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DOE OSTI2017

Modeling and Analysis of Coupled Flow Through Packed Beds as a Diagnostic Tool for Packed Bed Reactor Health - 17172

The US Department of Energy, Office of River Protection, is building the Waste Treatment and Immobilization Plant (WTP) in Richland, Washington, to immobilize the hundreds of millions of liters of radioactive waste stored at the Hanford Site in Washington state. The offgas generated during vitrification of these wastes requires treatment to remove and/or destroy chemical and radioactive hazardous components before release to the atmosphere. In particular, volatile and semi-volatile hydrocarbon components are to be catalytically oxidized to carbon dioxide and water, whereas nitrogen oxides (NOx) are to be catalytically reduced to nitrogen and water. These tasks are accomplished in a series of packed bed reactors. Achieving the requisite Destruction/Removal Efficiency (DRE) for each of the tracked components is highly dependent on the flow characteristics of the offgas through the packed beds. The flow kinetics may be monitored and characterized during the initial warm-up of the system. The gases, initially at 76 deg. C (169 deg. F), (a) flow through one side of a countercurrent heat exchanger to recover heat from the exiting gases (which have already passed through the catalytic beds), (b) are then further heated electrically to approximately 400 deg. C (750 deg. F; the preferred operating temperature of the Thermo Catalytic Oxidizer (TCO)), (c) pass through the TCO where the reduced components of the offgas are oxidized over a supported Pt catalyst bed, (d) are partially cooled to approximately 315 deg. C (600 deg. F) by addition of approximately 25% (by weight) diluted ammonia/air mixture, (e) pass through the Selective Catalytic Reducer (SCR) where the NOx components are converted to N{sub 2} and H{sub 2}O over a supported V2O{sub 5}/TiO{sub 2} catalyst bed, and (f) exit through the hot side of the countercurrent heat exchanger. The electrical heater is not powerful enough to raise the temperature of the offgas to the desired 400 deg. C without the contribution of the heat recovered in the heat exchanger. Moreover, the thermal mass of the beds and housings of the TCO and SCR require a significant input of thermal energy to reach the desired operating temperatures; the better part of a day is required to bring the unit from ambient temperatures to operating temperatures. Therefore, the temperature of the stream during the initial warm-up of the system at each point in the system is dependent upon the thermal mass, flow dispersion, and permeability of the components, quantified as characteristic/representative volumes and residence times. By monitoring the temperature of the offgas as it enters and exits each portion of the system during this warm-up period one may model the system as a series of small Continuously Stirred Tank Reactors (CSTR)s (referred to as 'CSTRlets') to explain the flow kinetics through the catalyst beds. These component CSTRlets are adjustable in size and number, that one may model the flow profile of the gases through the catalyst beds. For example, a single large CSTR will present an effluent temperature profile with a simple exponential decay shape. Concatenating more CSTRlets generate more complex sigmoidal responses. Concatenating one CSTRlet significantly larger than the others in the train also alters the shape of the effluent temperature profile, in a fashion not replicable by a train of equal-sized CSTRlets. The ultimate goal is to accurately emulate the entire system as a representative CSTRlet train. Once accomplished, this characterization may be used in subsequent warm-ups (following shutdown for catalyst maintenance, for example) or by an intentional step change in inlet offgas temperature to detect/diagnose anomalies in flow through the beds (e.g., channeling or partial blockage.) This work presents development of this model through analytical, calculational, and empirical means. - The analytic model is accomplished by Laplace transform analysis of each component to recover component process transfer functions. - The calculational portion is performed via brute force finite element calculation. - Fitting empirical data from the warm-up of the TCO skid allows one to select the hypothetical CSTRlet trains which best fit the TCO and SCR beds for the analytical and finite element calculations. The first portion of this paper presents the development of the analytical (Laplace space) model of the coupled system. Estimated values for the initial temperature, mass, specific heat, electrical heat, and relative quantities of the offgas, catalyst beds, stainless steel housings, and ammonia dilution stream are used to generate expected temperature profiles at several selected junctures of the system. The number and size of the component hypothetical CSTRlets are calculated from the analytical model and are used to predict temperature profiles at the exit of the system and at intermediate junctions between system components (where thermocouples are present in the actual unit.) The finite element calculations employ a series of CSTRlet volumes to produce a similar suite of temperature profiles, for comparison with the analytical model results. Finally, empirical data from the Factory Acceptance Test system warm-up are used to select and compare the results from each modeling strategy to select the CSTRlet train(s) which best model the behavior of the actual system. This work presents a novel method of system component characterization, of value to the WTP plant and similar vitrification facilities as well as chemical process industries unrelated to waste management. (authors)

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DOE OSTI2017

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↗
DOE OSTI2017

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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DOE OSTI2017

Phyto-screening for Volatile Organic Compounds in a Clay-based Glacial Till at the Niagara Falls Storage Site, New York - 17372

The Niagara Falls Storage Site (NFSS) and associated vicinity properties (VPs) occupy approximately 607 hectares (1,500 acres) of the original 3,035-hectare (7,500-acre) Lake Ontario Ordnance Works (LOOW). The NFSS was used to store radioactive materials, residues and waste from uranium ore processing between 1944 and 1952. The waste-handling practices produced an array of groundwater impacts throughout the site. Although the groundwater at the NFSS is naturally high in total dissolved solids that preclude untreated consumption, the groundwater contaminant levels in some areas of the site are very high and may pose risk to a Construction Worker receptor, as well as limit the options for future site re-use. Volatile organic compound (VOC) impacted soil in the northern portion of the site has produced a groundwater plume with significant concentrations of tetrachloroethene (PCE) and associated daughter compounds. Active groundwater remediation is impractical due to the clayey soils, as evident from adjacent property owners who have unsuccessfully attempted active remediation of their VOC plumes. Consequently, the Buffalo District may evaluate excavation options, with secondary phyto-remediative approaches as passive alternatives to lessen the groundwater impacts during the project's life cycle. A preliminary field-study in the VOC plume area sampled the interior of a Black Locust (Robinia pseudoacacia) tree using a Haglof tree corer. Detectable levels of tetrachloroethene (PCE), trichloroethene (TCE), and cis-1,2,-dichloroethene (cis-DCE) were found in the woody interior of the tree. Subsequent sampling and analyses were performed on proximal trees of varying species in and around the currently delineated VOC plume. The results indicate that poplar and native cottonwood variants preferentially contain the VOCs, whereas collocated hardwood trees produced non-detectable results. Additional sampling of Cottonwoods in the VOC-impacted area promoted further delineations of the impacts. The tree-sampling data can be considered a broad surrogate for soil sampling to optimally delineate the VOC plume without performing additional invasive sampling of the subsurface (e.g., drilling). The results also will be used to determine which native tree species preferentially uptake the VOCs from soil and groundwater. These species, and possibly other hybrid species known for VOC uptake and transpiration, would be propagated to promote interim phyto-extraction of the VOCs during the maintenance period between decision-document completion and likely remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP). (authors)

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DOE OSTI2020

Progress and Technology Assessment of Melt-Cast Ceramic Waste Forms

The Savannah River National Laboratory (SRNL) conducted research from FY2011 – FY2020 to develop ceramic waste forms to treat High-Level Waste (HLW) resulting from reprocessing of commercial Used Nuclear Fuel (UNF) as part of the Department of Energy (DOE) office of Nuclear Energy (NE) Fuel Cycle Technologies (FCT) research program. The objective of the research was to develop a reference ceramic waste form and process technology needed to immobilize the combined HLW raffinate stream, including undissolved solids and soluble technetium, and potentially the TRU waste stream resulting from reprocessing processes. This report compiles the relevant literature produced under, and pertinent to, the program objectives and provides a perspective on the technology development as well as prospects for future.

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DOE OSTI2019

Challenges with vitrification of Hanford High-Level Waste (HLW) to borosilicate glass – An overview

The Hanford site in Washington State houses ~56 million gallons of radioactive wastes stored in 177 underground tanks. The waste must be immobilized and permanently stored, and the plan is to separate the tank wastes into low activity waste (LAW) and high-level waste (HLW) streams. The U.S. Department of Energy is building a Waste Treatment and Immobilization Plant at Hanford site to separately vitrify these two waste streams in borosilicate glass using Joule-heated ceramic melters (JHCM). Although the process of nuclear waste vitrification seems to be well established, in practicality, it is faced with complex problems starting from the design of glass compositions, to processing in melters and long-term performance of the final vitrified waste forms. The article presents an overview of our current understanding of critical challenges related to the development and performance of HLW glasses.

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DOE OSTI2017

Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF composition using optical microscopy. This research establishes a viable alternative to more traditional techniques for evaluating nepheline crystallization in large numbers of glasses, which are prohibitively time consuming or otherwise impractical. The ultimate objective is to combine the nucleation and growth information obtained from DSC, like that presented in this report, with computer simulations of glass cooling within the canister to accurately predict nepheline crystallization in HLW during processing through WTP.

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DOE OSTI2022

Vitrification of Hanford Tank 241-AP-101 Waste and Simulant

Hanford tank 241-AP-101 (referred to herein as AP-101) is the second Hanford radioactive tank waste planned to be processed and vitrified. A simulant version of AP 101 waste was formulated from the best-basis inventory (BBI) for the Hanford Tank 241-AP-101 liquid with an assumed target dilution of the waste from the BBI sodium molarity of 8.61 M to the desired 5.5 M Na. After the addition of glass-forming chemicals (GFCs), the simulant melter feed was processed in a non-radioactive, continuous laboratory-scale melter (CLSM) system. The AP-101 simulant melter feed was charged into the CLSM for 6.11 h of processing, which produced 6.55 kg of glass, for an average glass production rate of 2275 kg m 2 d -1 . Since there were no processing issues with the AP-101 simulant melter feed, AP-101 melter feed made with actual waste was then processed in a CLSM system built into a contamination area in a radioactive environment. The melting behavior characteristics appeared similar for both the simulant and waste melter feeds. The AP-101 waste melter feed was charged into the CLSM for 12.14 h of processing, which produced 8.75 kg of glass, for an average glass production rate of 1530 kg m 2 d -1 . During the AP-101 waste melter feed charging, the pump used to move the feed reached a maximum and it is believed that if the pump had a greater capacity, a greater average glass production rate could have been achieved. A constituent of interest present in low quantities in the AP-101 waste is 99 Tc or its non-radioactive surrogate, Re, added to the AP-101 simulant. Analysis for the quantities of 99 Tc and Re in the AP-101 glass product resulted in an average single-pass retention from the melter feed during relative chemical steady state of 55 ± 2 % for 99 Tc and 45 ± 2 % for Re. Compared to the processing of other melter feeds, the retention of 99 Tc in the AP-101 glass was greater than in both AP-107 and AP-105 glass, while the retention of Re in the AP-101 was less than in the AP-107 glass, but greater than in the AP-105 glass. A spike of I was added into the AP-101 melter feed that could be detected above the analysis detection limits. However, the iodine was only detectable above the ~6 ppm limit in one glass pour: the pour immediately following the burn off of the cold cap, where the I level reached ~30 ppm. This event was significant because the glass was poured immediately after burn off and thus it is presumed that the iodine had yet to volatilize from the glass melt while idling. It is recommended to perform future tests with I spikes at greater levels so that it can be detected in additional glass pours to determine if the expected 50 % retention of I used in the Kim et al. glass models can be confirmed. Offgas liquid samples were analyzed for acetonitrile, which was present at greater concentrations in CLSM liquids than in other scaled melter systems. This result was expected based on unique conditions with the CLSM system including a small plenum space leading to low residence time for offgas and the rapidity of offgas cooling upon exiting the CLSM vessel due to the location and environment. About 90 % of the total acetonitrile captured during both the AP-101 simulant and waste CLSM runs was found in the offgas condensate and demister liquids, thus it is recommended that only those liquids be sent for analysis if future testing to study the presence of acetonitrile in offgas products is desired.

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DOE OSTI2016

Arvia Organics Destruction: An Innovative Treatment Approach for Dioxins, Furans and Other Problematic (Orphan) Organic Wastes - 16262

The safe and efficient disposal of radionuclide-contaminated organic wastes is a key component in the lifecycle of a nuclear power plant. Technologies are available for the processing of some of these wastes although there are a number of waste streams for which the disposition path is not clear or does not exist at all. For example, waste streams such as dioxin, furan and PCB-laden oils and aqueous organic matrices pose an issue for the US Dept. of Energy complex. These wastes present a challenge as they cannot be readily treated using existing treatment technologies because of legislative and material constraints such as gaseous phase radioactivity discharge limits, waste elemental content, by-product generation and waste resistance to thermal decomposition. Alternate methods of organics destruction can circumvent some of these issues by using different mechanisms of organics oxidation. One such innovative method, known as the Arvia organics destruction process, combines adsorption with electrochemical combustion and offers such an alternative. This technology, based on a patented material, Nyex{sup TM} that is able to act as both an adsorbent and as an electrode, allows the electrochemical combustion of adsorbed organics to occur on the material's surface. The result is the destruction of organics whilst the bulk of inorganic and radionuclide components remain in the aqueous phase. These components can subsequently be treated using existing waste processing technologies. This paper outlines recent work conducted on demonstrating the Arvia process against the destruction of dioxin, furan and PCB-laden wastes that demonstrates that the technology successfully removes these target species to below discharge standards in the aqueous phase. Based on these successful trials a test unit has been designed to demonstrate the process on actual radioactive wastes at the DOE Oak Ridge facility. During commissioning at Arvia's facility the unit achieved a destruction rate of over 1.5 ml of 100% organic waste per hour. The unit is currently at a partner laboratory facility based at Oak Ridge National Laboratories and has recently completed a further suite of simulant trials which are reported here. The above work showcases the development, testing and initial application of an innovative technology, first presented to the industry at the Waste Management 2012, as another option for the treatment of challenging wastes with expensive, complex or no current disposal routes. Over the years the technology has gone through a number of proving stages culminating in the testing on real orphan wastes based at the Oak Ridge Facility. The Arvia Organics Destruction Cell technology adds to the portfolio of available treatment technologies and will help facilitate the decommissioning and waste disposal responsibilities of the nuclear industry. (authors)

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DOE OSTI2024

Enhanced prediction of Cs removal by CST from Hanford tank waste with K accountability

Abstract The treatment of Hanford tank waste is one of the most challenging environmental cleanup activities to date. To expedite the processing of liquid waste stored in underground tanks in Washington State it is necessary to remove the significant dose contributor, 137 Cs. Crystalline silicotitanate ion exchanger is currently used to remove 137 Cs from the aqueous phase of Hanford tank wastes in preparation for vitrification at the Waste Treatment and Immobilization Plant (WTP). Improving the understanding of potassium impacts on ion exchange behavior of Cs will help in the operation of a critical component of one of the most complex treatment processes in the world today. Optimization of this process can result in significant cost savings and less waste production. Toward this effort, a series of batch contact tests varied in potassium concentration were conducted to look at the impact of potassium concentration on Cs distribution. Experimental distribution ratios ( K d ) were compared to the distribution ratios predicted using the ZAM model. A significant underprediction of Cs capacity in the presence of potassium was seen with the existing model. A revision of the equilibrium constants was determined and provided a statistically better fit for determining the Cs K d values in tank waste matrices.

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DOE OSTI2019

Treatment of Uranium Catalyst Waste for High Volume Reduction of the Final Waste to be Disposed - 19023

A large volume of problematic radioactive waste of spent uranium catalyst, approximately 7,000 drums was generated by a chemical company in Korea. The spent uranium catalyst has since been stored onsite awaiting a management strategy that is to minimize the volume/cost requirements for final disposal of the uranium-bearing waste. This work developed for the first time a process to greatly reduce the waste volume to be disposed while the secondary wastes generated from the process being released to the environment are controlled to meet the discharge criteria in Korea. (authors)

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DOE OSTI2015

Why accredit ISO/CEI 17025 a nuclear waste management laboratory? - 15112

When there is no storage, nor repository, available on a site, the entity in charge of nuclear waste management should find outlets in order to evacuate outside the site. Each outlet has his rules and waste acceptance criteria (packaging rules, physical/chemical/radiological criteria for example). Before you are able to evacuate your waste, an acceptance's process (or contract review) takes place. This process scans the entire measures implemented by the site in order to satisfy outlet's criteria. When the acceptance is obtained, shipments could be organized, and the process gets into an observation's phase. During this phase, the outlet could choose to inspect a finished nuclear waste package or to audit regularly the waste management process. In this last case, the outlet re-scans the entire documents and implemented measures. This audit lasts from one to several days. A way to tackle/approach the audit is to acknowledge, by a trusted third party (i.e. independent from the outlet), all or part of waste management system, in order to base one's reputation on this organism and win outlet's trust. For this purpose, the entity could choose certification or accreditation. Certification (ISO 9001 as often as not) aims at evaluating a system for managing the quality of one product or service. Accreditation (ISO 17025, for example) provides formal recognition for technical skills within a quality management system (clauses 5 and 4 of the ISO 17025 standard). In 2006, the LDM laboratory is the first French laboratory accredited, in accordance with ISO 17025 standard, for the characterization by gamma spectrometry on 100 L- and 200 L- radioactive waste drums (accreditation No 1-1712 - scope available on www.cofrac.fr). This accreditation allows the lab: - to update the quality management system, - to compile the entire technical file of the gamma-ray assay (description, evaluation of the uncertainties, calibration, proficiency testing programs, measurement procedures,...), - to pass previous identified constraints on to producers and nuclear waste package production, - to maintain permanent quality improvement loop, - to lighten the programs of outlets audits, through the acknowledgement of the technical competence of the laboratory to evaluate by gamma spectrometry the activity contained in our nuclear waste drums. On the other hand, the lab is re-evaluated periodically to ensure its continued compliance with requirements: within a 12 month's interval for an internal audit, with a 18 month's period for COFRAC (French sole official national accreditation body). Since 2006, the lab is accredited. His accreditation renewed in 2010. A new entire renewal audit is planned in 2015. (authors)

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DOE OSTI2020

In-Can Incineration and Vitrification Process for the Treatment of Transuranic Wastes - 20035

Technological alpha contaminated wastes are produced by the nuclear fuel cycle management industry, essentially from MOX fuel production units. These Plutonium Contaminated Materials (PCM) contain a mixture of organic, metal and mineral materials, contaminated with uranium, plutonium and americium. One treatment scenario under study is to incinerate the organic materials, melt the metallic fraction and incorporate the mineral fraction (including ashes from organics incineration) into a glass matrix without sorting or pretreatment (grinding for instance). A scale 1 pilot of this PIVIC process (Process for Incineration and Vitrification In Can) has been commissioned in 2018 at the CEA R and D vitrification facility in Marcoule site. A R and D program aiming to demonstrate the feasibility of this process is currently on going. This paper starts with a description of PIVIC process with explanation of the technologies involved. Then, main results of the first year of operation with the technological demonstration goal are presented. (authors)

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DOE OSTI2021

Selection of Glasses to Confirm the 0.65 Weight Percent Sulfate Solubility Limit for Sludge Batch 10

In preparation for Sludge Batch 10 (SB10) processing, projections of sulfate (SO 4 2- ) in glass at 36% waste loading (WL) were calculated in May 2020 for Tank 40 blend projections representing 0.7M and 0.85M Na wash endpoints. The projected SO 4 2- ) concentrations for either sludge-only (SO) or coupled processing with the Salt Waste Processing Facility (SWPF) were either near or exceeded the current Sludge Batch 9 (SB9) limit of 0.65 weight percent (wt.%). Four nominal glass compositions were selected based on SO and coupled processing for the 0.85M Na wash endpoint Tank 40 blend projection to conduct an initial evaluation of the SB10 sulfate solubility behavior. A sulfate salt phase was absent from each of the prepared glasses, which provided preliminary results that supported the 0.65 wt.% SO 4 2- ) limit for SB10.

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DOE OSTI2024

Development of Magnesium Oxysulfate Formulation for SRPPF Aqueous Recovery Liquid Solidification

The liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery Processes will be solidified into a stable form that is acceptable by Waste Isolation Pilot Plant (WIPP) for disposal. The current Aqueous Recovery flow sheet proposes to solidify the liquid effluent using a grout formula that was developed and tested for the former Waste Solidification Building process. This Portland cement based mixture results in a high pH (~13) leachate from the solidified waste form which is not acceptable to WIPP in the large quantities expected from production at SRPPF. Various cementitious materials were previously evaluated as alternative grout formulations to Portland cement and a MgO-based mix was identified as a promising alternative. A magnesium oxysulfate (MOS) cement formulation comprised of reactive magnesium oxide (MgO), anhydrous magnesium sulfate (MgSO 4 ), and sand, as a non-reactive heat sink provided good mixability, similar density to the original Portland-cement based mix, and a leachate pH of 9.4, within the assumed WIPP brine pH range. However, the MOS formulation exhibited an appreciable amount of heat generation, which resulted in premature setting of a large-scale test.

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DOE OSTI2020

Sulfonamide and Sulfonamido-phenol Ligands for Extraction of f-Elements from Alkaline High-Level Waste

Alkaline High Level Waste (HLW) has been accumulated at Hanford and Savannah River Sites as a result of reprocessing for nuclear weapons production during the cold war. A large volume (∼100 MGal) has been accumulated in carbon steel tanks at Savannah River (SRS) and Hanford. The tank waste contains three separate phases generated when NaOH was added to previously acidic Purex raffinates: 1-2 Supernatant liquid, salt-cake, and sludge. The sludge consists mainly of insoluble hydroxides of transition metals while the supernate and salt-cake contain caustic-soluble materials, including salts of highly radioactive fission products Cs(I) and Sr(II). Current treatment of alkaline HLW in SRS includes: i) The Actinide Removal Process (ARP), which is based on sorption of {sup 90}Sr and Actinides (An) on monosodium titanate (MST), also known as 'alpha-strike' process, followed by ii) Caustic Side Solvent Extraction (CSSX),4 which is used for the extraction of {sup 137}Cs by modified calixarenes in a hydrocarbon diluent. Residual actinides in some tanks are removed after CSSX by an additional ARP process commonly referred to as 'alpha-finishing'. Despite the success of ARP for Sr and An removal, as it is a sorption process, it represents the kinetic bottleneck of integrated salt waste processing. Hence potential introduction of additional organic ligands for actinide extraction (in a modified CSSX process) could simplify the overall integrated process, making it more efficient and economical, with less titanate needed and shorter sorption time, as some of the actinide component would be removed during CSSX. In this study tri-sulfonamide and o-sulfonamido-phenol (mono-sulfonamide) ligands have been studied as extractants for Sm(III), which is being used as an Am(III) surrogate. Our prior studies in the group using a tri-sulfonamide (iPr-tsa-B) showed favorable extraction for Sm(III) nitrate salts from alkaline solutions. Mono-sulfonamides possess similar orientation of binding sites to pyrocatechols, which have been found to be good ligands for Am(III) binding and extraction from alkaline media. Tri-sulfonamide of the type iPr-tsa-B6 (1 mM in CH{sub 2}Cl{sub 2} solution) was studied for Ln{sup 3+} extraction using Sm(NO{sub 3}){sub 3}.6H{sub 2}0 (10 and 25 μM) in alkaline solution of NaOH (0.05, 0.1, 0.2, 0.3 mM) / 0.1 M NaNO{sub 3}. Stripping of the organic phase was done using 0.1 M HNO{sub 3} and quantification of Sm{sup 3+} was done using ICP-OES at 359.3 nm. The need to improve stability of the complex led to synthesis of compounds with N-donor site closer to the central benzene ring to facilitate cation-π interactions. Synthesis of tri-sulfonamide type A: a) Chloromethyl methyl ether, SnCl{sub 4}, CH{sub 2}Cl{sub 2}, 0 deg. C, N{sub 2}, 4 h, 57%; b) NaN{sub 3}, reflux in H{sub 2}O/acetone for 22 h, 80%; c) PPh{sub 3}, THF/H{sub 2}O, 22 h, 79%; d) p-toluene sulfonyl chloride, Et{sub 3}N, 1,2-DCE, 22 h. Extraction: Sm(NO{sub 3}){sub 3}.6H{sub 2}O (2 mM) in 5 ml of aqueous NaOH (pH 10.5 - 14) + 6 ml of CH{sub 2}Cl{sub 2} solution of msa (20 mole equiv.) were rotated on a wheel (60 rpm; 20 h). Stripping: 5 ml of 0.1 M HNO{sub 3} + CH{sub 2}Cl{sub 2} solution of msa (after extraction, centrifugation and filtration) was rotated on the wheel (60 rpm; 20 h). Sm{sup 3+} was quantified using UV-Visible spectrophotometry.

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DOE OSTI2023

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

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DOE OSTI2014

Road Map for Development of Crystal-Tolerant High Level Waste Glasses

The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will also be addressed in this road map. The planned research described in this road map is motivated by the potential for substantial economic benefits (significant reductions in glass volumes) that will be realized if the current constraints (T 1% for WTP and T L for DWPF) are approached in an appropriate and technically defensible manner for defense waste and current melter designs. The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal-tolerant high-level waste (HLW) glasses targeting high waste loadings while still meeting process related limits and melter lifetime expectancies. The modeling effort will be an iterative process, where model form and a broader range of conditions, e.g., glass composition and temperature, will evolve as additional data on crystal accumulation are gathered. Model validation steps will be included to guide the development process and ensure the value of the effort (i.e., increased waste loading and waste throughput). A summary of the stages of the road map for developing the crystal-tolerant glass approach, their estimated durations, and deliverables is provided.

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