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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Decomposition of PVDF to delaminate cathode materials from end-of-life lithium-ion battery cathodes

The growing demand of electric vehicles and rapid consumption of rechargeable lithium-ion batteries (LIBs) require recycling of spent cathode active materials (CAMs) to reduce hazardous wastes and supply raw materials to LIB production. To separate CAMs from the cathode, direct calcination of polyvinylidene fluoride (PVDF) binder is widely applied, which leads to high energy consumption and release of toxic hydrogen fluoride. It is desirable to have an environmentally friendly and effective alternative to traditional direct calcination. In this study, five lithium salts, LiOAc (lithium acetate), LiNO 3 , LiCl, Li 2 CO 3 , and Li 2 SO 4 , were deployed and compared for their performance in recycling CAMs. A peel-off efficiency of up to 98.5% was achieved at a LiOAc to LiNO 3 molar ratio of 3:2, salt to cathode mass ratio of 10:1, and temperature of 300 °C at a holding time of 30 min. This system avoids corrosive chemicals and minimizes particle agglomeration of recycled products. Compared with sodium salt systems (NaOAc-NaNO 3 ) or direct calcination, the LiOAc-LiNO 3 system prevented high reaction temperature or further lithium loss and minimized crystal structure and morphological changes. In conclusion, a decomposition mechanism of PVDF through adsorption of HF and fluorine substitution was proposed.

25 ENERGY STORAGE↗
DOE OSTI2018

An Efficient, Solvent-Free Process for Synthesizing Anhydrous MgCl 2

A new efficient and solvent-free method for the synthesis of anhydrous MgCl2 from its hexahydrate is proposed. Fluidized dehydration of MgCl 2 ·6H 2 O feedstock at 200 °C in a porous bed reactor yields MgCl2·nH2O (0 < n < 1), which has a similar diffraction pattern as activated MgCl2. The MgCl 2 ·nH 2 O is then ammoniated directly using liquefied NH 3 in the absence of solvent to form MgCl 2 ·6NH 3 . Calcination of the hexammoniate complex at 300 °C then yields anhydrous MgCl 2 . Both dehydration and deammoniation were thoroughly studied using in situ as well as ex situ characterization techniques. Specifically, a detailed understanding of the dehydration process was monitored by in situ PXRD and in situ FTIR techniques where formation of salt with nH 2 O (n = 4, 2, 1, <1) was characterized. Given the reduction in thermal energy required to produce dehydrated feedstock with this method compared with current strategies, significant cost benefits are expected. Overall, the combined effect of activation, macroporosity, and coordinated water depletion allows the formation of hexammoniate without using solvent, thus minimizing waste formation.

Motkuri, Radha K.↗
DOE OSTI2025

Solvation governs cation transference in glyme-based lithium battery electrolytes

The efficacy of electrochemical systems is governed by the cation transference number, which represents the fraction of current carried by the working ion. Energy is wasted when field-induced motion also drives anions and solvent molecules, decreasing the transference number to near-zero. We present a systematic study of cation transference in a series of electrolytes: tetraglyme (TG), octaglyme (OG), and poly(ethylene oxide) (PEO) mixed with lithium bis(trifluoromethanesulfonyl)imide. In all three electrolytes, starting from the dilute salt concentration limit, the experimentally measured cation transference number decreases with increasing concentration, reaching a minimum between -0.1 and -0.2, before rising back to positive values. Explicit measurements of field-induced species' velocities by electrophoretic nuclear magnetic resonance indicate that negative cation transference numbers in TG and OG electrolytes are dictated by solvation interactions with minimal contribution from anion-cation interactions. Simulation-based solvation structures indicate that OG serves as a bridge between TG and PEO. Multi-charge positive clusters, which are negligible in TG, become increasingly important at higher chain lengths (OG and PEO). As migrating cations drag their solvation shells, this solvation-induced motion is amplified in glyme electrolytes because of covalent interactions between solvating glyme molecules and free glyme molecules.

Im, Julia↗
DOE OSTI2018

Glass-bonded iodosodalite waste form for immobilization of 129I

Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. The pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.

sodalite, iodosodalite, waste forms, iodine↗
DOE OSTI2018

Inorganic Ba-Sn Nanocomposite Materials for Sulfate Sequestration from Complex Aqueous Solutions

Selective sequestration of sulfate (SO42-) in the form of barite (BaSO4) from alkaline solutions of high ionic strength containing carbonate at high concentrations resembling legacy radioactive waste generated during plutonium cold war era production is problematic due to the preferential formation of BaCO3. Incorporation of sulfate into the insoluble and thermally stable BaSO4 phase can potentially benefit radioactive waste processing in reducing operational challenges and suppressing volatilization of other waste components such as technetium-99. To enhance selectivity of SO42- sequestration, a series of Ba(II)-Sn(IV) materials was prepared using simple hydrothermal synthesis from different Sn(II) and Sn(IV) precursors. Structural characterization indicated that all obtained products predominantly contained BaSn(OH)6 and Ba2SnO2(OH)4(H2O)10 crystalline phases which were disrupted upon exposure to SO42- due to formation of BaSO4. Performance of the Ba(II)-Sn(IV) materials was tested using complex alkaline solutions simulating radioactive waste containing 0.094 M SO42- and 0.5 M CO32- among other constituents. About 54 – 66% of SO42- was converted to BaSO4 when a quantity of Ba(II)-Sn(IV) material containing approximately a stoichiometric amount of Ba2+ relative to SO42- was used. In comparison, previous studies indicate negligible BaSO4 formation under similar conditions when a simple Ba2+ salt is used. This improvement is attributed to the selective replacement of the Ba(II) stannate by SO42-. Thermal stability of the sulfate-loaded product material up to 1100 °C was demonstrated. Obtained materials offer a convenient and economical option for the selective sequestration or removal of SO42- from carbonate containing solutions.

Johnson, Isaac↗
DOE OSTI2019

Analysis of Tank 38H (HTF-38-19-18, -19) and Tank 43H (HTF-43-19-20, -21) Samples for Support of the Enrichment Control and Corrosion Control Programs

SRNL analyzed samples from Tank 38H and Tank 43H to support ECP and CCP. Overall, the results of the analysis indicate the concentrations of species in the Tank 38H surface sample increased significantly from the previous sample with the sodium concentration increasing from 1.09 M in the previous sample to 3.19 M in the current sample. However, it should be noted that the previous Tank 38H surface sample was very dilute compared with typical Tank 38H samples of the recent past. The Tank 38H sub-surface sample shows a composition fairly similar to the previous sample’s composition. Both the Tank 43H surface and sub-surface samples appear slightly more dilute than the previous samples based on the sodium concentrations. For both Tank 38H and Tank 43H, the subsurface samples appear slightly more concentrated for most species than the surface samples. These results indicate some success in increasing the salt solution concentration and reducing stratification in Tank 38H, but some stratification is still evident in both Tank 38H and Tank 43H.

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

Removal of Tritium from the Gaseous Effluents Produced by the Recovery of Zirconium from Used Fuel Cladding

U.S. regulations will require the removal of tritium ({sup 3}H) from gaseous effluents associated with nuclear fuel reprocessing and recycling. One process that may generate {sup 3}H-containing gaseous effluents is the recovery of zirconium from used fuel cladding. A chlorination process developed by Collins et al. is designed to convert the zirconium present in Zircaloy{sup R} cladding to recoverable ZrCl{sub 4}. The greatest economic benefit from including zirconium recycle as part of a larger used nuclear fuel (UNF) recycle effort would be a decrease in the disposal cost of used fuel cladding. Used fuel cladding is considered high-level waste (HLW). As zirconium is the second largest element by mass (25%) contained in most UNF assemblies (with uranium being the largest), recovery of the zirconium could result in a substantial decrease in the amount of HLW intended for geologic disposition. The reclaimed ZrCl{sub 4} product, if sufficiently free from radioactive impurities, could then be disposed of as low-level waste or reused in nuclear applications. The chlorination process uses chlorine gas (Cl{sub 2}) and an inert carrier such as argon (Ar) to convert the zirconium present in cladding to ZrCl{sub 4}. ZrCl{sub 4} is more volatile than other components in the cladding, causing evaporation of a relatively pure gaseous ZrCl{sub 4} stream. The product ZrCl{sub 4} is recovered as a solid salt by condensation when cooled to less than about 300 deg. C. A fraction of the {sup 3}H (0-96%) produced in nuclear fuel during irradiation is found in zirconium-based cladding, possibly as zirconium hydride (the wide range is a remnant of various researchers' analyses of cladding of various irradiations and cooling times). Heating the cladding to high temperatures, such as would be done during the chlorination process, will release the {sup 3}H. At the temperatures of the zirconium recycle process, the {sup 3}H is likely to react with excess Cl{sub 2} to produce tritiated HCl ({sup 3}HCl), although this could be avoided if the cladding and fuel were treated to remove {sup 3}H prior to chlorination. A previous analysis showed that the concentration of {sup 3}HCl in the off-gas of the zirconium recycle process will be in the range of parts per billion to a few parts per million. Should {sup 3}H mitigation be required for the zirconium recycle process, it will be completed through the abatement of {sup 3}HCl from this off-gas stream. The intent of the testing described in this summary was to conduct proof-of-principle demonstrations for two {sup 3}HCl removal methods. The zirconium recycle off-gas stream differs from many industrial HCl-containing streams in that the concentration of HCl requiring removal is extremely low. Demonstrations of HCl removal at ppm concentrations were not found in the literature, making the proof-of-principle approach attractive as a way to identify the most promising method for continued development. An important consideration throughout these experiments was that current research efforts on the zirconium recycle process are conducted in a hot cell and any {sup 3}HCl removal methods should be easily deployed in a hot-cell environment.

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

Molten salt electrochemical upcycling of CO 2 to graphite for high performance battery anodes

The efficient transformation of CO 2 into a value-added material is a potential strategy to help mitigate climate effects caused by CO 2 emissions. One potential CO 2 conversion product is graphite which is an important and versatile material extensively used in many applications including as an anode for lithium-ion batteries (LIBs). Commercial graphite, however, is traditionally synthesized via the energy intensive Acheson process (>3000 °C) and the performance of such graphite can be limited under fast charging conditions which is important for vehicle electrification. We report the electrochemical transformation of CO 2 to highly crystalline nano-graphite with a controlled microstructure in a carbonate molten salt at 780 °C. The use of a nickel foam electrode and controlled electrochemical parameters during the molten salt conversion process yielded pure graphite at a lower temperature compared to the Acheson process. Moreover, when investigated as an anode material for LIBs, the CO 2 -converted graphite exhibited high reversible capacity, long cycle life, and excellent rate capability even under fast charging conditions. This process provides a way to potentially reduce carbon emissions through the utilization of waste CO 2 by converting it into value-added graphite suitable for fast charging, high-energy-density batteries for vehicle electrification.

25 ENERGY STORAGE↗
DOE OSTI2021

Spent Crystalline Silicotitanate Storage Study—Post AP-105DF Processing

A Tank-Side Cesium Removal (TSCR) system is under development by Washington River Protection Solutions to support initial production of immobilized low-activity waste (LAW) by feeding Hanford tank supernate from tank farms to the Hanford Waste Treatment and Immobilization Plant (WTP) LAW Facility. Tank waste supernate will be filtered to remove suspended solids and then Cs will be removed by processing it through crystalline silicotitanate (CST) ion exchange media manufactured by Honeywell UOP, LLC. The Cs-loaded CST columns will be stored indefinitely, with a goal of eventual CST removal and treatment. Thus, the spent CST needs to be recoverable after undetermined storage time. Previous testing with AP-105 simulant showed that rinsing the CST bed with 3 bed volumes (1.4 apparatus volumes [AVs]) of 0.1 M NaOH resulted in a dried bed that maintained flow characteristics indicative of ease of recovery. This study explored the intermediate conditions (between feed dried in place and the 1.4 AVs of 0.1 M NaOH rinse) to evaluate CST bed properties after: 1) stoppage with feed in place; 2) stoppage after draining feed; 3) stoppage after 0.7 AV of 0.1 M NaOH rinse through column; 4) stoppage after 1.5 AVs of 0.1 M NaOH rinse through column. Post processing, each column was heated at 50 °C for 19 days under pseudo-storage conditions to simulate the expected dried and stored CST bed conditions. Testing was conducted at the small scale (12-mL bed volume); actual, Cs-depleted, AP-105 tank waste was used as the feed. Post-dried CST bed physical properties (angle of repose and penetration depth) were measured to evaluate how CST moved and flowed. All process stop-conditions resulted in a solidified CST bed except for the final condition, 1.5 AVs of 0.1 M NaOH rinse. At this small scale, the three CST beds presented an issue for retrievability after the short storage period (19 days at 50 °C). The latter case confirmed the results from simulant testing. The testing was intended to provide a preliminary assessment of issues that may arise from desiccation of CST during storage with the indicated salt solutions in place. Since these were small-scale tests, the processing system did not scale to full scale conditions exactly; however, the tests did provide insight into the impact on the dried and stored CST bed after stopping processing at an earlier step (upset condition) than normal. These results indicate that if an upset condition occurs at TSCR, a dilute hydroxide rinse should be considered before the CST dries from internal heating.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Analytical Results from Final Salt Batch 9 Routine DSSHT and SEHT Monthly Samples

Strip Effluent Hold Tank (SEHT) and Decontaminated Salt Solution Hold Tank (DSSHT) samples from several of the “microbatches” of Integrated Salt Disposition Project (ISDP) Salt Batch (“Macrobatch”) 9 have been analyzed for 238Pu, 90Sr, 137Cs, elemental constituents (Inductively Coupled Plasma Emission Spectroscopy - ICPES), and in some cases anions (Ion Chromatography Anions - IC-A).

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

Wash Penalty Factors: Insight into the Effects of Water Wash Down of Building Materials for Radionuclide/Nuclear (Rn) Decontamination - 20153

The release of radiological material to the urban environment can have a devastating socio-economic impact. In an urban release scenario, areas around the dispersal epicenter and any areas contaminated via atmospheric deposition may require extensive decontamination. Most recently, our group reported the decontamination efficacy of pressurized water washing of common porous building materials, and compared the penetration depth of the contaminations before and after exposure to pressurized tap water. Here, we continue evaluation of the decontamination efficacy of wash solutions based on potassium (K{sup +}) salt for cesium (Cs), strontium (Sr), and europium (Eu) contaminations on common building surfaces. The surfaces included: the cement used in concrete, the aggregate used in concrete, intact concrete, roadway asphalt, asphalt shingles, and latex paint. We deposited solutions of {sup 137}Cs, {sup 85}Sr, and {sup 152}Eu onto numerous common building materials and exposed these materials to a static bath or low-pressure flow of tap water, 0.1 M potassium chloride (KCl), and 0.5 M KCl. The decontamination efficacy and the depth profile for residual contamination were measured to determine the conditions under which applying a wash solution has benefit compared to physically removing the surface material. To aid in these analyses, we introduced a 'wash penalty factor' to quantify the degree to which contamination migrated into the bulk material by the action of the wash method. On asphalt, 70-80% of the radionuclides were found to be within 0.02 mm of the surface. Cs{sup +} bonded to negative charge sites on the surface of the mineral aggregate (used in concrete) so decontamination was improved by ion exchange with K{sup +}. Sr{sup 2+} is soluble and removed by water, while not bonding to aggregate or demonstrating solubility in the liquid asphalt. Eu{sup 3+} is precipitated and not removed in gentle wash but displays mobility within the pore space that suggests its presence as a colloid. Concrete is more porous than asphalt, and 80% of the radionuclides were within 0.2 mm of the surface for {sup 137}Cs and {sup 152}Eu and 50- 80% for {sup 85}Sr. On concrete, Cs{sup +} bonds similarly to asphalt. Sr{sup 2+} bonds to the cement or precipitates from the alkaline pore water. Eu{sup 3+} sorbs to the cement and then precipitates. In brick, the results for Cs{sup +} and Sr{sup 2+} suggested that diffusion of ions from deeper in the coupon was important. Eu{sup 3+} is removed poorly due to precipitation. We will explain and report example wash penalty factors that vary from 0 to > 1 for the various methods and the implications for deciding when to implement wash down methods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Closed-loop pressure retarded osmosis draw solutions and their regeneration processes: A review

Pressure-Retarded Osmosis (PRO) is an osmotic process that has been used to harvest energy from salinity gradients using a semi permeable membrane. A comparison between open-loop PRO (OLPRO) and closed-loop PRO (CLPRO) was made regarding their performance and costs. In CLPRO, where the diluted draw solution is re-concentrated in the regeneration system to be reutilized in the process, has recently received an intensive focus as the most viable configuration for a standalone power plant. The choice of the PRO draw solution in CLPRO is crucial to garner a high osmotic pressure as the key for the feasibility of the process. Here, in this review, the draw solutions are critically evaluated in the literature in terms of energy output as well as the method of regeneration used to recirculate them. A set of practical criteria has been suggested to appraise the adequacy of the solution for CLPRO application. It was concluded that NH 3 – CO 2 theoretically can produce 170 W/m 2 of power density. Inorganic draw solutes such as NaCl can generate high power density up to 87 W/m 2 . Organic draw solutes with their remarkably low reverse salt flux (RSF) have promising potential for future application in PRO. Similarly, the regeneration systems of the diluted draw solutions have also been reviewed and discussed. How the energy consumption of the regeneration process affects the feasibility of CLPRO is explained. For the specific case of osmotic heat engines (OHEs), when the energy of the regeneration process is supplied by heat waste, the range of applicability of the heat waste in CLPRO in terms of efficiency is defined and compared to Organic Rankine Cycle (ORC). The results showed that CLPRO has better efficiency than ORC for temperatures T < 80 °C, which makes it a promising process or low-grade heat energy recovery. In addition, a PRO-RO hybrid system coupled with solar power can reduce the net specific energy consumption (SEC) to 0.39 kWh/m 3 . The conditions that regeneration processes should operate under to make PRO viable are discussed in the last section. Overall, the study indicates the key factors for optimizing the performance of CLPRO process.

42 ENGINEERING↗
DOE OSTI2019

Energetics of electrochemically mediated amine regeneration process for flue gas CO 2 capture

The electrochemically mediated amine regeneration (EMAR) process presents an alternative route to the conventional thermal regeneration of amines for carbon capture from a flue gas source. In this study, chemical and electrochemical measurements were performed to validate the speciation and activity coefficient models that describe the solution thermodynamics. Electrochemical kinetics and efficiencies were identified and process level energetic calculations were performed using these validated models to understand the effects of various operating parameters on the energy demand for CO 2 capture. An isothermal EMAR process requires about 52 kJe/molCO 2 , and the net energy demand can be reduced to about 37 kJe/molCO 2 if waste heat is available at temperatures below 90 °C. The EMAR process is able to desorb CO 2 at pressures up to 20 bar with negligible additional energy penalty. These results suggest that the EMAR process is competitive with the conventional thermally regenerated amine processes. Preliminary assessments also indicate the solvent fomulation with NaCl requires less electric energy among other conventional supporting salt candidates (e.g. Na 2 SO 4 ) for electrolytic processes. Furthermore, advanced process designs, process optimization, and the identification of improved solvents may offer the potential to further reduce the energy demand for CO 2 capture.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Fiscal Year 2020 Filtration of Hanford Tank Waste 241-AP-105

Bench-scale filtration testing of 15 liters of 241-AP-105 supernatant was conducted using a Mott inline filter Model 6610 (media grade 5) backpulsed dead-end filter (BDEF) in the hot cells of the Radiochemical Processing Laboratory at Pacific Northwest National Laboratory. The as-received samples were diluted with raw water from the Hanford Site to assess the propensity for solids to form upon dilution and to assess the impact those solids have on filtration performance. The feed was split into two batches. The first batch was settled and decanted as planned in Tank Side Cesium Removal operations. The BDEF was used to filter this AP-105 feed at a targeted flux of 0.065 gpm/ft 2 and exhibited no measurable pressure increase during filtration. The second batch was fed to the BDEF with solids resuspended. This was also operated at the targeted flux of 0.065 gpm/ft 2 . In this case, the target pressure differential of 2 psid (resistance of 6.24 × 10 10 m -1 ) was reached in slightly more than 12 h. A backpulse restored the filter flux; however, the time to the next backpulse was slightly shorter. These tests indicate that after the addition of raw water, settle and decant effectively removed solids from the feed. It is not possible to assess the time required for these solids to settle based on the result from these tests. However, if sufficient time is allowed for the solids to settle, no measurable filter fouling was observed in current testing. In contrast, if insufficient time is allowed for solids to settle, filter fouling will occur relatively quickly and will likely require backpulses more frequently than every 12 hours (1.91 m 3 /m 2 volume filtered). Solids concentrated from backpulse solutions displayed large variability in composition with Al-silicates and mixed chromium-aluminum oxide phases. Amounts of Mg, Fe, and Ca were also found in the presence of Na and O. These solids could be grouped in three categories: (1) soluble Na-salts that formed large block-like particles but were made up of fibrous materials, (2) amorphous Al-phases, and (3) colloidal iron oxides with an average particle diameter of 4.2 µm.

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

Interactive web mapping tools and custom subsurface cross-sections for interdisciplinary geologic investigation

Using Python-based geospatial analytics, open-source web mapping technologies, geophysical data models, and subsurface stratigraphy models from the Regional Geology Geologic Framework Model database assembled by Los Alamos National Laboratory, we developed a suite of web-based geologic investigation tools to identify and understand subsurface structures and geophysical properties concerning salt and shale formations within the contiguous United States. Coupled with a web map interface, these tools allow for the interactive visualization of various geologic data and demonstrate the ability to quickly generate custom subsurface cross-sections, borehole charts, and diagrams for azimuthal orientation data. These capabilities were developed for stakeholder and researcher use to facilitate informed decision making for spent nuclear waste disposition. However, these capabilities provide a flexible model for a variety of subsurface investigation needs, and we have demonstrated this flexibility by adapting these tools to meet visualization needs for various subsurface models within a web-based platform.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

High-Fidelity and High-Performance Computational Simulations for Rapid Design Optimization of Sulfur Thermal Energy Storage

Industrial process heating (IPH) accounts for approximately 70% of US manufacturing energy use and is primarily produced by fossil fuel combustion. Approximately 1500 TWht (approximately 60%) of IPH demand is in the temperature range of 100-300. Industrial applications in this temperature range include drying, hydrothermal processing, thermal enhanced oil recovery, food and beverage, bioethanol production, etc. Cost-effective thermal energy storage (TES) that increases the utilization of waste and renewable heat (solar, geothermal, etc.) could provide significant energy savings and reliable heat sources, decrease emissions, and increase US manufacturing competitiveness through reductions in fuel consumption. TES development has historically been dominated by technologies suitable for deployment with concentrating solar power (CSP). State-of-the-art thermal storage deployed commercially with power tower CSP plants uses a 60%/40% NaNO3/KNO3 molten salt and operates between temperatures of approximately 280 degrees Celsius and 570 degrees Celsius using a two-tank configuration. However, these nitrate salts are unsuitable for operation outside of this temperature range due to a high freezing point of approximately 220 degrees Celsius, and limits on high-temperature salt stability and corrosion resistance of containment alloys. Other materials being investigated for TES include those based on: (1) sensible energy storage (various molten salt compositions, inert solid particles, rocks or pebble beds, sulfur, water, concrete, graphite, etc.), (2) latent energy storage in materials that undergo solid-liquid phase change at relevant temperatures (organic materials for low-temperature applications, inorganic salts and/or metals for high-temperature applications), or (3) thermochemical energy storage (hydrides, hydroxides, carbonates, metal oxides, etc.). The application temperature and challenges pertaining to storage material and/or containment cost, energy density, long-term thermal and cyclic stability, and charge/discharge heat transfer effectiveness drive material selection for a given IPH or electricity generation application. Sulfur is a cheap commodity at $80/ton compared to $1100 - 1300/ton for conventional salts. When using a metric of storage cost per kWh, sulfur costs around 2-3 $/kWh. Previous sulfur TES development focused on high temperature (>600 degrees) concentrated solar power applications with sulfur encapsulated in pipes and flow of gaseous HTF (air) in the shell side. However, for lower-temperature IPH applications in the range of approximately 100-300 degrees Celsius Element 16 adopted a compact and scalable TES design with molten sulfur in the shell and HTF pipes submerged in the molten sulfur bath. The low-cost molten sulfur TES for dispatchable IPH has deployment potential for broad applications. The spatial and temporal evolution of the HTF and sulfur temperature is critical to the TES system performance, and thus detailed modeling can improve understanding of the performance and facilitate design improvements. Using high performance computing and computational fluid dynamics (CFD) a low-cost molten sulfur thermal energy storage (TES) system for industrial process heating (IPH) applications was developed. The unique challenges in CFD modeling of sulfur TES are the sharp property changes of sulfur relevant to the working temperatures. Above 159, liquid sulfur undergoes polymerization, and the viscosity of sulfur rapidly increases by several orders of magnitude between 159 degrees Celsius and 188 degrees Celsius, followed by a decrease in viscosity beyond 188 degrees Celsius due to thermal bound dissociation. In addition, various concentrations of H2S impurities can also modify sulfur viscosity. This numerical challenge is especially relevant to transient simulation of the sulfur TES charging and discharging processes as the extreme property variations limit the applicability of traditional heat transfer correlations. Transient CFD simulations including the temperature-dependent sulfur properties and geometric complexity of the TES design were used to predict the effect of natural convection during charging and discharging on the heat transfer process, sulfur temperature uniformity, charge/discharge rates, and performance of the storage devices. The CFD model was validated with experimental results for a full charge and discharge cycle. The work will show 3D and 2D simulation comparisons aimed to facilitate rapid design iterations and a machine learning based design optimization approach.

CFD↗
DOE OSTI2024

Compositional Variation Tests on DuraMelter 100 with LAW Sub-Envelope A1 Feed (LAWA44 Glass) in Support of the LAW Pilot Melter, VSL-02R62N0-4, Rev. 0 (Nov 2002)

The two tests described in this report utilized blended feed (glass formers plus waste simulant) prepared by Optima Chemicals according to VSL specifications generating about 1. 7 metric tons of glass. Sugar was added (at VSL) to the nominal feed at a "sugar ratio" of 0.5 for each of the two variation tests1; however, since the sugar addition is assumed to be "blind" to the variations, the actual sugar ratios were 0.57 and 0.44. The DMl00-WV melter was used in order to provide a direct comparison with the LAW tests previously conducted on the same melter. Two 100-hour melter tests were conducted: one with a 15% deficiency in simulant and one with 15% excess in simulant. Key operating parameters including cold cap coverage, feed rate, and glass pool temperature were held constant to investigate the effects of the glass compositional changes on processing characteristics (including salt formation) and the product glass. The bubbling rate was adjusted to provide the desired glass production rate with a near complete cold cap (90-100% of melt surface covered with feed). Quantitative measurements of glass production rates, melter operating conditions (temperatures, pressures, power, flows, etc.), and off-gas characteristics (NOx, SO 2, CO, particulate load and composition, and acid gases) were made for each test.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Effects of Chemical Wastes on Concrete through Spectroscopic and Computational Models - 18405

Liquid mercury was used as a coolant for early fast reactors, as catalyst in the dissolution of uranium alloy fuels, lithium isotope separation and other processes. Cesium is produced during nuclear fission reactions within the reactor. The interaction between chemical contaminants and solid structures can cause structural damages (e.g. dislocations) and consequently modify the material's mechanical properties. Another problem associated with the interaction of chemical contaminants and metal structures is corrosion pitting. Mercury intrusion porosimetry (MIP) has been utilized to investigate the pore structure of the cement material. The overall goal of this project is to determine the physiochemical interactions of Hg and Cs to steel and concrete using spectroscopic analysis: - Evaluate the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques under standard conditions; - Evaluate the impact of environmental conditions (temp., salt content, pH, and time) on the physicochemical interactions of Hg and Cs to steel and concrete using spectroscopic techniques. Conclusion and Future Work: Adsorption process depends on the surface morphology. Surface morphology analysis demonstrated that concrete samples had a heterogenous surface. Adsorption of Mercury and Cesium are more favorable onto heterogeneous regions such as edges, grooves, and wrinkles. The mechanism for cation removal by concrete includes surface complexation and ion exchange. Overall adsorption of Mercury was more favorable than Cesium onto concrete. Continuing investigations will be focusing on application considerations such as multicomponent interactions and desorption studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗