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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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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Results for the September Bimonthly Calendar Year 2022 Tank 50 Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the 2022 September bimonthly sample of Tank 50 salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from Tank 50 to the Saltstone Production Facility (SPF), where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the Tank 50 sample was obtained. The chemical and radionuclide contaminant results from the characterization of the 2022 September bimonthly sampling of Tank 50 were requested by SRMC personnel via a Task Technical Request (TTR). Details of the testing are presented in the Savannah River National Laboratory (SRNL) Task Technical and Quality Assurance Plan (TTQAP). This Technical Report is the TTR deliverable relating Salt Solution Analysis from the SRMC request. The following facts pertaining to the WAC are drawn from the analytical results, including analytical uncertainty, provided in this report. 1) WAC LIMITS and TARGETS were met for all analyzed chemical and radioactive contaminants for which the detection limits are below the WAC LIMITS and TARGETS; 2) Measured average concentrations of nitrate, nitrite and total mercury are approximately 24%, 9% and 12% of the WAC LIMITS, respectively; 3) Measured average concentrations of Tc-99 and I-129 are approximately 19% and 20% of the WAC LIMITS, respectively; 4) All other radionuclide average concentrations are at 4% or less of the WAC LIMITS and TARGETS.

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

Chromium and Vanadium Incorporation in Sulfate-Containing Sodium Aluminoborosilicate Glass

In the composition space for Hanford low-activity nuclear waste glass, the waste loading of some formulations is limited by poor incorporation of sulfate. In particular, certain elements identified in previous studies tend to raise or lower sulfate solubility in borosilicate glass. The aim of the current study is to understand how vanadium and chromium each affect a sulfate-containing sodium aluminoborosilicate glass structure. Glass transition temperature, mass density, visible absorption, and Raman scattering were measured to investigate changes in the glass structure. Both Cr6+ (CrO42-) and Cr3+ were found in the chromium-containing glasses while vanadium primarily existed in the 5+ oxidation state in the vanadium-containing glasses. Electron probe microanalysis determined elemental compositions to assess retention of components. Increasing Cr2O3 caused saturation of Cr concentration within the glass and formation of crystalline eskolaite (Cr2O3). On the other hand, all the targeted V was incorporated into the glass. A fraction of batched S was not incorporated in any glass, presumably due to volatilization during melting. Apparently, the details of incorporation of Cr, and to an extent V, are different depending on whether sulfur is added as in this study or in oversaturated conditions as described in the literature, where Cr partitions to a sulfate-containing salt phase.

Borosilicate glass, Waste Glass, vanadium chemistr↗
NASA NTRS1992

Removing Solids From Supercritical Water

Apparatus removes precipitated inorganic salts and other solids in water-recycling process. Designed for use with oxidation in supercritical water which treats wastes and yields nearly pure water. Heating coils and insulation around vessel keep it hot. Locking bracket seals vessel but allows it to be easily opened for replacement of filled canisters.

Hong, Glenn T.↗
NASA NTRS2003

Less-Toxic Coatings for Inhibiting Corrosion of Aluminum

Two recently invented families of conversion- coating processes have been found to be effective in reducing or preventing corrosion of aluminum alloys. These processes offer less-toxic alternatives to prior conversion-coating processes that are highly effective but have fallen out of favor because they generate chromate wastes, which are toxic and carcinogenic. Specimens subjected to these processes were found to perform well in standard salt-fog corrosion tests.

Minevski, Zoran↗
DOE OSTI2020

Fluorine Limits and Impacts in High-Level Waste Glass Compositions

The impact of elevated fluorine (F) content on Hanford high-level waste (HLW) glasses has not previously been studied in detail. This effort represents the first systematic study to determine what F concentration limits should be used for the design of alkali-borosilicate-based Hanford Waste Treatment and Immobilization Plant (WTP) HLW glasses, and to document the technical basis for that limit. If alkali borosilicate glass made from Hanford HLW can accommodate a large amount of F, the large capital costs for complex sludge washing facilities may be avoided, as would much of the operational costs and negative schedule impacts associated with handling the large volumes of water required to dissolve these salts. In order to determine a limit for F in likely HLW glass compositions, an evaluation was conducted on glasses with F ≤ 0.90 mass% from previous nuclear waste glass studies. The collected dataset contains 239 glasses (232 HLW glasses and 7 LAW glasses) including 109 glasses with 0.9 ≤ F mass% ≤ 2.5, 116 with 2.5 < F mass% ≤ 8.0, and 14 with F mass% ≥ 8 (max. F mass% = 17.42). The collected composition and property data were analyzed to determine the basis for the F tolerance, i.e. the maximum F concentration that can be processed without potential issues. Fluorine volatility, product consistency test (PCT) response, liquidus temperature (T L ), glass melt viscosity, and crystallinity have been investigated. No limits for F concentration can be made based on F volatility, T L , or glass melt viscosity, because the data show that high F in glasses do not indicate, with high probability, being restricted by those property constrains. However, crystallinity and PCT response were used to estimate the F tolerance. The results show that glasses with high F (≥ 0.90 mass%) are more likely to form large fractions of F-containing crystal phases which may increase PCT responses, i.e. decrease the glass durability. Based on the results of crystallinity and PCT data, the F tolerance of 4.5 mass% is estimated. There is no evidence of other glass components, such as calcium oxides and alkali metal oxides have combined impacts with F on the glass properties. Overall, the available high-F glass data is limited, especially in the designed HLW glass composition regions. Future work on formulation and testing of HLW glasses with F ≥ 0.9 mass% will close the data gaps and expand operational flexibility with respect to the fluoride tolerances. Volatility of F from melters and corrosion of materials in contact with glass melts are important for processing of high-F wastes; yet no test data are currently available. It is recommended tests be conducted to address these two potential issues.

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

Brine Availability Test in Salt (BATS) (FY25 Final Closeout)

This fiscal year 2025 (FY25) closeout report summarizes the final experiments conducted and data collected in phase 2 of the Brine Availability Test in Salt (BATS 2). The design and interpretation of the test was previously funded by the US Department of Energy’s Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Disposition Program, under the Disposal Research and Development (R&D) program of the Office of Spent Fuel & Waste Science and Technology (SFWST) but they are not funding it anymore (since May 2025). The experiment was located underground at the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico, which is a DOE Office of Environmental Management (DOE-EM) site managed by the Carlsbad Field Office (CBFO). DOE-EM funds the WIPP Test Coordination Office (TCO), which provided critical implementation support for the continued execution of BATS. A high-level summary of processes expected in a salt repository are given in Kuhlman & Mills (2025).

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

Investigation of neutron noise in a micro-scale, natural circulation molten salt fission battery system

The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. A specific novel micro-MSR, natural circulation, battery design concept involves placing all primary components within a single reactor vessel containment without the need of primary forced pumping. This small modular, integral design presents potential cost savings while producing safety, reliable, and transportable carbon-free power for decades. This investigation evaluates the neutron noise induced by density and flow fluctuations in a natural circulation MSR battery concept being developed at the University of Idaho. This study finds numerical solutions to the one-dimensional, one-group, coupled diffusion equations for a bare, homogeneous core, and evaluates neutron flux and delayed neutron precursor concentration noise due to core flow and fuel salt density perturbation sources. Noise analysis shows both point kinetic and space-dependent behavior is present despite the small size, low-flow, closely-coupled nature of the natural circulation MSR concept. Analysis shows that the low flow of the convective system results in noise behavior typically representative of a larger forced-circulation MSR system. Finally, noise techniques may also be useful in MSR core diagnostics as a non-invasive, low-cost radiological option to traditional monitoring schemes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Multi-Physics Investigation of a Natural Circulation Molten Salt Micro-Reactor that Utilizes an Experimental In-Pile Device to Improve Core Physics and System Thermal-Hydraulic Performance.

INL employee PhD Dissertation - The Molten Salt Reactor (MSR) concept is a rapidly evolving Generation IV design that has recently attracted favorable attention due to the potential for reducing waste generation, realizing passive safety features, and seizing on the opportunity for cost effective economics. This thesis investigates the performance benefit of a new device invented by the doctoral candidate. The device is referred to as a Wrapped Helix around an Inclined Plane or WHIP. The WHIP is protected under a provisional patent filed with the USA Patent Office on 28 September 2021 under application number 63/261,776, BEA docket number BA-1254. The WHIP device can be located in-core or near-core to promoted enhanced thermal-hydraulics and neutronics performance. While the WHIP can be employed in a variety of solid or liquid fuel designs, this thesis investigates the device’s benefit in the application of a natural-circulation, micro-molten salt nuclear battery concept (MsNB). This thesis will specifically investigate the temperature coefficient of the MsNB fuel (FLiNaK) using novel temperature sensitivity techniques unique to Serpent particle transport code, evaluate the WHIP’s thermal-hydraulic and neutronic performance effects using both established (STAR-CCM+) and novel (Python code developed to estimate the circulation effective delayed neutron fraction, ßef f ) analytical and numerical methods, evaluate the neutron noise behavior of the MsNB and how the WHIP may alter the character of the MsNB’s transfer function, and how the WHIP affects the autonomous, load-following performance under transient power conditions using Python code developed by the candidate. The sum body of this work, in part, has been published in three journal articles as the timing of the provisional patent process has allowed. Results show that creative utilization of WHIP engineering design and function reduces reactor system volume, fuel loading, control/stability in the buoyant regime

buoyant flow↗
DOE OSTI2023

SCALE depletion capabilities for molten salt reactors and other liquid-fueled systems

Nuclear reactor systems that use fuel dissolved in a liquid have the potential for enhanced safety characteristics, improved fuel-cycle outcomes, and more efficient isotope-production configurations. In these reactor systems, the fueled liquid may simultaneously undergo irradiation, physical and chemical removal processes, and fueling. The modeling and simulation of this transmutation and decay with material additions and removals is an ongoing research area. An accurate simulation tool is critical to the reactor and fuel-cycle design, reactor deployment, and source-term characterization for these advanced reactor systems. The work described herein involved implementing, testing, and applying the capability to perform reactor physics simulations within the Oak Ridge National Laboratory-developed SCALE suite for nuclear systems analyses and design, leveraging much of its pedigree in quality-assurance and reactor-analysis capabilities. The functionalities to simulate irradiation with material feeds and removals had been added in ORIGEN, and the TRITON reactor physics sequence was extended to calculate the total removed material and track external nonirradiated mixtures to estimate separate processing or waste streams. Results from these capabilities align with analytical expectations obtained from ORIGEN for simplified test cases and with expectations for a molten salt reactor application. This implementation, available with the SCALE 6.3 release, provides for a more efficient and accurate material accountability methodology, allowing for the characterization, design, and analysis of the complete isotopic material inventory of advanced liquid-fueled systems for a variety of applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing, and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create a stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes the PE for high-temperature processing. After removing the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using a Fe-based low-temperature (<1500 °C) catalytic process. The PE-derived graphite powder showed excellent electrochemical performance as an anode material for lithium-ion batteries (LIBs) with a capacity of up to 302 mAh/g at 0.5 discharge/charge cycles per hour (0.5 C) and capacity retention of 100% after 415 cycles. This method illustrates there are opportunities for upcycling large quantities of PE waste to produce graphite powders.

Huynh, Ngoc Tien [NETL Site Support Contractor, Na↗
DOE OSTI2024

Evaluating Iodine Immobilization Technologies: Cermets, Polycermets, and Polyhalmets

The work in this report documents the efforts conducted to assess the feasibility of some of the ideas documented in Pacific Northwest National Laboratory invention disclosure reports (IDRs) including: 1) Iodine capture in polyacrylonitrile (PAN)-containing composite sorbents (32451-E). In this work, the composites evaluated included Ag0, Bi0, Cu0, Bi2S3, and Cu2S embedded in PAN. 2) Metal iodide removal from these sorbents through dissolution in dimethyl sulfoxide (DMSO) (32729-E). In this work, PAN dissolution was evaluated for multiple types of sorbents including Ag-Pan, Bi-PAN, Cu-PAN, Bi2S3-PAN, and Cu2S-PAN. 3) Using metal-sulfide sorbents for iodine capture (32647-E). In this work, the composites evaluated under this IDR included Ag2S, Bi2S3, and Cu2S embedded in PAN. 4) Using low-melting metals to immobilize (encapsulate) iodine-loaded and polymer-containing sorbents into polymer-ceramic-metal (called polycermet) or polymer-halide-metal (called polyhalmet) composite waste forms (32625-E). In this work, the iodine-loaded PAN composites included AgI-PAN, BiI-PAN, and CuI-PAN. 5) Ceramic-metal composite waste form synthesis of polymer-containing materials using low-melting metals like bismuth, tin, or bismuth-tin alloys (32537-E). In this work, the metals evaluated included Bi, 58Bi-42Sn eutectic. 6) Cermets for immobilizing commercial sorbents loaded with radioiodine (32806-E). In this work, AgIX (iodine-loaded silver faujasite zeolite) was evaluated in cermet form.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Modernization at the Y-12 National Security Complex: A Case for Additional Experimental Benchmarks

Efforts are underway at the Y-12 National Security Complex (Y-12) to modernize the recovery, purification, and consolidation of un-irradiated, highly enriched uranium metal. Successful integration of advanced technology such as Electrorefining (ER) eliminates many of the intermediate chemistry systems and processes that are the current and historical basis of the nuclear fuel cycle at Y-12. The cost of operations, the inventory of hazardous chemicals, and the volume of waste are significantly reduced by ER. It also introduces unique material forms and compositions related to the chemistry of chloride salts for further consideration in safety analysis and engineering. The work herein briefly describes recent investigations of nuclear criticality for 235UO2Cl2 (uranyl chloride) and 6LiCl (lithium chloride) in aqueous solution. Of particular interest is the minimum critical mass of highly enriched uranium as a function of the molar ratio of 6Li to 235U. The work herein also briefly describes recent investigations of nuclear criticality for 235U metal reflected by salt mixtures of 6LiCl or 7LiCl (lithium chloride), KCl (potassium chloride), and 235UCl3 or 238UCl3 (uranium tri-chloride). Computational methods for analysis of nuclear criticality safety and published nuclear data are employed in the absence of directly relevant experimental criticality benchmarks.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Industrial Waste to Value in the Circular Economy

This paper will outline a waste to value project currently underway. The project will profitably process 100,000 metric tons per year of secondary aluminum salt slag, which is currently landfilled, into recycled salt and refractory aggregate.

Ormond, Paul↗
DOE OSTI2025

Uncertainty improvement of 22 Na based radioactive tracer dilution for determining total mass of pyroprocessing molten salt systems by 154 Eu removal

To determine the total salt mass of the molten salt systems for pyroprocessing spent nuclear fuels, a 22 Na based radioactive tracer dilution was studied in Idaho National Laboratory in recent years. This 22 Na based RTD technique was deemed feasible, but due to the gamma energy peak of 22 Na coinciding with one of the energy peaks of 154 Eu radioisotope in the molten salt, the uncertainty of the 22 Na radioactivity in the 22 Na-spiked salt samples was quite high. To improve the uncertainty of the 22 Na based RTD technique, we proposed to chemically remove the 154 Eu of the salt samples by DGA resin for gamma spectroscopy. The effectiveness of removing 154 Eu on uncertainty improvement was evaluated. Furthermore, it was found that (1) the 154 Eu fission product effect on the uncertainty and detection limit can be effectively eliminated by chemically removing the 154 Eu during the salt sample preparation and (2) the uncertainty of 22 Na radioactivity in the salt samples for electrorefining was significantly improved from 13% to 2%, showing the potential of practical engineering application of 22 Na based RTD as a safeguards technique for molten salt systems for pyroprocessing spent nuclear fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Synthesizing Highly Crystalline Graphite Powder from Bulk Polyethylene Waste for Lithium-Ion Battery Anodes

Upcycling plastic waste into graphite can potentially be used, in conjunction with other methods, to manage existing waste materials and diversify graphite supply chains. However, synthesizing large quantities of crystalline graphite powder from plastic waste, particularly polyethylene (PE), remains a challenge because PE decomposes into light gases during thermal processing and simple methods do not exist at any appreciable size scale to address this challenge. In this work, a method is developed for air processing bulk forms of PE waste to create stable carbon char that does not readily decompose during high-temperature processing. This method employs solid additives in the form of salts, which are combined with the PE melt during air processing to increase the effective surface area of the melt and improve the oxygen-driven chemistry that stabilizes PE for high-temperature processing. After removal of the solid salt additives from the PE-derived char, it is converted into a highly crystalline bulk graphite powder using an Fe-based catalytic process. The PE-derived graphite anode in a lithium-ion coin cell showed a specific capacity of 345 mAh/g at 0.05C with an initial Coulombic efficiency of 87% and reversible capacity retention of ~100% at different current rates. It also showed a specific capacity of up to 313 mAh/g at 0.5 discharge/charge cycles per hour (0.5C) and Coulombic efficiency of 99.9% after 250 cycles, indicating excellent electrochemical performance as an anode material for lithium-ion batteries. This method illustrates that there are opportunities for upcycling large quantities of PE waste to produce graphite powders suitable for use in LIBs.

25 ENERGY STORAGE↗
DOE OSTI2019

Three Years of Multilevel Monitoring Data at the Riverton, Wyoming, Processing Site That Show Contaminant Increases After River Flooding Events and a Large Recharge Event

The tailings pile at the Riverton, Wyoming, Processing Site has been removed, but contaminant concentrations in the shallow aquifer increase after large recharge events. The Riverton site experienced flooding near the Little Wind River in early May 2016, February 2017, and June 2017. The May 2016 flooding was from a large rain event. The February 2017 flooding was due to an ice jam, and the June 2017 flooding was due to mountain snowmelt runoff. In addition, the site experienced high water tables in April and May 2017, as this was the second wettest spring in Riverton since records began in 1918. Past reports indicated the accumulation of solid-phase uranium and molybdenum in evaporites above the respective contaminant groundwater plumes. Given these data, multilevel monitoring wells were installed in 2015 to monitor seasonal changes in groundwater quality. Data from three years of groundwater monitoring are summarized in this report and compared with the flooding and large recharge events. The results indicate that evaporite-related constituents (especially sodium, sulfate, and chloride) present across the site in a silt layer can be released to the shallow sand and gravel aquifer during flooding and large recharge events. The solid-phase and water-phase data confirm that these evaporites are mainly sodium sulfate and sodium chloride salts with some calcite and gypsum.

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

Fundamental Study of Uranium Dendrites Electrochemically Separated in Molten Salt with the Presence of Cerium

In this study, uranium was electrochemically separated in the presence of cerium in eutectic LiCl-KCl at varying concentrations of cerium and varying overpotentials. Harvested dendrites were then examined and analyzed for their morphological features and elemental composition using various analytical techniques. Conclusions: Increasing the applied overpotential causes the dendrites to evolve from relatively simple and large platelet-like crystals (25 mV), to a grid-like network with crystals having jagged and smooth features (50 mV, 100 mV), and finally into a very chaotic and complex crystal morphology (200 mV) for all uranium-cerium compositions explored. Varying the uranium:cerium concentration ratio did not have much effect on the uranium dendrite morphologies. EDS results show that cerium is not co-deposited on the surface of the dendrite crystals, ICP-MS results do show a presence of cerium in the volumetric composition. This is likely due to salt occlusion during the dendrite formation. XRD has been performed on several samples at the 25 mV overpotential with 1 wt% U and 0 wt% Ce in the molten salt eutectic. These results indicate that there is not salt occlusion within the dendrites during electrochemical separation. Additionally, the uranium metal quickly oxidizes during preparation for XRD.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗