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DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using GeoMelt ICV - WM2020 Conference Paper

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

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DOE OSTI2020

Vitrification of Inactive Incineration Ash in a Full Industrial Scale In Can Melter - 20502

In the framework of the H2020 THERAMIN project funded by the European Commission, CEA performed a pilot-scale trial consisting of the vitrification of inactive ash (Al, Ca, Si, Zn, Bi, K, Mg, etc.) using the In- Can Melter technology. These ashes were coming from incineration of simulated technological wastes such as cotton, latex, neoprene, polyethylene. The borosilicate glass feasibility was first evaluated at a laboratory scale (10-100 g of glass). The volatile behavior of fly ash was controlled by pelletizing to avoid pipes clogging when performing the pilot scale test. A 10 wt.% inorganic binder was added to obtain cohesive pellets. The glass frit composition was selected and its ratio was fixed at 50 wt.%. After this step, a bench test (1 kg of glass) was done to verify the absence of container corrosion. Finally, it was followed by a pilot scale test (50 kg of glass) with 75,000 ash pellets used. The same glass microstructure and chemical composition were observed at each scale. Typical crystals of apatites (Ca, P), zincochromites (Zn, Cr, Al) and bismuth alloys (Bi, Sb) dispersed into a borosilicate glass matrix were observed. The chemical durability of the wasteform was measured during 28 days in a 90 deg. C water. Although there were durable crystalline phases, the hydrolysis rate of the vitreous part is relatively high because of the B and Na contents. However, leach rate profiles suggest that an alteration layer will form, leading to an alteration rate decrease. (authors)

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DOE OSTI2020

Composition Measurements of the LAW Phase 4 Glasses

In this report, the Savannah River National Laboratory provides results from the analyses of glass compositions for a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched versions of the glasses. The resulting data will be used in the development of enhanced property/composition models for waste vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with the targeted compositions. For the quenched glasses from Set 1, measured concentrations of B 2 O 3 and SiO 2 were generally above the targeted values. Measured concentrations for MgO and P 2 O 5 were generally low for glasses. Measured concentrations of ZrO 2 were both below and above the targeted values for glasses. For the quenched glasses from Set 2, measured concentrations of AL 2 O 3 , B 2 O 3 , P 2 O 5 , SiO 2 , and ZrO 2 were generally below the targeted values. Measured concentrations of Na 2 0 were both below and above the targeted values for glasses. The measured concentrations of chlorine, fluorine, and SO 3 were below the targeted values for most of the study glasses, likely because of volatility during melting. Overall, there were no indications of errors in batching of the simulated waste glasses. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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DOE OSTI2026

Evaluation of Pu Solubility in Glass for Sludge Batch 11

Stainless-steel clad Pu from Japan’s Fast Critical Assembly (FCA) reactor is currently being dispositioned at the Savannah River Site. The electrolytic dissolver, operated by Savannah River Nuclear Solutions in H-Canyon, is being utilized to dissolve the material. The resulting solutions are transferred to the Concentration, Storage, and Transfer Facilities, operated by Savannah River Mission Completion (SRMC), for subsequent vitrification at the Defense Waste Processing Facility (DWPF). In support of the FCA mission startup, a preliminary evaluation was conducted by the Savannah River National Laboratory to assess the impact of the FCA discards on the liquid waste system.

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DOE OSTI2023

Product Consistency Test Results for the LAW ML1 Glasses

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DOE OSTI2022

Product Consistency Test Results for the Environmental Management Headquarters (EMHQ) Low-Activity Waste (LAW) Glasses

This report provides the results of the Product Consistency Test leachates from the Environmental Management Headquarters Low-Activity Waste glasses, a series of simulated nuclear waste glasses designed and fabricated at the Pacific Northwest National Laboratory. The series included quenched versions of the glasses. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The measured concentrations of the analytes in the test blank samples were below detection limits. The measured concentrations of B, Li, Na, and Si in the Approved Reference Material-1 samples included with the Product Consistency Tests fell within control chart values indicating proper test performance. The measured glass compositions for the study glasses were close to target values: therefore, little difference was seen when evaluating the normalized values using the targeted or measured glass compositions. The $NC_B$, $NC_{Na}$, and $NC_{Si}$ values were less than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L for all glasses.

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DOE OSTI2023

Product Consistency Test Results for the HLW HAlG Glasses

This report summarizes the chemical analysis of Product Consistency Test leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of quenched simulated nuclear waste glasses designated High-Level Waste High Aluminum Glass that were designed and fabricated at PNNL. The reported data is provided to be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration NCi. NCi of several elements was computed for both the target and measured glass compositions. Several of the glasses exhibited NC B , NC Li , and/or NC Na values that were greater than the Environmental Assessment (EA) benchmark values. Several of the leachates from approved reference material (ARM) glasses included with the study glasses had elemental concentrations greater than the expected ranges. Two water blanks contained measurable amounts of Si; measurements confirmed by rerun samples. One water blank also contained a measurable amount of Na.

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DOE OSTI2021

Product Consistency Test Results for the LAW Phase 5 Glasses

This report provides the results of the Vapor Hydration Test solutions and the Product Consistency Test leachates from the Low-Activity Waste Phase 5 glasses, a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory. The series included quenched and canister-centerline cooled versions of the glasses. These data will be used in the development of enhanced property/composition models for waste glass vitrification at Hanford. Several of the blanks had detectable amounts of sodium and/or silicon. The measured concentrations (mg/L) of boron, sodium, and silicon for the low-activity test reference material glass included with the Product Consistency Tests were much lower than the expected low-activity test reference material test results. A review of the Product Consistency Test data indicated that there was generally little difference between the normalized values based on targeted or measured glass composition. Heat treatment had various impacts on the normalized concentration values. Most of the glasses had NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment and Immobilization Plant low-activity waste constraint of 4 g/L. The release rates for boron, sodium, and silicon were highly correlated for the study glasses.

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DOE OSTI2020

Product consistency test results for the LAW Phase 4 gases

In this report, the Savannah River National Laboratory provides chemical analysis of Product Consistency Test (PCT) leachates from a series of simulated nuclear waste glasses fabricated at the Pacific Northwest National Laboratory (PNNL). The series included quenched and canister-centerline cooled (CCC) versions of the glasses. The resulting data will be used in the development of enhanced property/composition models for waste vitrification at Hanford. For some of the glass leachates, minor scatter among the triplicate values of some analytes were observed. For other leachates, there were more significant differences among the triplicate values. A review of the PCT data noted that there was little difference between the normalized values based on targeted or measured glass composition. Several of the study glasses have normalized concentration of element “i” (NCi) values that are greater than the Hanford Tank Waste Treatment and Immobilization Plant immobilized low-activity waste constraint of 4 g/L for boron (B), sodium (Na), and silicon (Si). The results of these glasses will help ensure the ability of advanced glass performance models to appropriately predict acceptable compositions. For the study glasses with NCi values exceeding 4 g/L, the CCC heat treatment samples had generally lower NCi values than quenched samples. The samples of the Environmental Assessment (EA) reference glass included with each PCT set had generally consistent NCi values. The release rates for boron (B), potassium (K), lithium (Li), sodium (Na), and silicon (Si) were highly correlated for the study glasses.

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DOE OSTI2022

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

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

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DOE OSTI2020

Mercury Speciation of Hanford 241-AP-107 Tank Waste Samples - 20384

The concentration of key mercury species in the Hanford tank waste has not been previously measured. Based on process knowledge, an estimated inventory of 2,100 kg of mercury is assumed and was distributed in varied amounts across 149 single-shell tanks (SSTs) and 28 double-shell tanks (DSTs). Limited analyses of individual SSTs and DSTs for total mercury were all below detection, yielding upper limits in concentration estimates based solely on instrument detection limits, resulting in an uncertain mercury inventory for Hanford tanks. Accurate and reliable chemical speciation and quantification is desired to anticipate, predict, and abate the mass movements of various mercury species through the Hanford waste processing flowsheet. Highly sensitive separation and quantification methods for total, elemental, and monomethyl mercury species, in tank waste, have been developed at the Pacific Northwest National Laboratory's (PNNL) Radiochemical Processing Laboratory (RPL). A non-radiological environmental method to quantify mercury species has been adapted for application to Hanford tank 241-AP-107 raw supernatant feed, filtered feed, and cesium-decontaminated samples. These samples represent feed inventory and pretreatment process effluents prior to immobilization by vitrification. This method separates and preconcentrates total, elemental, and monomethyl mercury onto selective solid substrates with little-to-no radioactive background. The solid samples can then be released from strict radiological control, allowing for analytical work to be conducted in low-level radiological facilities. Separation of mercury species was demonstrated at PNNL's RPL in Richland, Washington before transport and analytical quantification using cold vapor atomic fluorescence spectrometry (CVAFS) at PNNL's Marine Science Laboratory (MSL) in Sequim, Washington. Additional studies were conducted to assess the stability of monomethyl mercury in distilled matrices and preconcentrated onto the Carbotrap{sup R} solid substrate to assess sample degradation over time during transport. (authors)

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DOE OSTI2020

Field-Scale Lysimeter Studies of Glass and Cementitious Waste Forms at the Hanford Site - 20392

The Hanford site Integrated Disposal Facility (IDF) will receive waste forms from vitrification activities at the Hanford Waste Treatment and Immobilization Plant (WTP). The waste form inventory to be disposed of at the IDF will consist of vitrified low-activity waste (LAW) in glass forms and solidified (or encapsulated) secondary wastes in cementitious forms. The IDF is a near-surface burial facility located near Hanford's Waste Treatment and Immobilization Plant in the 200 East Area of the site's Central Plateau. An extensive set of laboratory experimental data has been collected to support IDF performance assessment calculations of the eventual degradation of waste forms and mobility of contaminants. This paper describes the start of a field experiment to generate waste-form performance data on a larger scale (tens of centimeters), over a longer duration (five years or more), and under field conditions representative of the IDF. Results from this study are expected to improve model descriptions of contaminant mobility, reduce uncertainties about the representativeness of laboratory results in the IDF performance assessment, improve stakeholder confidence in the safe disposal of treated waste at the IDF, and facilitate the adoption of informed facility designs with the potential to reduce operational costs. A lysimeter test facility near Hanford's 200 West Area has been repurposed to carry out long-term assessments of waste forms buried in sediments excavated from the IDF. The waste forms are buried in large drainage lysimeters, open caissons that are two meters in diameter and three meters deep. Waste forms are buried in up to six locations in each lysimeter. Samplers collect pore water and drainage water beneath the waste forms and air from the pore space adjacent to waste forms. Sensors measure temperature, soil water content, soil water tension, and drainage flux. One lysimeter containing cementitious waste forms and one containing glass waste forms have been completed, along with a control lysimeter (without waste forms). The cementitious waste forms are Cast Stone grout formulations using a liquid secondary waste simulant and a generic Hanford high-salt simulant, as well as a hydrated lime-based grout formulation using a liquid secondary waste simulant. Technetium-99 and iodine-127 were added as tracers to monitor contaminant mobility. Glass waste forms were fabricated in two formulations (LAWA44 and ORLEC28), with rhenium and molybdenum tracers to monitor waste form degradation. The predicted performance of the waste forms in the lysimeters was modeled prior to installation to determine waste form size, tracer concentration/activity, and locations of sensors and samplers. (authors)

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DOE OSTI2023

Product Consistency Test Results for the LAW HPVR Glasses

This report summarizes the chemical analysis of Product Consistency Test (PCT) leachates received from Pacific Northwest National Laboratory (PNNL). The leachates are from a series of simulated nuclear waste glasses designated Low-Activity Waste High PCT and VHT Response (LAW HPVR) glasses that were designed and fabricated at PNNL. The glasses included both quenched and canister centerline cooled glasses. The reported data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. The elemental release for the study glasses is reported as normalized concentration (NC i ). NC i of several elements was computed for both the target and measured glass compositions, which were similar, resulting in no significant differences when computing the NC i values. The majority of the glasses exhibited NC B , NC Na , and NC Si values that were greater than the Hanford Tank Waste Treatment Plant (WTP) low-activity waste constraint of 4 g/L. Several blank solutions included with the study glasses had K, Na, and Si concentrations above the analytical detection limit. Additionally, several of the reference glasses that were included with the study glasses had Na concentrations that were higher than expected. The unexpected analyte concentrations in the reference glass and blank solutions were insignificant with respect to the reported NC i values, and are noted for completeness and their experimental relevance.

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DOE OSTI2024

Studies on the Vitrification Potential of High Risk Spent Nuclear Fuels

This research explored vitrification of U-Zr nuclear fuel that exhibits an explosion hazard when processed by common nitric acid dissolution. Glass frits were chosen based on properties that had been previously measured as well as being demonstrated successfully in production at the Defense Waste Process Facility (DWPF). Zirconium and Aluminum metal powders were chosen to represent the spent fuels after literature demonstrated that the potential phase that causes the issue for explosive reaction in the effluent stream is tied to the Zirconium-III phase field. DSC measurements were carried out on various simulants and waste forms to better understand thermal behavior of these materials. Zirconium (Zr) was tested with both nitric acid and the new effluent flowsheet that is under development by the SRNL Chemical Process Control (CPC) group. The nitric acid test did show the expected exothermic behavior as well as how the behavior is affected by particle size of the powders used. In contrast, the effluent samples did not show an exothermic peak. However, the concentration of metal in the samples was orders of magnitude lower than in the nitric acid testing and further testing of the effect of concentration on the exothermicity that could be measured confirmed a dampened signal. Tests were run that included frit, both DWPF Frit 510 and Iron Phosphate frit, exhibiting complex curves that included multiple endothermic and exothermic peaks during heating and cooling. The final crucible samples appear similarly to the bulk glasses pointing to these peaks being a part of the dissolution of the simulant into the base glasses.

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DOE OSTI2020

IER-519 CED-2: Final Design for Thermal/Epithermal eXperiments (TEX) with Absorbers to Provide Validation Benchmarks for Hanford Tank Farms

The Hanford tank farms contain 56 million gallons of waste across 177 tanks. The primary criticality safety concern for the waste is the plutonium inventory in waste solids – approximately 670 kg in total. Criticality safety analysis credits the absorption and dilution properties of the large quantities of other elements (aluminum, chromium, iron, manganese, nickel, silicon, sodium, and zirconium) present in the waste. Of these, iron and manganese are by far the most significant neutron absorbers, particularly for the waste compositions of highest criticality safety concern. The criticality safety analyses at the Hanford Waste Treatment and Vitrification Plant (WTP) and the Savannah River tank farms also credit iron and manganese as the primary neutron absorbers to demonstrate subcriticality.

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DOE OSTI2024

The Utility of Waste Tank Historical Reviews in Bulk Waste Removal Operations at the Savannah River Site – 25271

SRMC is actively working to remove, treat, and dispose radioactive waste generated by the separation facilities at SRS since their initial operations in the 1950s. The separation facilities at SRS have produced nuclear materials for a variety of purposes, particularly national defense, and continue to support the disposition of spent fuel through the Accelerated Basin Deinventory program. In almost 70 years of operation, nearly 625,000 m3 (165 million gallons) of radioactive waste have been generated and transferred to the tank farm facilities at SRS [1]. As a result of volume reduction (e.g., evaporation) and waste solidification (e.g., vitrification), approximately 127,000 m3 (33.5 million gallons) of material remain as of June 30, 2024 [2]. This liquid waste has since been stored in 51 large underground waste tanks present on the site. These waste tanks may contain up to 4,921 m3 (1.3 million gallons) of radioactive waste each in the form of saltcake or sludge. SRMC’s contract is to treat and dispose of this waste, clean the tanks, and operationally close them. To date, 8 of the 51 waste tanks have been operationally closed. Waste retrieval and tank closure activities are ongoing in an additional 17 tanks through either operations in the field or in design [3].

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DOE OSTI2023

Heat transfer enhancement due to cold cap motion from bubbling in a waste glass melter

In this study, a computational fluid dynamics (CFD) model was developed to model the motion of a solid cold cap in a waste glass melter. Forced convection bubblers at the base of the melter release air into the molten glass, which forms large bubbles that travel upwards to the cold cap and augment heat transfer from the glass to the cold cap. The CFD model employs the Navier- Stokes equations to solve for the fluctuating flowfield using a rigid body motion dynamic fluid body interaction module. This allows for movement of the floating body in response to the bubbling forces calculated at each time step. Further, the heat flux delivered to the cold cap by the convective bubbling is studied as a function of the bubbling rate. Results for the moving cold cap are compared with the computed heat flux trends for a stationary cold cap. The heat flux delivered to the cold cap from the glass is 25% higher for the case with the moving cold cap. The heat flux was found to be proportional to v0.6 as opposed to v0.9 (where v is the normalized bubbling rate) for the stationary cold cap.

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DOE OSTI2023

Crystalline versus Glassy Nature of Iron Phosphate Waste Forms Subjected to Different Slow Cooling Curves

This report provides experimental details and associated results on the elemental distributions (compositions), crystalline/amorphous phase distributions, and microstructures for iron phosphate reference materials (DPF5-336) made with different cooling curves and starting from different temperatures (i.e., 1050°C and 1200°C). The primary goal of this work was to evaluate the properties of the DPF5- 336 reference material after it was melted and cooled at different rates to simulate the cooling profile of different types of canisters. Crystal fractions across the SCC#2, SCC#3, and SCC#5 samples were all similar ranging in crystal content of 35.66–41.14 mass% with amorphous fractions being the balance (64.34–58.86 mass%, respectively). This shows that the heat treatment profile did not seems to greatly affect either the total crystal content or the phase distributions. The quenched sample was almost completely amorphous (99.55 mass%) with very small peaks identified as Li 3 Fe 2 (PO 4 ) 3 whereas the SCC-treated materials showed peaks for what appear to be ten separate phases present in various concentrations. Another goal of these experiments was to see how the phase distribution affected chemical durability of these waste forms, but that information is being collected at Argonne National Laboratory and will be published in a separate report

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