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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Vanadium metal-organic framework for selective adsorption

A permanently porous vanadium(II)-containing metal-organic framework (MOF) with vanadium(II) centers and methods for synthesis of such MOF frameworks are provided. Methods for using such compounds to selectively react with N2 over CH4 are provided. In the synthetic methods, a vanadium source, such as VY2(tmeda)2, where Y is a halogen and tmeda is N,N,N′,N′-tetramethylethane-1,2-diamine and a H2(ligand) are reacted in the presence of acid in a solvent at between 110° C. and 130° C. to form an intermediate product. The intermediate product is collected and washed with a washing agent, such as DMF and acetonitrile, and the vanadium(II) based MOF is activated by heating the washed intermediate product to at least 160° C. under dynamic vacuum.

Long, Jeffrey R.↗
DOE OSTI2012

Evaluation of Background Concentrations of Contaminants in an Unusual Desert Arroyo Near a Uranium Mill Tailings Disposal Cell - 12260

The U.S. Department of Energy (DOE) Office of Legacy Management (LM) manages 27 sites that have groundwater containing uranium concentrations above background levels. The distal portions of the plumes merge into background groundwater that can have 50 μg/L or more uranium. Distinguishing background from site-related uranium is often problematic, but it is critical to determining if remediation is warranted, establishing appropriate remediation goals, and evaluating disposal cell performance. In particular, groundwater at disposal cells located on the upper Cretaceous Mancos Shale may have relatively high background concentrations of uranium. Elevated concentrations of nitrate, selenium, and sulfate accompany the uranium. LM used geologic analogs and uranium isotopic signatures to distinguish background groundwater from groundwater contaminated by a former uranium processing site. The same suite of contaminants is present in groundwater near former uranium processing sites and in groundwater seeps emanating from the Mancos Shale over a broad area. The concentrations of these contaminants in Many Devils Wash, located near LM's Shiprock disposal cell, are similar to those in samples collected from many Mancos seeps, including two analog sites that are 8 to 11 km from the disposal cell. Samples collected from Many Devils Wash and the analog sites have high AR values (about 2.0)-in contrast, groundwater samples collected near the tailings disposal cell have AR values near 1.0. These chemical signatures raise questions about the origin of the contamination seeping into Many Devils Wash. (authors)

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

Effects of Rangeland Evapotranspiration on Groundwater Recharge, Discharge, and Flow at the Tuba City, Arizona, Disposal Site - 17159

The US Department of Energy Office of Legacy Management is modeling groundwater flow and contaminant transport at a former uranium mill site near Tuba City, Arizona. A goal is to project groundwater travel times and flow volumes between the former mill site and Moenkopi Wash, a down-gradient stream. Aquifer recharge and discharge are sensitive model parameters; however, assigning representative rates is inexact and involves approximations. We applied a remote sensing algorithm to determine spatially variable evapotranspiration (ET), precipitation recharge, and groundwater discharge rates for use in a large scale groundwater flow model for the site. ET is the combination of evaporation and plant transpiration from soil and groundwater. We estimated landscape-scale ET over a 13-year period (2000 to 2012) for distinct plant communities within a 3531-hectare groundwater model domain (GMD), and then evaluated effects of ET on groundwater recharge and discharge within the GMD. Our empirical algorithm was derived from ground ET measurements, multispectral satellite imagery, and temperature data. Groundwater recharge or discharge rates were calculated for each plant community (or ET Zone) as the difference between precipitation (PPT) and ET rates. Recharge occurred in plant communities where PPT exceeded ET; discharge occurred where ET exceeded PPT. Estimates of groundwater recharge rates for upland plant communities that survive on meteoric water ranged from 0 to 88 mm yr{sup -1}. Estimates of groundwater discharge for phreatophyte plant communities ranged from 23 mm yr{sup -1} to 150 mm yr{sup -1}. Phreatophytes are plants that survive by extracting groundwater. Zero recharge occurred in plant communities where ET rates equaled annual average precipitation. An increase in groundwater recharge (PPT > ET) was associated with past land disturbances and heavy livestock grazing in upland areas. Groundwater discharge (ET > PPT) was highest in riparian phreatophyte communities, but lower than optimal in upland phreatophyte communities impacted by heavy grazing. Protecting phreatophyte communities from grazing at Tuba City could potentially increase ET from 153 mm yr{sup -1} to at least 500 mm yr{sup -1}. The validity of using the ET algorithm to estimate recharge and discharge was evaluated by comparing the net volumetric outflow rate for the model domain with base flow measurements collected from Moenkopi Wash. Modeled outflow approximated measured flow gains in the Wash, increasing our confidence in the ET algorithm and in the model. Results suggest that rangeland management practices that reduce groundwater recharge rates and increase groundwater discharge rates should be evaluated as part of an overall groundwater remediation strategy. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Groundwater Remediation in a Flood plain Aquifer at Shiprock, New Mexico - 16097

The Shiprock, New Mexico, Disposal Site is the location of a former uranium and vanadium ore-processing mill that operated from 1954 to 1968. During milling years, tailings leachate and other mill-related liquids migrated into underlying sediments, resulting in the contamination of groundwater beneath the river terrace containing the mill and a nearby flood plain alluvial aquifer adjacent to the San Juan River. The US DOE Office of Legacy Management manages the site and is implementing the current groundwater remedy for the flood plain area, consisting of natural flushing supplemented by extraction of alluvial groundwater to accelerate reduction in contaminant concentrations. The alluvial aquifer contains elevated concentrations of sulfate, uranium, and nitrate derived largely from the historical eastward migration of groundwater through terrace alluvium and underlying bedrock to the flood plain. The bedrock formation beneath the entire Shiprock site is Mancos shale, which forms an escarpment that separates the terrace from the flood plain. Contamination in the flood plain alluvial aquifer is influenced by a dynamic flow system that changes seasonally and from year to year. Simulations with a flood plain flow model based on monitored groundwater levels and coincident river-surface elevations show that background groundwater flow, unaffected by remediation pumping, is characterized by distinct hydraulic conditions for three separate periods each year. During months of low river flow in winter and early spring, river losses to the aquifer create a large hypo-rheic zone in the south half of the flood plain. Simultaneously, recharge of continually flowing surface water emptying onto the north half of the flood plain from Bob Lee Wash creates divergent, radial flow from the wash outlet and diverts northwestward-migrating contaminant plumes to the northeast where they discharge to the river. Subsequent bank storage processes in May and June from high snow melt runoff change background flow directions, increasing the spread of contamination in the alluvium. From mid-summer through fall, river flows are typically similar in value to those observed prior to the onset of snow melt, but groundwater elevations in the aquifer are noticeably lower than water levels observed in winter and early spring and may induce river losses to the aquifer. Several different contaminant sources have impacted the alluvial aquifer. Much of the contamination originated as tailings leachate and raffinate wastewater that infiltrated the terrace subsurface and eventually reached the flood plain at the escarpment. Additional sources of contamination in the aquifer included discharge of mill effluent to Bob Lee Wash and a pond at the base of the escarpment, and aqueous mobilization of solid-phase contamination in the form of windblown deposits and contaminants adsorbed to alluvial sediments. The degree to which remnant, secondary-source processes, such as leaching of solid-phase deposits and discharge of terrace groundwater across the escarpment, contribute to existing contaminant plumes is uncertain. The flood plain remediation system consists of two groundwater extraction wells, pumping from two horizontal wells placed in trenches near the base of the escarpment, a sump collecting discharges from seeps in the escarpment wall, and discharge of collected water to an evaporation pond on the terrace. Flow-model simulations indicate that most contaminated areas are captured by remediation pumping, resulting in several zones of freshwater between the river and the escarpment. Despite the highly transient nature of groundwater flow beneath the flood plain and the varied contaminant sources, initiation of remediation pumping in 2003 and coordinated operation of the system's various components have successfully removed large amounts of contaminant mass and helped prevent discharge of contaminated groundwater to the river. The percentage reductions of dissolved sulfate, uranium, and nitrate mass in the aquifer estimated for the first 9 years of remediation range from 40% to 72%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Demonstration of Wide Area Radiological Decontamination and Mitigation Technologies for Building Structures and Vehicles - 16508

Researchers at the U.S. Environmental Protection Agency in collaboration with the Department of Homeland Security conducted the 'Wide-Area Urban Radiological Contaminant, Mitigation, and Clean-Up Technology Demonstration' in Columbus, Ohio on June 22-25, 2015. EPA's National Homeland Security Research Center (NHSRC) demonstrated five wide-area radiological decontamination technologies (including strippable coatings, gels, and chemical foam technologies) on an urban building. Decontamination technologies were applied to remove the contaminants from the building's surfaces by physical, chemical, or other methods, which in practice could reduce radiation exposure level. In addition, NHSRC teamed with the Department of Homeland Security (DHS) to demonstrate several radiological contaminant mitigation technologies including building and vehicle wash technologies, as well as several approaches to contain wash water and radioactive particles. 'Radiological contaminant mitigation' technologies are measures taken to reduce adverse impacts of radiological contamination on people and the environment, and facilitate such purposes as restoration of first responder services and critical infrastructure. Radiological contaminant mitigation technologies are designed for containing and removing radiological contamination on the surface in the first hours or days following a radiological event (early phase response). Such technologies include 'radiological particle containment', which is design to prevent the spread of particles which might result from vehicle or foot traffic. Radiological particle containment technologies are applicable for early phase response to contain the radionuclides and to reduce radiation dose to responders and the public. Mitigation also includes 'gross decontamination' technologies, which perform a type of decontamination that is conducted with the goal of reducing contamination levels. This reduction may not meet final cleanup levels, but may be useful to mitigate some public hazard or contain contamination. The purpose of the demonstrations was to educate potential end-users and stakeholders of this technology about a 'Toolbox of Options' for radiological decontamination, as well as radiological contaminant mitigation. Both demonstrations were conducted using a 75-year old brick building and the surrounding area (including parking lots) in Columbus, OH. No radioactive contaminants were applied during either demonstration, as the objective was to duplicate and implement realistic operational conditions for these technologies. Surrogate contaminants such as particle tracers were used in several demonstrations. The decontamination technologies were used in a scaled-up setting with application to the building. Contaminant mitigation technologies were demonstrated on the building as well as on vehicles. Example technology application techniques/accessories included an articulating boom lift, repelling boatswain chair, stand-alone surface material structures, high-volume foam applicators, fire truck foam applicator, a vehicle wash tent for vehicles, particle tracers to simulate radiological contaminants, and high- and low-technology liquid containment approaches. Results and stakeholder observations from the demonstration are now publically in a report entitled 'Technical Report for the Demonstration of Radiological Decontamination and Mitigation Technologies for Building Structures and Vehicles' available at http://www.epa.gov/hsresearch Example information that was obtained included decontamination rate, contaminant mitigation and containment capacity, user friendliness of each technology, the required utilities (electric, water, etc.) for each technology, skill of worker required, and the cost. The condition (color, texture, integrity, etc.) of each building material present on the structure along with all structural components such as gutters, windows, doors, etc. were carefully examined and documented. All demonstrations were open to individuals, organizations, and local, state, federal, tribal, and international governments who may be involved with implementing or planning radiological incident response. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

The Effect of Glass Chemistry on Caesium Volatility in Iron Phosphate Glass Systems - 19344

The UK currently has a stockpile of 99,000 m{sup 3} of intermediate level waste (ILW), with this figure expected to rise to 191,000 m{sup 3} by 2125. Current immobilisation options include cementation and thermal treatment, with cementation being the preferred candidate due to the simplicity of processing. However, thermal processing such as vitrification could offer significant volume reductions, up to 75 % for some waste-streams. In comparison, cementation is generally associated with a volume increase, with a typical waste package seeing a volume increase of 200 - 400 %. For vitrification to be accepted as the preferred treatment option of ILW in the UK, the issue of volatility during the vitrification process must be addressed. Volatile components of ILW such as Cs-137 and Sr-90 account for almost 50 % of the current total radioactivity produced by the UK's ILW stockpile. Therefore, the efficient retention of these elements into the glass immobilisation medium must be considered. Various approaches can be employed when trying to increase volatile retention, however this project has focussed on the influence of glass chemistry on volatility. A series of iron phosphate glasses containing 5 mol % of B{sub 2}O{sub 3}, MnO and ZnO have been produced, where additives have been substituted in place of Fe{sub 2}O{sub 3} in a standard 40Fe{sub 2}O{sub 3}-60P{sub 2}O{sub 5} system. All glass samples were characterised using Raman spectroscopy, SEM-EDX, thermal analysis, XRD and XRF. The glasses were then doped with caesium carbonate and re-melted in the system described below. A simple off-gas system has been developed and commissioned to capture and characterise volatiles released from the glass during melting. The off-gas system has been refined several times, with the most recent design comprising of a sealed stainless steel system, connected to two gas washing bottles containing 150 ml of 50 % nitric acid. Nitrogen gas is passed through the system for the duration of the experiment at a flow rate of ∼ 0.9 L/min to ensure efficient carryover of the volatiles released from the glass melt into the solutions in the gas washing bottles. The total retention of the volatiles can be measuring by analysing the contents of the gas washing bottles using ICP-MS and XRF of the final glass product. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Obtaining Real Production Data Through On-Component Printed SIR Patterns

Cleanliness validation of printed circuit assemblies has become increasingly important as electronic assemblies become smaller, denser, and more challenging to clean, prompting an increase in reliance in custom solutions. With increasingly fast paced development cycles, long lead times and costs of quality custom surrogate test boards become prohibitive in obtaining cleanliness data that is a true representative of the final product. In this study, aerosol jet printing, an additive manufacturing technology for electronics, was used to manufacture surface insulation resistance (SIR) test structures that were evaluated on their ability to detect cleanliness defects and remain stable when clean, as well as their survivability to standard electronics wash processes. Additionally, in an effort to further increase the agility of true product testing, SIR test structures were printed directly onto ball-grid-arrays (BGAs) and tested the survivability of conformally printed leads for data acquisition. SIR test structures displayed good wash survivability on standard FR4 and BGAs, including wraparounds for on-component prints, and high sensitivity to cleanliness defects. The presented results demonstrate the feasibility of printed SIR structures as a quick-turn, highly customizable solution for wash validation with potential for increased fidelity cleanliness testing.

SIR, Additive Manufacturing, Printed Electronics, ↗
DOE OSTI2024

Method for leaching rare earth elements and critical minerals from organically associated materials

A method for extracting rare earth elements and critical minerals including adding an acid to a mixture comprising organically bound rare earth elements. The mixture is maintained at a pH of 0.25 to 4 for a period of time, resulting in a liquor and a leached mixture. The liquor is removed from the leached mixture to form a dewatered cake. The dewatered cake is washed to form a washing liquid. The washing liquid is recycled to create a second slurry comprising organically bound rare earth elements.

Theaker, Nolan↗
DOE OSTI2024

Destructive Examination Protocol for 3013-Container-Package Storing Plutonium-Bearing Materials

The 3013-container-package consists of a convenience, inner and outer container and is used for long-term storage of plutonium-bearing materials. A destructive examination (DE) protocol has been developed to examine the container package visually and with microscopic aid to find any corrosion conditions that could result in the loss of the integrity of the container package over its lifetime. The DE protocol contains three main steps: initial container examination, helium (He) leak testing, and detailed imaging and analysis of the inner container closure weld region (ICCWR). The ICCWR has been determined to be bounding, defined as exhibiting worst case conditions for stress corrosion cracking (SCC) of the inner container. To assess SCC in the ICCWR, the inner container lid is cut into quarters and the weld and He-leak testing gasket are removed. Then a citric acid wash is performed to remove adherent chlorides from the ICCWR. The wash is then sent for analysis to determine the concentration of chlorides in the ICCWR. While the analysis for chloride concentration is being performed, the quarter sections are further sectioned into 1/8th subsections by cutting each section in half. These subsections are washed using nitric acid to remove corrosion products. Then each subsection is imaged using a Wide Angle 3-D Measurement System (WAMS). After analysis of microscope images for potential SCC, additional imaging can be performed, including subsurface imaging. After review, a determination is made of whether the container integrity may potentially affect the safe storage of the material.

container↗
DOE OSTI2026

Determining the Solubility Behavior of Kogarkoite in Simulated Nuclear Waste

Kogarkoite (Na 3 FSO 4 ) is a sparingly soluble fluoride–sulfate double salt that has been identified in high level nuclear waste sludge at the Hanford Site and, more recently, in sludge batch compilation samples at the Savannah River Site (SRS). Due to its complex dissolution behavior, which exhibits an inverse dependence on sodium ion activity, the presence of this mineral poses significant challenges to waste retrieval and processing. Incomplete dissolution during sludge washing can lead to the retention of fluoride and sulfate in the high-level waste feed, potentially causing the formation of corrosive, immiscible molten salt layers, known as "glass gall,” in vitrification melters. Current efforts to optimize flowsheet parameters and wash-water volumes are hindered by the absence of a commercially available, certified reference material, which prevents the accurate calibration of analytical methods and the verification of dissolution kinetics. To address this critical gap, this research focuses on the laboratory synthesis of pure Kogarkoite to serve as a standard for comprehensive solubility and washing performance testing. A coupled synthesis and simulant campaign was executed using an evaporative crystallization protocol designed to replicate the dynamic concentration effects observed in tank farm operations. Thirteen simulant matrices were prepared by dissolving systematically varied ratios of sodium fluoride (NaF) and sodium sulfate (Na 2 SO 4 ) in deionized water under three distinct caustic regimes: 0.0 g (control), 4.0 g (~1 M), and 12.0 g (~3 M) sodium hydroxide (NaOH). While thermodynamic equilibrium models suggest that high-caustic environments should favor the stability of the double salt7, results from this evaporative study at 25 0 C revealed a distinct kinetic divergence. Simulants with high hydroxide loading predominantly yielded large, blocky crystals of sodium sulfate decahydrate (Na 2 SO 4 .10H 2 O). Successful synthesis of pure Kogarkoite was achieved exclusively in specific NaOH-free compositional windows, where the precipitate manifested as fine, opaque granular aggregates. Ion chromatography (IC) analysis confirmed phase purity through the simultaneous stoichiometric depletion of both fluoride and sulfate from the supernatant. This successful synthesis establishes a reproducible route to generate bulk Kogarkoite, enabling the subsequent phase of quantitative dissolution testing using inhibited water to optimize sludge-batch assembly.

Sarker, Md Sharif [Florida International Univ. (FI↗
DOE OSTI2011

Regeneration of field-spent activated carbon catalysts for low-temperature selective catalytic reduction of NOx with NH3

In the process of producing liquid crystal displays (LCD), the emitted NOx is removed over an activated carbon catalyst by using selective catalytic reduction (SCR) with NH3 at low temperature. However, the catalyst rapidly deactivates primarily due to the deposition of boron discharged from the process onto the catalyst. Therefore, this study is aimed at developing an optimal regeneration process to remove boron from field-spent carbon catalysts. The spent carbon catalysts were regenerated by washing with a surfactant followed by drying and calcination. The physicochemical properties before and after the regeneration were investigated by using elemental analysis, TG/DTG (thermogravimetric/differential thermogravimetric) analysis, N2 adsorption-desorption and NH3 TPD (temperature programmed desorption). Spent carbon catalysts demonstrated a drastic decrease in DeNOx activity mainly due to heavy deposition of boron. Boron was accumulated to depths of about 50 {mu}m inside the granule surface of the activated carbons, as evidenced by cross-sectional SEM-EDX analysis. However, catalyst activity and surface area were significantly recovered by removing boron in the regeneration process, and the highest NOx conversions were obtained after washing with a non-ionic surfactant in H2O at 70 C, followed by treatment with N2 at 550 C.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2026

Techno-Economic Analysis of Recycling Strategies for Catalyst and Acid During Catalytic Graphitization

With the aim of meeting the urgent demand for active anode materials (AAM) in energy storage systems, bio-based graphite (biographite) emerges as an affordable solution to de-risk the turbulent supply chain of critical minerals. Anode grade biographite requires high crystallinity and purity, which can be achieved by catalytic graphitization with iron, followed by acid washing. Therefore, a well-conceived process integration that recycles catalyst can be the starting point to commercialization. This study evaluates closed-loop catalyst recovery, and byproducts valorization scenarios through a technoeconomic framework to help understand the scale-up potential of biographite. For the acid washing, three reactors in series meet the required biographite purity at 99.95%. Iron and acid recovery can reduce material consumption and waste generation by ~95%, albeit at the expense of ~80% increase in capital costs. Recovery scenarios present similar capital and operational expenses, yielding minimum selling prices (MSP) near $6 kg-1 of biographite. Monte Carlo methodology reveals that feedstock price accounts for ~60% of MSP variance, followed by plant capacity ~20%. The likelihood of reaching a competitive profit margin of 30% in the U.S. AAM market sits at 85% average for recovery scenarios, and 103% when iron oxide is sold as byproduct. Additionally, an IRR >= 15% can be achieved for half of Monte Carlo simulations, representing promising early-stage results. Biographite production offers a strategic pathway to stabilize the anode market beyond China by integrating established technologies for a scalable, economically viable, and sustainable process. The role of catalyst recovery and byproducts utilization is critical for advancing the biomaterials industry.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

Determination of mercury distribution inside spent compact fluorescent lamps by atomic absorption spectrometry

Highlights: Black-Right-Pointing-Pointer New treatments for CFL are required considering the aim of Directive 202/96/CE. Black-Right-Pointing-Pointer It is shown that most of the mercury introduced into a CFL is in the phosphor powder. Black-Right-Pointing-Pointer Experimental conditions for microwave-assisted sample digestion followed by AAS measurements are described. Black-Right-Pointing-Pointer By washing the glass it is possible to reduce the concentration below legal limits. - Abstract: In this study, spent compact fluorescent lamps were characterized to determine the distribution of mercury. The procedure used in this research allowed mercury to be extracted in the vapor phase, from the phosphor powder, and the glass matrix. Mercury concentration in the three phases was determined by the method known as cold vapor atomic absorption spectrometry. Median values obtained in the study showed that a compact fluorescent lamp contained 24.52 {+-} 0.4 ppb of mercury in the vapor phase, 204.16 {+-} 8.9 ppb of mercury in the phosphor powder, and 18.74 {+-} 0.5 ppb of mercury in the glass matrix. There are differences in mercury concentration between the lamps since the year of manufacture or the hours of operation affect both mercury content and its distribution. The 85.76% of the mercury introduced into a compact fluorescent lamp becomes a component of the phosphor powder, while more than 13.66% is diffused through the glass matrix. By washing and eliminating all phosphor powder attached to the glass surface it is possible to classified the glass as a non-hazardous waste.

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

Urban Mining: Quality and quantity of recyclable and recoverable material mechanically and physically extractable from residual waste

Highlights: • Material recycling and recovery from residual waste by physical and mechanical process has been investigated. • About 6% of recyclable can be extracted by NIR and 2-3Dimension selector. • Another 2% of construction materials can be extracted by adopting modified soil washing process. • Extracted material quality is quite high even some residual heavy metal have been detected by leaching test. - Abstract: The mechanically sorted dry fraction (MSDF) and Fines (<20 mm) arising from the mechanical biological treatment of residual municipal solid waste (RMSW) contains respectively about 11% w/w each of recyclable and recoverable materials. Processing a large sample of MSDF in an existing full-scale mechanical sorting facility equipped with near infrared and 2-3 dimensional selectors led to the extraction of about 6% w/w of recyclables with respect to the RMSW weight. Maximum selection efficiency was achieved for metals, about 98% w/w, whereas it was lower for Waste Electrical and Electronic Equipment (WEEE), about 2% w/w. After a simulated lab scale soil washing treatment it was possible to extract about 2% w/w of inert exploitable substances recoverable as construction materials, with respect to the amount of RMSW. The passing curve showed that inert materials were mainly sand with a particle size ranging from 0.063 to 2 mm. Leaching tests showed quite low heavy metal concentrations with the exception of the particles retained by the 0.5 mm sieve. A minimum pollutant concentration was in the leachate from the 10 and 20 mm particle size fractions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

DESORPTION OF SORBATES FROM MST, MMST, AND CST UNDER VARIOUS CONDITIONS

The Small Column Ion Exchange (SCIX) Program (formerly referred to as the Modular Salt Processing (MSP) Project) seeks to deploy equipment to remove the {sup 134,137}Cs, {sup 90}Sr, and alpha-emitting radionuclides (principally {sup 238,239,240}Pu and {sup 237}Np) from the high level waste salt solutions. The equipment is installed within a high level waste tank to take advantage of the shielding provided by the waste tank. The process will involve adding monosodium titanate (MST) to the waste tank (i.e., Tank 41H) to sorb the Sr and select actinides, removing the MST and entrained sludge with in-riser rotary microfilters, and subsequently using ion-exchange columns containing crystalline silicotitanate (CST) to remove the Cs. After being loaded with Cs, the CST will be ground to reduce the particle size and then transferred into another waste tank (e.g., Tank 40H). The MST and sludge solids stream will be transported to a sludge batch preparation tank (i.e., Tank 42H or Tank 51H) once the SCIX batch is processed. Both streams, MST/solids and CST, will ultimately be transported into and vitrified inside the Defense Waste Processing Facility (DWPF). A series of experiments were performed to examine desorption from monosodium titanate (MST), modified monosodium titanate (mMST), and crystalline silicotitanate (CST) under various conditions. The first two experiments examined desorption from MST and CST under two different sludge treatment processes, aluminum dissolution and sludge washing. Desorption of all sorbates was observed to varying degrees under the aluminum dissolution conditions. The extent of desorption ranged from < 3% to about 50% after 4 weeks, with Pu exhibiting the lowest desorption. At the end of the experiment, the temperature was reduced from 65 C to 25 C and the tests monitored for an additional two weeks. After reducing the temperature, partial resorption of the sorbates was observed with both MST and CST. Under the sludge washing conditions, no desorption of sorbates was observed with MST; however, some additional sorption did occur. For CST, a small amount of Cs leached from the material during the first day of testing, but no further leaching was observed over the remaining test period. The final test was designed to examine the possibility of desorption from both MST and mMST upon increasing the solid to liquid phase ratio. The results of these tests indicated some desorption of Pu from MST within the first two weeks after changing the phase ratio, then resorption of some of the leached Pu over the remaining 4 weeks of the experiment. No desorption of any sorbates was observed for mMST under these conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2010

The hydrothermal reactions of monosodium titanate, crystalline silicotitanate and sludge in the modular salt process: a literature survey

The use of crystalline silicotitanate (CST) is proposed for an at-tank process to treat High Level Waste at the Savannah River Site. The proposed configuration includes deployment of ion exchange columns suspended in the risers of existing tanks to process salt waste without building a new facility. The CST is available in an engineered form, designated as IE-911-CW, from UOP. Prior data indicates CST has a proclivity to agglomerate from deposits of silica rich compounds present in the alkaline waste solutions. This report documents the prior literature and provides guidance for the design and operations that include CST to mitigate that risk. The proposed operation will also add monosodium titanate (MST) to the supernate of the tank prior to the ion exchange operation to remove strontium and select alpha-emitting actinides. The cesium loaded CST is ground and then passed forward to the sludge washing tank as feed to the Defense Waste Processing Facility (DWPF). Similarly, the MST will be transferred to the sludge washing tank. Sludge processing includes the potential to leach aluminum from the solids at elevated temperature (e.g., 65 C) using concentrated (3M) sodium hydroxide solutions. Prior literature indicates that both CST and MST will agglomerate and form higher yield stress slurries with exposure to elevated temperatures. This report assessed that data and provides guidance on minimizing the impact of CST and MST on sludge transfer and aluminum leaching sludge.

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

Plutonium loading capacity of reillex™ hpq anion exchange column - AFS-2 plutonium flowsheet for mox

Radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the dependence of column loading performance on the feed composition in the H-Canyon dissolution process for plutonium oxide (PuO 2 ) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). These loading experiments show that a representative feed solution containing ~5 g Pu/L can be loaded onto Reillex™ HPQ resin from solutions containing 8 M total nitrate and 0.1 M KF provided that the F is complexed with Al to an [Al]/[F] molar ratio range of 1.5-2.0. Lower concentrations of total nitrate and [Al]/[F] molar ratios may still have acceptable performance but were not tested in this study. Loading and washing Pu losses should be relatively low (<1%) for resin loading of up to 60 g Pu/L. Loading above 60 g Pu/L resin is possible, but Pu wash losses will increase such that 10-20% of the additional Pu fed may not be retained by the resin as the resin loading approaches 80 g Pu/L resin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed

The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗