Search NASA⌕ Search

SEARCH · Search NASA

Results for “WASTE PROCESSING”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 901 records · Page 50
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Ion Exchange Column Thermal Modeling: Prevention of Process Effluent Boiling - 18577

The Savannah River Site treats and disposes of legacy high-level waste, which resulted from 60 years of nuclear material production, medical research, and other nuclear related work. Complex mixtures of soluble and insoluble solids occupy tanks of which the South Carolina Department of Health and Environmental Control recognizes as the single largest environmental risk state-wide. The risk is reduced through removal, treatment, and solidification of decontaminated salt waste. Presently, SRS decontaminates the soluble salt component of the waste, a solution containing actinides and cesium, via implementation of the Actinide Removal Process and the Modular Caustic Side Solvent- Extraction Unit. Caustic solvent extraction is set to increase in scale by late 2018, with startup of the newly constructed Salt Waste Processing Facility. In conjunction with Salt Waste Processing Facility startup, Savannah River Remediation has been tasked by the Department of Energy to demonstrate modular ion exchange technology. The Tank Closure Cesium Removal project aims to demonstrate the feasibility of modular ion exchange processes for future Department of Energy applications. The Tank Closure Cesium Removal system is scheduled to begin operation in mid-2018. Dissolved salt solution, or influent, will be fed through a filter and ion exchange columns in lead-lag configuration, producing effluent. The effluent is fed to a separate high-level waste tank, for future disposition at Salt-stone Production Facility. Once the demonstration project is complete, the loaded ion exchange columns will be purged, dried, and placed in interim safe storage, awaiting final disposition. The ion exchange media selected for the Tank Closure Cesium Removal system is Honeywell UOP IONSIV{sup C} R-9120. The media is a prime candidate for separation of cesium from radioactive high-level waste. UOP R-9120{sup C} is inorganic and has high capacity, selectivity, and affinity for cesium. During operation, influent is fed to the ion exchange columns, contacting UOP R-9210{sup C} packed within the media vessel. Contact initiates loading of isotopes, and given ample time the solid phase (ion exchange media) and solution achieve thermodynamic equilibrium. UOP R-9120{sup C} is a prime candidate for separation of cesium and dissolved fission products. Although the properties of UOP R-9120{sup C} are favorable for separation, the same properties lead to accumulation of a heat source in the ion exchange bed. Cesium-137 and its daughter product, barium-137 m, generate a clear majority of heat (i.e. ∼ 94%). During operation salt solution flows through each column and transports heat from the ion exchange bed. Once operation is terminated, the effluent contained within the porous bed volume becomes essentially static. In the absence of engineering controls, the heat generated may boil the effluent. Rapid vaporization of aqueous salt solution would lead to bed over pressurization. Over pressurization and an associated release of high-level waste would result in adverse impacts to human health and the environment. This paper focuses on modeling the heat transfer associated with the Tank Closure Cesium Removal system ion exchange column. The model evaluates the maximum heat generation and the associated radionuclide loadings that prevent boiling within of the media vessel, under no flow conditions. A safety significant control, maintaining loadings less than the established loading limit ensures associated impacts to human health and the environment are prevented. (authors)

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

CST Batch Contact Equilibrium Testing of TCCR Tank 10H Batch 1 and 1A Waste Supernate Samples

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with crystalline silicotitanate media to process radioactive waste solutions for the removal of Cs-137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium saltcake solids) within the tank followed by at-tank ion exchange column treatment. Equilibrium batch contact tests conducted in the SRNL Shielded Cells laboratory and associated analyses indicated that both the Tank 10H Batch 1 (pre-production) and 1A (production) waste supernates are unique solutions for which CST cesium removal performance is lower than predicted by ZAM modeling. Cesium distribution coefficients (K d ), percent removal, and CST loading data for the TCCR production supernate batch are provided in Table ES-1. ZAM-based cesium loading predictions are compared to the observed loading for the production batch in Figure ES-1, where significantly lower performance was observed than predicted by the model. Even lower cesium loading results (versus ZAM) were observed previously for the in-tank “teabag” batch contact samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Improving PCM Waste Management Through New Approaches to Screening for LLW Treatment and Disposal - 20145

A major step forward in the management of Plutonium Contaminated Material (PCM) Waste has been taken at the Sellafield site with the introduction of new capabilities to enable wastes previously conservatively designated as PCM Intermediate Level Waste (ILW) to be re-classified for treatment as Low Level Waste (LLW). The Sellafield site, located in the North West of England, within the United Kingdom (UK), has operated plutonium processing plants and waste facilities for almost 70 years. During this time management of the raw PCM waste produced has presented a variety of challenges. The waste is currently processed by high force waste compaction in the Waste Treatment Complex (WTC). The WTC converts 200 litre drums of raw PCM waste into a compacted, grout encapsulated monolith form within stainless steel containment, ready for emplacement within the UK's Geological Disposal Facility (GDF) when it becomes available. To help progress some of the wastes waiting for treatment, recent improvements have been made with the introduction of new waste assay instrumentation which enables the identification of a 'candidate' waste potentially suitable for disposal via the LLW treatment route. Opening a new LLW route for previously designated PCM has been a major advance in PCM waste management policy. It has required a range of technical assessment arrangements to be developed, tested and commissioned and involved the engagement of a wide range of stakeholders. In particular, very detailed technical analysis methods have been developed to ensure that there is a robust and conservative approach to the final sentencing of waste. Following completion of commissioning tests, the first 86 m{sup 3} of previously designated PCM waste was successfully treated through the LLW route and consigned to LLW disposal in June 2019. The direct benefit of sentencing waste to the correct waste route (LLW) is the significant cost savings across the complete range of waste management activities. This work is also strategically important for demonstrating to a range of stakeholders including workforce and regulators, that waste previously designated as PCM can be re-classified in a systematic and justifiable way without increasing risk to personnel or the wider environment. By ensuring that waste is being managed appropriately it will be possible to carry out more remediation work safely, in less time, and at a lower cost. (authors)

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

Volatile species of technetium and rhenium during waste vitrification

Volatile loss of technetium (Tc) during vitrification of low-activity wastes is a technical challenge for treating and immobilizing the large volumes of radioactive and hazardous wastes stored at the U.S. Department of Energy's Hanford Site. There are various research efforts being pursued to develop technologies that can be implemented for cost effective management of Tc, including studies to understand the behavior of Tc during vitrification, with the goal of eventually increasing Tc retention in glass. Furthermore, one of these studies has focused on identifying the form or species of Tc and Re (surrogate for Tc) that evolve during the waste-to-glass conversion process. This information is important for understanding the mechanism of Tc volatilization. In this paper, available information collected from the literature is critically evaluated to clarify the volatile species of Tc and Re and, more specifically, whether they volatilize as alkali pertechnetate and perrhenate or as technetium and rhenium oxides after decomposition of alkali pertechnetate and perrhenate. The evaluated data ranged from mass spectrometric identification of species volatilized from pure and binary alkali pertechnetate and perrhenate salts to structural and chemical analyses of volatilized materials during crucible melting and scaled melter processing of simulated wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Implications of Non-Bingham Rheology

One of the key challenges now facing the US Department of Energy (DOE) is the fate of radioactive waste remaining from World War II and the Cold War which is stored underground in tanks some 75 feet in diameter and over 30 feet tall. Over time, the waste has segmented into multiple layers with sludges and slurries at the bottom with salt crust layers often at the top and liquid in between. DOE’s current official baseline plans call for remaining sludges and slurries to be removed from the tanks and converted into a more stable glass waste form. Minimizing worker exposure to radiation drives DOE to use slurry processing techniques to suspend, mobilize, transport, mix, and process the waste. Therefore, a clear and quantitative understanding of Hanford waste rheology is essential for the success of the DOE mission. Historically much of the waste has been characterized using Eugene Bingham’s model more than a century old that provides a straight line fit to higher shear rate data with the intercept suggesting a yield stress and the slope providing the consistency. Yet, Bingham fits overestimate the shear stress at a given shear rate for low to intermediate shear rates, exactly the range of shear rates typically encountered in pipe flow, where shear rates peak along the pipe wall and vanish in the center. This model produces a fictitious yield stress for some of the wastes that do not exhibit yield phenomena. While over estimating the yield stress may be prudent, safe, and conservative for some applications (e.g., pump sizing to ensure that pumps can handle yield stresses), over estimating the rheology may be inaccurate and non-conservative for other applications (e.g., eroding settled particle beds). Therefore, this paper evaluates the slurry rheology of Hanford and Savannah River wastes using a more modern rheological model that fits the full range of experimental data. Although a bias has been recognized and alternative models proposed, the magnitude of this bias and the implications for tank waste have only been qualitatively suggested. The purpose of this paper is to evaluate quantitatively implications of the poor quality of fit between a Bingham model for rheology and experimental data at modest shear rates. We first demonstrate the magnitude of the bias between the data and the Bingham extrapolation. We then evaluate quantitatively the velocity profile under laminar conditions. This analysis shows that the bias may be large (hundreds of percent or more) at modest shear rates and that modest shear rates dominate pipe velocity profiles.

Rheology, Radioactive waste↗
DOE OSTI2017

Phase 1 Testing Results of Immobilization of WTP Effluent Management Facility Evaporator Bottoms Core Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate, along with entrained, volatile, and semi-volatile metals, such as Hg, As, and Se. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate that get recycled to the melter, and is a key objective of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of earlier tasks was to formulate and prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations and use it in evaporator testing to predict the composition of the effluents from the Effluent Management Facility (EMF) evaporator to aid in planning for their disposition. The objective of this task was to test immobilization options for this evaporator bottoms aqueous stream. This document describes the method used to formulate a simulant of this EMF evaporator bottoms stream, immobilize it, and determine if the immobilized waste forms meet disposal criteria.

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

Stabilization and Solidification of Nitric Acid Effluent Waste at Y-12

Consolidated Nuclear Security, LLC (CNS) at the Y-12 plant is investigating approaches for the treatment (stabilization and solidification) of a nitric acid waste effluent that contains uranium. Because the pH of the waste stream is 1-2, it is a difficult waste stream to treat and stabilize by a standard cement-based process. Alternative waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the nitric acid effluent wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Composition Measurements of the HLW APPS Glasses

This report provides the results from the chemical analyses of the glass compositions of the High Activity Waste Aspen Process Performance Simulation (HLW APPS) study glasses, a series of simulated nuclear waste glasses designed and fabricated at Pacific Northwest National Laboratory. These data will be used in the development, validation, and implementation of enhanced property/composition models for waste glass vitrification at Hanford. Chemical analyses were performed on a representative sample of each of the quenched glasses to allow for comparisons with targeted compositions. The relative differences between the targeted and measured concentrations of F- for three glasses and B 2 O 3 for five glasses were greater than 10%. These results can be used in further characterization of this series of glasses, including the normalization of Product Consistency Test results.

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

Solid Secondary Waste Form Testing Activities for the Integrated Disposal Facility Performance Assessment Support - 17567

Solid secondary waste (SSW) that will be generated at the Waste Treatment and Immobilization Plant (WTP) and from other activities at the Hanford site is destined for disposal at the Integrated Disposal Facility (IDF). Construction of the IDF was undertaken based on a Performance Assessment (PA) performed in 2001. In 2016, a SSW data package was prepared to support the development of an updated PA (2017 IDF PA). There are many inputs required to support the IDF PA. Data needs were identified based on experience from the previous PAs and from the development of conceptual models and modeling approaches to be applied for the 2017 PA. Interactions with the IDF PA team and the inventory data package developer helped to focus detailed efforts on a few key waste streams, contaminants of concern, and disposal configurations that are expected to be the primary contributors to the dose resulting from SSW disposed at the IDF. The key waste streams include: High Efficiency Particulate Air (HEPA) filters, ion exchange resins, granular activated carbon, and Ag-mordenite. The key parameters addressed in the 2016 data package include: saturated hydraulic conductivity, moisture characteristic curves, effective diffusion coefficient, distribution coefficient and porosity. The IDF PA team also identified a set of key contaminants that are expected to be the primary concern for the PA calculations (Tc-99, I-129, Cs-137, Sr-90, uranium isotopes (and total uranium), chromium, mercury and nitrate). These contaminants were the focus of the SSW data package. Specific formulations had not been identified for cementitious materials that will be used to encapsulate or stabilize SSW, and no IDF-specific experiments had been conducted to obtain the data for the PA. Thus, for the 2016 SSW data package, recommended property values were provided for a range of representative candidate materials based on a review of existing literature. However, starting in FY17, IDF-specific experiments will be performed for Hanford SSW expected to contribute most significantly to the dose for the IDF. This paper will summarize the information obtained for the 2016 data package and the plans for further testing to refine key inputs for the IDF PA. (authors)

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

New Formulation of Cast Stone for {sup 99}Tc Immobilization in Sulfate Rich Secondary Waste Stream - 16098

A low-temperature, cementitious waste form called Cast Stone (or grout) has been selected for solidification of the Effluent Treatment Facility (ETF)-treated wastes at the Hanford Site. Because the waste stream to be solidified may be relatively high in sulfate, the current Cast Stone dry blend formulation with ordinary Portland cement (OPC), fly ash (FA), and blast furnace slag (BFS) (47% BFS, 45% FA, and 8% OPC) may not be adequate. Thus a new formulation, called secondary waste grout, was proposed and tested. The secondary waste grout uses hydrated lime [Ca(OH){sub 2}] to tie up the sulfate initially such that ettringite [Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26(H{sub 2}O)] forms rapidly instead of forming as the cementitious waste form cures. Three different liquid waste simulants (242-A evaporator, Environmental Restoration Disposal Facility [ERDF] leachates, and Hanford Tank Waste Treatment and Immobilization Plant [WTP] off-gas condensates) were prepared and solidified using this new dry blend formulation. Baseline grout dry mixes were 20 wt% hydrated lime (HL), 35 wt% OPC, and 45 wt% BFS, and two different water-to-dry mix (w/dm) ratios (0.5 and 0.6) of free water mass to mass of dry blend were tested. After curing for a minimum of 28 days, the grout solidified waste forms were characterized and used for Tc leachability testing using the EPA 1315 Method. X-ray diffraction analysis confirmed that ∼12-15% of ettringite was formed in early stage of 28-day curing for the grouts prepared with the high sulfate WTP simulant and HL addition. In addition, the new formulation of secondary waste grout prepared with HL (20%), OPC (35%), and BFS (45%) at either 0.5 or 0.6 water to dry mix ratio yields higher average Tc interval leachability index (LI) values (12.5 to 13.9) after 28 to 90 days of leaching using deionized water, which is higher than the desired {sup 99}Tc LI value of 12 for grout destined for burial in the Integrated Disposal Facility at the Hanford Site. (authors)

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

Radionuclide Behaviour During Thermal Treatment of LLW and ILW Using In-Container Vitrification - 19239

Thermal treatment is being considered in the UK for the treatment of a wide range of low and intermediate level wastes (LLW and ILW). To support this objective, a range of in-container vitrification trials were conducted to test a wide range of simulated waste streams. The vitrified products have been visually inspected and samples removed for chemical analysis. This paper presents preliminary results of this analysis with some comment on the spatial and chemical heterogeneity of the products. Treatment of a sludge simulant in a steel skip failed to fully melt the metal and produced a glass encapsulated waste-form with clearly visible porosity. Clinoptilolite produced a dense uniform glass but with some chemical heterogeneity. Some operational issues were encountered when treating water-rich sludge, but these were successfully resolved and resulted in a broadly homogenous vitrified product. Melting simulated mixed decommissioning waste produced a chemically homogenous product, despite the heterogenous nature of the waste stream. Uranium and concrete contained in a steel 'top hat' presented significant treatment issues. The top hat failed to fully melt, and this appears to have prevented the waste contained within the top hat from mixing with the glass external to the top hat. As with the steel skip containing sludge, this again left a glass encapsulated, rather than fully vitrified, product. Attempts to melt bulk PVC (an analogue of soft PCM) initially led to significant operational issues with the melt being terminated early and the waste apparently combusting rather than melting. This was resolved by pre-treating the PVC by shredding it and mixing this with the frit to give a larger contact area for melting. This approach produced a homogenous well mixed glass. Overall it appears the properties of the melt product are highly dependent on the physical and chemical properties of the waste stream being treated. (authors)

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

Drying of Crystalline Silicotitanate (CST) Beds by Air Flow

The Direct Feed Low-Activity Waste (DFLAW) process has been proposed to support early production of immobilized low-activity waste (LAW) at the Hanford Site. The tank farm contractor Washington River Protection Solutions, LLC (WRPS) is developing the Tank-Side Cesium Removal (TSCR) system to accomplish the treatment objectives. The TSCR system is being designed and built by AVANTech, Inc. for WRPS and will use a filter to remove any entrained solids in the LAW and will use columns of crystalline silicotitanate (CST) ion exchange media to remove 137 Cs. A key step in preparing used columns of CST for transportation and interim storage is draining and drying of the CST beds. The purpose of this study is to document testing activities that quantify the drying of full-height beds of CST material in a laboratory column that is scaled to represent the drying process of the planned full-scale CST columns to be used in the TSCR system. The purpose also includes demonstrating that CST material within the scaled column, when dried with air injection, will pass the U.S. Environmental Protection Agency’s Paint Filter Liquids Test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Characterization of CST Post-Processing AP-105 Hanford Tank Waste

The primary goal of the Tank Side Cesium Removal (TSCR) system, under development by Washington River Protection Solutions, LLC (WRPS), is to remove entrained solids and 137 Cs from the Hanford tank waste supernate to expedite production of low-activity waste. Ion exchange (IX) testing of 10.9 L of waste from Hanford tank 241-AP-105 (AP-105), performed by Pacific Northwest National Laboratory, used a lead-lag-polish column format, with a bed volume of 10 mL per column, to decontaminate tank waste supernate using crystalline silicotitanate (CST) as the IX media. The AP-105 Cs IX processing test, discussed elsewhere, resulted in a shorter transition zone (i.e., steeper load curve) than those defined by wastes from tanks 241-AP-107 and 241-AW-102.1 The shorter transition zone was indicative of a matrix effect retarding Cs capacity. Therefore, aliquots of spent CST from the lead, lag, and polish columns were subjected to a digestion protocol to quantify analytes retained by the CST and extrapolate the impact on Cs capacity. The spent CST was digested using a combination of 5 M HNO 3 and H 2 O 2 with vigorous heating and stirring. Due to the radiation dose accompanying the 137 Cs on the CST columns, a secondary Cs separation by ammonium molybdophosphate embedded in polyacrylonitrile (AMP-PAN) was performed to separate the 137 Cs from the CST so the samples could be contact-handled for analysis outside of a shielded facility.

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

Conversion of a wet waste feedstock to biocrude by hydrothermal processing in a continuous-flow reactor: grape pomace

Wet waste feedstocks present an apt opportunity for biomass conversion to fuels by hydrothermal processing. In this study, grape pomace slurries from two varieties, Montepulciano and cabernet sauvignon, have been converted into a biocrude by hydrothermal liquefaction (HTL) in a bench-scale, continuous-flow reactor system. Carbon conversion to gravity-separable biocrude product up to 56 % was accomplished at relatively low temperature (350 C) in a pressurized (sub-critical liquid water) environment (20 MPa) when using grape pomace feedstock slurry with a 16.8 wt% concentration of dry solids processed at a liquid hourly space velocity of 2.1 h-1. Direct oil recovery was achieved without the use of a solvent and biomass trace mineral components were removed by processing steps so that they did not cause processing difficulties. In addition, catalytic hydrothermal gasification (CHG) was effectively applied for HTL byproduct water cleanup using a Ru on C catalyst in a fixed bed producing a gas composed of methane and carbon dioxide from water soluble organics. Conversion of 99.8% of the chemical oxygen demand (COD) left in the aqueous phase was demonstrated. As a result, high conversion of grape pomace to liquid and gas fuel products was found with residual organic contamination in byproduct water reduced to <150 mg/kg COD.

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

Dry Pretreatment of Used Nuclear Fuel for Tritium Removal: A Review

Virtually all the elements in the periodic chart are generated as a result of atomic fission in nuclear fuel. Several of the elements are volatile or can readily react with other chemicals to form volatile molecules that can be released during the processing of used nuclear fuel for the recovery of the valuable products. Among the volatile fission products, {sup 129}I, {sup 14}C, {sup 3}H, {sup 85}Kr, and {sup 79}Se constitute the main environmental concern for uncontrolled release. Storage of used fuel allows the short-lived isotopes to decay significantly (e.g., {sup 3}H decays in half every 12.32 years and {sup 85}Kr every 10.76 years). However, depending on regulatory limits they may remain an environmental concern up to about 50-100 years of fuel storage. In contrast, the concern for environmental release of the highly mobile {sup 129}I and {sup 14}C remains for geological times (half-lives of 15.7 million years for {sup 129}I and 5730 years for {sup 14}C). Processing of used commercial nuclear fuel requires significant head-end operations to remove the fuel from the protective coverings (e.g., assemblies). The current industrial process includes cropping the end caps of the fuel assemblies, chopping of the fuel pins and/or bundles into short segments, and prolonged leaching of the segments using concentrated nitric acid (HNO{sub 3}). The infrastructure, processes, and waste treatment systems in the head-end operations, including the associated off-gas system, are some of the larger and more costly components of the entire plant. The bulk of the volatile fission products are released to the off-gas during the leaching and dissolution. However, a small but significant amount will transfer to the following steps of processing such as solvent extraction and will show up in many different waste streams. The effective capture of {sup 3}H and iodine throughout the different processes and disposition of associated waste streams is costly and complicated. Up-front removal and capture is highly desirable. Currently, at the U.S. Department of Energy national laboratories and internationally, several dry pretreatment technologies are being studied and developed with the potential to simplify and increase the safety of subsequent storage or chemical recycling processes. These studies include (1) alternative decladding methods, (2) oxidation of exposed fuel following either conventional shearing or chemical decladding, and (3) the collection and stabilization of released volatile radionuclides. Several dry processes are being developed for the up-front removal of tritium and other fission products. They can be combined with different decladding alternatives and, in general, produce a fine powder that can be fed to aqueous or pyrochemical separations processes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cesium Removal Performance Comparisons of Crystalline Silicotitanate Media Batches with Savannah River Site Waste Simulant

The Tank Closure Cesium Removal (TCCR) system uses ion exchange columns filled with Crystalline Silicotitanate (CST) media to process radioactive waste solutions for the removal of Cs137. TCCR currently focuses on dissolving Savannah River Site (SRS) Tank 10H waste (primarily sodium salt cake solids) within the tank followed by at-tank ion exchange column treatment. Plans are underway to prepare and install a second TCCR unit at SRS. Capacity and particle size differences exist between archived (IE-911) and more recently-prepared CST media batches (9120- B and 9140-B). Side-by-side comparison testing was performed to evaluate the cesium removal performance of each batch to aid in selecting the preferred CST batch and media characteristics to load into the second TCCR unit. Batch contact equilibrium and flow-through column tests have been conducted with three CST batches using an SRS Average Simulant. Simulant batch contact equilibrium cesium loading results for the three CST batches (including two lots of one batch) are provided in Table ES-1. The 35 °C data indicates that the archived IE-911 batch has a higher cesium capacity than recently-prepared CST media and that the minor TCCR CST 9120-B media lot (2099000035) has similar cesium removal performance to the major TCCR lot (2099000034). Tests conducted at 25 ºC for the archived IE-911 CST batch indicated lower cesium removal performance with this simulant batch than was observed recently with a different SRS Average Simulant batch. A dilution factor (DF) is typically utilized when modeling engineered CST media cesium loading performance to account for mass contributions from the binder material. In cases where CST performance is lower than expected, this factor includes corrections for the binder and for low performance. The ZAM (Zheng, Anthony, and Miller) Isotherm Model DF values are provided in Table ES-1 for each simulant batch contact result. DF values near 0.5 were determined for the 9120-B and 9140-B CST batch contact tests while DF values near 0.6 (20% higher) were determined for the tests with IE-911. These dilution factors are lower than recently observed with a different SRS Average Simulant batch (9140-B DF = 0.68; IE-911 DF = 1.0). Based on these results, it appears that some component in the simulant solution used for equilibrium testing may have resulted in reduced cesium loading on the CST media.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2004

Improved Method of Purifying Carbon Nanotubes

An improved method of removing the residues of fabrication from carbon nanotubes has been invented. These residues comprise amorphous carbon and metal particles that are produced during the growth process. Prior methods of removing the residues include a variety of processes that involved the use of halogens, oxygen, or air in both thermal and plasma processes. Each of the prior methods entails one or more disadvantages, including non-selectivity (removal or damage of nanotubes in addition to removal of the residues), the need to dispose of toxic wastes, and/or processing times as long as 24 hours or more. In contrast, the process described here does not include the use of toxic chemicals, the generation of toxic wastes, causes little or no damage to the carbon nanotubes, and involves processing times of less than 1 hour. In the improved method, purification is accomplished by flowing water vapor through the reaction chamber at elevated temperatures and ambient pressures. The impurities are converted to gaseous waste products by the selective hydrogenation and hydroxylation by the water in a reaction chamber. This process could be performed either immediately after growth or in a post-growth purification process. The water used needs to be substantially free of oxygen and can be obtained by a repeated freeze-pump-thaw process. The presence of oxygen will non-selectively attach the carbon nanotubes in addition to the amorphous carbon.

Delzeit, Lance D.↗