to 50 deg. C); Oxidant addition time (10 mins to 2 hours); Amount of oxidant (1 x to 1.5 x stoichiometric requirements); Solution pH (3 to 13); Glycolate concentration (125 to 250 mg/kg). Offgas Analyzers: Mass spectrometer and FTIR spectrometer. Samples: pulled at 7 different times throughout the run to determine the amount of organic and inorganic carbon and concentrations of various organics through ion chromatography (IC). Only 3 of the runs have been completed at this point, all with permanganate. Based on the first 3 runs, permanganate seems to oxide the majority of the organics and the run at pH 8 and 32.5 deg. C had the most effective oxidation. The rest of the runs will need to be performed to determine the best oxidant and optimum values of the other variables of interest. Once the experiment is completed, DWPF will select one of the oxidants. Testing will continue to determine its performance during foam-over and other non-ideal conditions. Finally, the oxidant will be tested with actual radioactive waste to confirm its effectiveness.

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DOE OSTI2025

Coal-Waste-Enhanced Filaments for Additive Manufacturing of High-Temperature Plastics and Ceramic Composites

In the United States, coal waste from over a century of mining and burning coal for heat and electricity has accumulated as mountains of coal fly ash and bottom ash and acre-size ponds, coal fines and gob. These materials can be a problem for local communities and water systems. A cost-effective process to utilize high volumes of these coal wastes in a high-value product would be beneficial to those communities by reducing the amount of waste and providing jobs, manufacturing components, and materials from the waste. Many coal-to-products technologies (e.g., carbon fibers, graphene, carbon foam) rely on carefully choosing the starting material and then altering it chemically or thermally to make the products work. Due to the wide variability of composition and coal content in typical coal waste streams, many high-volume coal waste streams are likely to be unsuitable for use in those technologies. Semplastics’ technology has been shown to utilize most types of coal waste successfully without any pre-selection or pre-processing requirements other than a nominal particle-size reduction for wastes like bottom ash. This characteristic of Semplastics’ solution may enable the use of much larger volumes of a wider range of coal wastes than other coal-to-products technologies. In this project, Semplastics leveraged its unique experience with both coal waste (fly ash or coal combustion residuals), resin materials, and 3D printing to develop 3D printer filaments using common coal wastes – bituminous coal fines and fly ash – and researched the feasibility of using other forms of coal waste as fillers. Simple 3D-printed parts were successfully produced from the coal waste enhanced filaments, which were found to have improved strength and stiffness.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2016

Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing High 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. The DWPF SPC system is known as the Product Composition Control System (PCCS). The DWPF will soon be receiving wastes from the Salt Waste Processing Facility (SWPF) containing increased concentrations of TiO 2 , Na 2 O, and Cs 2 O . The SWPF is being built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to process TiO 2 concentrations >2.0 wt% in the DWPF, new viscosity data were developed over the range of 1.90 to 6.09 wt% TiO 2 and evaluated against the 2005 viscosity model. An alternate viscosity model is also derived for potential future use, should the DWPF ever need to process other titanate-containing ion exchange materials. The ultimate limit on the amount of TiO 2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge batch, the waste loading of the sludge batch, and the frit used for vitrification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Single phase melt processed powellite (Ba,Ca)MoO4 for the immobilization of Mo-rich nuclear waste

Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to i) promote the formation of a Cs containing hollandite phase and ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

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DOE OSTI2015

Maintaining department of energy (DOE) waste disposition pathways in today's uncertain budgetary environment - 15129

For many years the Department of Energy (DOE) and its prime contractors have worked collaboratively with commercial waste transportation, treatment, and disposal providers to develop and implement outlets for the efficient management and disposition of radioactive wastes, particularly 'no-path-forward' or 'difficult-to-treat' wastes. Both the DOE and commercial providers have made substantial investments to develop capabilities and infrastructure to address these wastes and both have benefited from this collaboration. For example, DOE realized significant cost savings and avoidances when a commercial facility became available that allowed closure of the TSCA incinerator in Oak Ridge or when commercial capabilities were expanded that precluded the need to build a new Hanford M-91 waste treatment facility, estimated at $540 million for design, permitting, and construction. In addition, expanded capabilities developed at commercial facilities ensured a known, if not guaranteed, source for managing waste and improving efficiencies for wastes which previously had no viable disposition outlet or that required extended schedules and cost for completion. Such wastes include sodium- and mercury-bearing wastes, classified shapes, or many forms of suspect transuranic (TRU) wastes. (authors)

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DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

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DOE OSTI2016

Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO 2 Containing Glasses

Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been 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 relying on 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 determine, 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. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO 2 -, Na 2 O-, and Cs 2 O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO 2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO 2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and the modified property acceptable region limits for the durability constraints be incorporated in the next revision of the technical bases for PCCS and then implemented into PCCS. It is also recommended that an reduction of constraints of 4 wt% Al 2 O 3 be implemented with no restrictions on the amount of alkali in the glass for TiO 2 values ≥2 wt%. The ultimate limit on the amount of TiO 2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge batch, the waste loading of the sludge batch, and the frit used for vitrification.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Form and Aging of Plutonium in Savannah River Site Waste Tank 18

Process modeling; injection molding; long-fiber thermoplastic; fiber orientation; fiber lengt This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH) 4(am) or hydrous plutonium oxide, PuO 2(am,hyd) and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu 4+ that has been substituted for other metal ions in crystal lattice sites, Pu 4+ occluded within hydrous metal oxide particles and Pu 4+ adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO 2(am,hyd) is also likely to be present in deposits and scales that have formed on the steel surfaces of the tank. Over the operational period and after closure of Tank 18, Ostwald ripening has and will continue to transform PuO 2(am,hyd) to a more crystalline form of plutonium dioxide, PuO 2(c) . After bulk waste removal and heel retrieval operations, the free hydroxide concentration decreased and the carbonate concentration in the free liquid and solids increased. Consequently, a portion of the PuO 2(am,hyd) has likely been converted to a hydroxy-carbonate complex such as Pu(OH) 2 (CO 3 ) (s) . or PuO(CO 3 ) · xH 2 O (am) . Like PuO 2(am,hyd) , Ostwald ripening of Pu(OH) 2 (CO 3 ) (s) or PuO(CO 3 ) · xH 2 O (am) would be expected to occur to produce a more crystalline form of the plutonium carbonate complex. Due to the high alkalinity and low carbonate concentration in the grout formulation, it is expected that upon interaction with the grout, the plutonium carbonate complexes will transform back into plutonium hydroxide. Although crystalline plutonium dioxide is the more stable thermodynamic state of Pu(IV), the low temperature and high water content of the waste during the operating and heel removal periods in Tank 18 have limited the transformation of the plutonium into crystalline plutonium dioxide. During the tank closure period of thousands of years, transformation of the plutonium into a more crystalline plutonium dioxide form would be expected. However, the continuing presence of water, reaction with water radiolysis products, and low temperatures will limit the transformation, and will likely maintain an amorphous Pu(OH) 4 or PuO 2(am,hyd) form on the surface of any crystalline plutonium dioxide produced after tank closure. X-ray Absorption Spectroscopic (XAS) measurements of Tank 18 residues are recommended to confirm coordination environments of the plutonium. If the presence of PuO(CO 3 ) (am,hyd) is confirmed by XAS, it is recommended that experiments be conducted to determine if plutonium carbonates are transformed back into PuO 2(am,hyd) upon contact with grout.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Electrochemically Enhanced Carbonate Precipitation into Building Materials: A Scalable Carbon Sequestration Strategy

Decarbonization goals across hard-to-abate industries have prompted an urgent need for advanced carbon capture and storage technologies. Sequestering CO2 into carbonate minerals is a scalable method of carbon management with the ability to produce value-added carbon negative materials from waste streams for the construction industry. Waste streams rich in Ca and Mg such as nickel mine tailings, iron/steel slag, and reverse osmosis brines can store 7.6 Mt CO2/year as minerals. Additionally, CO2 mineralization in acid-neutralization processes currently present in industrial waste treatment can eliminate associated CO2 emissions of lime processing by improving process circularity. The carbonate minerals formed from these waste sources are valuable as components of carbon-negative concrete, which have the potential to sequester 1.8 billion Mt of CO2/year. Electrochemical means of CO2 mineralization improves the kinetics of the thermodynamically favorable mineralization process, lessening or eliminating the high energy requirements of traditional methods. Here, we investigate benchtop scale electrochemical CO2 mineralization of alkaline mining waste, highlighting the effects of key constituents in mining waste on the mineralization process.

carbon capture↗

DOE OSTI2025

Strontium Speciation in Relevant Tank Waste Components Examined by Electrospray Ionization Mass Spectrometry

The identification of chemical species formed in complex nuclear waste is crucial for the development and employment of advanced separations technologies to remediate the Hanford site by processing tank waste. The current Tank Side Cesium Removal (TSCR) process deployed at Hanford utilizes crystalline silicotitanate (CST) ion exchange (IX) media to aid in the separation of low-activity waste for proper treatment and disposal. The inorganic IX media is highly selective for Cs but has been shown to also remove Sr from caustic simulants and small-scale IX processing of Hanford tank waste.(Fiskum, Rovira et al. 2019, Fiskum, Campbell et al. 2021, Westesen, Campbell et al. 2022) Quantitative Sr removal has not been observed in all tank waste supernates tested; thus, to better understand Sr removal and effectively predict processing behavior through TSCR, it is necessary to first investigate Sr speciation in tank waste. This work utilized electrospray ionization mass spectrometry (ESI-MS) to identify ionic Sr complexes that form in the presence of NO 3 –, NO 2 –, OH–, and Cl–. Although our results show that NO 3 –, NO 2 –, and OH– are competitive for Sr 2+ binding, previous data from IX studies indicate that [SrOH] + is not the dominant species of concern in tank waste processing schemes.(Fiskum, Campbell and Trang-Le 2020) Our results show that the [Sr(NO3)]+ species and the [Sr(NO2)] + species form in considerable abundances, which may affect the ability to separate Sr using CST in nuclear waste separation processes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

Highlights: ► In the marketing of recyclable materials, the waste-pickers are the least wins. ► It is proposed creating a network of recycling cooperatives to achieve viability. ► The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS1980

U.S. program assessing nuclear waste disposal in space - A status report

Various concepts for the space disposal of nuclear waste are discussed, with attention given to the destinations now being considered (high earth orbit, lunar orbit, lunar surface, solar orbit, solar system escape, sun). Waste mixes are considered in the context of the 'Purex' (Plutonium and Uranium extraction) process and the potential forms for nuclear waste disposal (ORNL cermet, Boro-silicate glass, Metal matrix, Hot-pressed supercalcine) are described. Preliminary estimates of the energy required and the cost surcharge needed to support the space disposal of nuclear waste are presented (8 metric tons/year, requiring three Shuttle launches). When Purex is employed, the generated electrical energy needed to support the Shuttle launches is shown to be less than 1%, and the projected surcharge to electrical users is shown to be slightly more than two mills/kW-hour.

Rice, E. E.↗

DOE OSTI2016

Effect of B on Crystallization of Li and Na aluminosilicates - 16323

The vitrification of Hanford high-level waste (HLW) requires careful balance of glass-forming additives and nuclear waste components to maximize the waste loading but prevent adverse effects such as crystallization of aluminosilicates upon cooling in canisters. Aluminosilicate crystallization removes chemically durable glass-formers from the melt, leaving behind a residual glass that is enriched in less durable components, such as alkali and boron. In our previous work, it has been shown that the composition of the starting melt has a large influence on the phase and composition of the nucleating crystals, typically nepheline (NaAlSiO{sub 4}) when wastes are simultaneously high in Na and Al. Chemical partitioning allows for additional crystallization in the residual glass. Interestingly, in HLW glass the primary phases to form are those of sodium or lithium aluminosilicates despite the presence of boron, a common additive known to reduce crystallization and lower viscosity and melting point. In this work, we seek to better understand aluminosilicate crystallization in the presence of boron. Twelve glasses were prepared along the tie lines of NaAlSiO{sub 4}-NaBSiO{sub 4} and LiAlSiO{sub 4}-LiBSiO{sub 4}. Glasses were heat-treated following a schedule that traversed the crystallization temperatures of carnegieite (NaAlSiO{sub 4}), nepheline, and eucryptite (LiAlSiO{sub 4}) as determined by thermal analysis. Samples were analyzed using X-ray diffraction (XRD), Raman spectroscopy, and differential scanning calorimetry (DSC). It was observed that Li glasses tended to crystallize at higher boron levels than Na glasses. This behavior may result from the relationship between the structure of the glass network and the structure of the resulting crystal and the relative need for atomic rearrangement prior to crystallization in the Na bearing glass. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Hanford and Savannah River Site Programmatic and Technical Integration

Abstract only. The Hanford Site and the Savannah River Site (SRS) were the primary plutonium production facilities within the U.S. nuclear weapons complex. Radioactive wastes were generated as part of these missions and are stored in similar fashion. The majority of radioactivity maintained by the two sites is located in underground carbon steel tanks in the physical form of supernatant, saltcake, or sludge. Disposition of SRS tank waste is ongoing by converting it into glass (pathway for sludge and radionuclides separated from supernatant or dissolved saltcake) or cement (pathway for the decontaminated supernatant and dissolved saltcake). Tank closure activity has also begun at SRS and will continue for the duration of mission. The Hanford tank waste inventory is roughly 2/3rds larger than SRS's by volume- but nominally half the radioactivity. The baseline disposition path includes high-level and low-activity waste vitrification with separate disposition of contact-handled transuranic tank waste. Retrieval of tank waste from aging single shell tanks (SSTs) into double-shell tanks (DSTs) is currently ongoing. As vitrification commences later this decade, Hanford will be in a similar operations mode as SRS. Site integration is increasing as the missions align. The ongoing integration is centered on key issues that impact both sites- regardless of mission timeframe. Three recent workshop exchanges have been held to improve communication with the primary intent of improving operations and technical work organization. The topics of these workshops are as follows: DST space utilization, optimization, and closure; Waste Feed Qualification; and, Cementitious Waste Forms. Key goals for these and future exchanges include aligning research and technology, preparing for joint initiatives (to maximize budgetary value for the customer), and reviewing lessons learned. Each site has played a leading role in the development of technology and operational practices that can be used to meet current challenges and to minimize the impact of future challenges. This paper provides an overview of the exchanges held, but predominately focuses on the team development and actions leading from the workshops.

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NASA NTRS1992

Glass corrosion in natural environments

Experiments carried out during the progress period are summarized. Experiments carried out involving glass samples exposed to solutions of Tris have shown the appearance of 'spikes' upon monitoring glass dissolution as a function of time. The periodic 'spikes' observed in Tris-based media were interpreted in terms of cracking due to excessive stress in the surface region of the glass. Studies of the interactions of silicate glasses with metal ions in buffered media were extended to systems containing Al. Caps buffer was used to establish the pH. The procedures used are described and the results are given. Preliminary studies were initiated as to the feasibility of adding a slowly dissolving solid compound of the additive to the glass-water system to maintain a supply of dissolved additive. It appears that several magnesium compounds have a suitable combination of solubility and affinity towards silicate glass surfaces to have a pronounced retarding effect on the extraction of uranium from the glass. These preliminary findings raise the possibility that introducing a magnesium source into geologic repositories for nuclear waste glass in the form of a sparingly soluble Mg-based backfill material may cause a substantial reduction in the extent of long-term glass corrosion. The studies described also provide mechanistic understanding of the roles of various metal solutes in the leachant. Such understanding forms the basis for developing long-term predictions of nuclear waste glass durability under repository conditions. From what is known about natural highly reduced glasses such as tektites, it is clear that iron is dissolved as ferrous iron with little or no ferric iron. The reducing conditions were high enough to cause metallic iron to exsolve out of the glass in the form of submicroscopic spherules. As the nuclear waste glass is much less reduced, a study was initiated on other natural glasses in addition to the nuclear waste glass. Extensive measurements were carried out on these glasses in order to characterize their magnetic properties. Results of these studies are described.

Thorpe, Arthur N.↗