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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Optical spectroscopy and modeling of uranium gas-phase oxidation: Progress and perspectives

We report that studies related to U gas-phase oxidation through plasma- and thermo-chemistry are important for many fields, including environmental monitoring, forensic analysis, debris analysis in a weapon detonation event, and nucleation physics. Recently, significant efforts have been made to understand the chemical pathways involved in the progression from U atoms to diatoms (UO) and polyatomic molecules (UxOy), employing optical spectroscopy tools and computational modeling. In many studies, laser ablation of U or a U-containing flow reactor are used as a highly resource-efficient, repeatable, tunable, and lab-scale testbed for studying gas-phase oxidation in U plasmas. The spectroscopic analysis of high-temperature gas-phase oxidation of U is challenging due to the congested U spectra, resolution limitations of instrumentation, and the numerous chemical reaction pathways possible. This article focuses on the current understanding and challenges related to studying U plasma chemistry, specifically U gas-phase oxidation and molecular formation, via optical spectroscopy of plasmas and associated computational and spectral modeling. The physical and chemical processes involved in the evolution from U atoms to U oxide molecules to nanoparticles and agglomerates (i.e., debris) are discussed in the context of optical spectroscopic studies. The article concludes by highlighting opportunities for future research efforts based on existing knowledge published in the literature.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Shinkolobweite, from the Shinkolobwe Mine, Democratic Republic of Congo: A New Mineral Containing Uranium in the Rare Pentavalent Oxidation State

ABSTRACT Shinkolobweite, Pb1.333[U5+O(OH)(UO2)5O4.67(OH)5.33](H2O)5, is a new lead uranyl oxide-hydroxide hydrate mineral containing hexavalent and pentavalent uranium from the Shinkolobwe mine, Democratic Republic of Congo. Crystals of shinkolobweite are dark reddish-brown prisms up to 0.5 mm in length, occurring on a matrix of massive uraninite associated with fourmarierite, rutherfordine, soddyite, and sklodowskite. Crystals are translucent with subadamantine luster and light bronze-yellow streak, are flattened on {010}, are elongated on [001], and exhibit the forms {100}, {010}, {101}, and . The mineral is non-fluorescent under both longwave and shortwave ultraviolet illumination. It has a Mohs hardness of ∼2 and exhibits brittle tenacity with perfect cleavage on {010}, imperfect cleavage on {100}, and even fracture. The calculated density is 5.853 g/cm3 based on the empirical formula. Electron probe microanalysis provided the empirical formula Pb1.290U6+4.876U5+1.166O27H16.633 based on 27 O apfu and U5+:U6+ determined by X-ray photoelectron spectroscopy. Shinkolobweite is orthorhombic, superspace group Pnnm(0b0)000, a = 14.4808(4), b = 7.0681(8), c = 11.9423(3) Å, V = 1222.32(15) Å3, modulation wave vector [0 1/3 0], and Z = 2. The structure was refined from 8959 reflections to a final R1 = 0.0736 for all reflections. Uranyl oxide-hydroxide sheets in shinkolobweite adopt the β-U3O8 topology and possess (3 + 1) commensurately modulated ordering that results from the long-range ordering of U5+ and U6+ in the sheet, as well as the position and occupancy of interlayer Pb2+ cations. Observations of a topological transition between α-U3O8 and β-U3O8 type sheets in shinkolobweite supplements our understanding of U5+ mineral oxidation and stability.

Mineralogy↗
DOE OSTI2024

UO 2 solubility and chemical interactions in molten LiCl-Li 2 O

Here, this study investigated the solubility of UO 2 in LiCl-2 wt% Li 2 O at 650 °C with O 2 partial pressures up to 114 torr. Experiments were run in which UO 2 powder was contacted with the molten salt for up to 120 h. Salt samples were taken and analyzed for U and corrosion product concentration using inductively coupled plasma mass spectrometry and for Li 2 O concentration using titration. The highest measured U concentration in the salt was 0.06 wt%, and O 2 partial pressure was observed to have little to no effect on the solubility. Increasing O 2 partial pressure did increase the concentration of corrosion products in the salt. The concentration of Li 2 O in the molten salt progressively decreased with time in contact with UO 2 . This can be explained by the formation of Li 2 UO 4 via reaction of UO 2 with Li 2 O.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Evaluation of the Change in Uranium Mobility in Sediments from the Hanford 300-FF-5 Stage B Polyphosphate Field Injection

The purpose of this study was to evaluate the change in uranium mobility in the periodically re-wetted zone (PRZ) and upper unconfined aquifer of the Hanford Site’s 300-FF-5 operable unit as a result of Stage B polyphosphate (PO 4 ) injections. These 300 Area sediments are beneath or near the former North Process Pond, which has previously been shown to contain metatorbernite [Cu(UO 2 ) 2 (PO 4 ) 2 •8H 2 O] and uranophane [Ca(UO 2 ) 2 (SiO 3 OH) 2 •5H 2 O], along with aqueous and adsorbed uranium (U), and small amounts of U in calcite as a result of acidic disposal with a significant number of co-contaminants. Previous laboratory-scale studies have identified multiple mechanisms that decrease uranium mobility with polyphosphate injection: (a) precipitation of a low-solubility uranium-phosphate mineral (autunite); (b) precipitation of a Ca-phosphate solid that incorporates U; (c) precipitation of Ca-phosphate mineral (apatite group minerals) that coats existing uranium phases (i.e., U in calcite, U in Fe-oxides), reducing uranium leaching; and (d) aqueous Ca-U-carbonate species adsorption onto apatite. Given the significant amount of apatite precipitation, other U phases such as carbonates could also coprecipitate with apatite. To quantify differences in solubility, leachability, and phases of uranium between pre- and post-PO 4 -treated sediments, different types of analyses were conducted, including analysis of uranium in (a) a carbonate solution extraction from sediment over 1000 h (41.67 days), (b) six sequential liquid extractions from sediment over approximately 1 week, (c) leaching of uranium from sediments in one dimensional (1-D) column experiments over 2 months, and (d) solid phase characterization of select sediments. An additional extraction and solid phase measurements were conducted to evaluate phosphate precipitate formation. Metals were also analyzed in sequential liquid extractions as indicators of (a) amendment injection delivery, (b) co-contaminant movement, and (c) changes in solid phases. Aqueous and solid phase carbonate measurements were also conducted to evaluate changes in carbonates.

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

Uptake of uranium by carbonate crystallization from reduced and oxidized hydrothermal fluids

This work evaluated the immobilization of uranium (U) through incorporation into calcite under reduced and oxidized conditions. We investigated how much U could be entrapped by calcite crystallizing in chloride solutions in autoclaves at temperatures from 162 to 300 °C. The oxidation state of U was set by controlling oxygen fugacity via redox buffers. Uranium was introduced into calcite growth media as a solid oxide compound or U aliquot. We found the uptake of tetravalent U by calcite is higher than that of hexavalent U by up to four orders of magnitude. Furthermore, we estimate that crystallization of a few mg of calcite immobilizes all dissolved U when 1 kg of solution is saturated with UO 2 under reduced hydrothermal conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Laser-Induced Thermal Decomposition of Uranium Coordination Compounds with Non-oxidic Ligands to Produce Nitride and Carbide Materials

The production of ceramics from uranium coordination compounds can be achieved through thermal processing if an excess amount of the desired atoms (i.e., C or N), or reactive gaseous products (e.g., methane or nitrogen oxide) is made available to the reactive uranium metal core via decomposition/fragmentation of the surrounding ligand groups. Here, computational thermodynamic approaches were utilized to identify the temperatures necessary to produce uranium metal from some starting compounds—UI 4 (TMEDA) 2 , UCl 4 (TMEDA) 2 , UCl 3 (pyridine) x , and UI 3 (pyridine) 4 . Experimentally, precursors were irradiated by a laser under various gaseous environments (argon, nitrogen, and methane) creating extreme reaction conditions (i.e., fast heating, high temperature profile >2000 °C, and rapid cooling). Despite the fast dynamics associated with laser irradiation, the central uranium atom reacted with the thermal decomposition products of the ligands yielding uranium ceramics. Residual gas analysis identified vaporized products from the laser irradiation, and the final ceramic products were characterized by powder X-ray diffraction. The composition of the uranium precursor as well as the gaseous environment had a direct impact on the production of the final phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

A Rapid Sintering Method for Cerium Nitride Pellet: A Uranium Mononitride Surrogate

Uranium mononitride (UN) is a candidate fuel material for light water reactors with higher uranium (U) loading and thermal conductivity than uranium dioxide (UO 2 ). However, the sintering of UN pellets is challenging as the UN powder particles oxidize rapidly at high temperatures unless the oxygen concentration is extremely low. Oxidation during sintering either reduces the relative density of the sintered UN pellet or disintegrates the sintered UN pellet to powder. To address this problem, the present work developed a rapid sintering method for producing highly densified UN surrogate pellets with minimal oxidation. Cerium nitride (CeN) is used as a surrogate for UN to reduce radiation hazards. With the custom-developed fast-heating system, the sintering process was completed within 150 s. The sintering atmosphere was flowing nitrogen (N 2 ). The sintered CeN pellet density was 95% of the theoretical density (TD) or higher. The microstructure was uniform with a 10–25 µm grain size as demonstrated by scanning electron microscopy (SEM) and contained trivial levels of oxides as demonstrated by X-ray diffraction (XRD). The resultant pellets indicate that the rapid sintering method is a promising method to make UN fuel pellets with equivalent or higher density to pellets made by conventional sintering methods, while also being more efficient in time and costs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Tailoring Redox Active Ligands for Probing the Reactivity of Actinides

With this project, we aim to further enhance our understanding of fundamental f-element chemistry, including electronic structure and bonding behaviors. Key project goals include the characterization of actinide complexes bearing redox-active ligands, including those of lower-oxidation state uranium, and examining how structural changes to the ligand and coordination sphere affect structure and bonding. Redox-active and redox non-innocent ligands have been used widely in transition metal chemistry, but uranium complexes with redox-active ligands are surprisingly rare. We recently reported the synthesis of the redox-active ligand, “phen-BIAN” (N,N’-bis(iminophenol)acenaphthene), in an investigation of the electronic behavior of uranium and other actinides. This ligand framework borrows features from two classes of Schiff base ligands—the tetradentate O-N-N-O binding pocket from salens, and the redox-active α-diimine unit and backbone from Ar-BIANs (N,N’-bis[(aryl)imino]acenaphthenes, which can accept up to four electrons upon reduction. To continue this work, the electronics of uranyl and thorium complexes will be probed using the reduced forms of these ligands, using these in the preparation of lower-oxidation state and non-oxo uranium complexes. Additional members of the phen-BIAN family will be synthesized, including naphthol and thiol derivative. With these we can characterize the influence of substitutent groups on the redox activity, coordination geometry, and covalent interactions. The work proposed here entails extensive preparation, structural and spectroscopic characterization, of an array of actinide complexes. This will allow us to address probative questions about the nature of actinide bonding, the degree of covalency, the validity of lanthanides as models for the actinides, hard-soft interactions, magnetic interactions between metals in bimetallic complexes and what combination of electronic and steric affects produce an actinide selective ligand. The latter will assist greatly in providing a benchmark against which to compare and evaluate the distinct behaviors exhibited by 5f-element complexes. This will enable us to learn about the differences in binding between the 4f lanthanides and the 5f elements like uranium and plutonium. Such difference in binding can be exploited in separations and the development of new materials.

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

Uranium(VI) attenuation in a carbonate-bearing oxic alluvial aquifer

Uranium minerals are commonly found in soils and sediment across the United States at an average concentration of 2–4 mg/kg. Uranium occurs in the environment primarily in two forms, the oxidized, mostly soluble uranium(VI) form, or the reduced, sparingly soluble reduced uranium(IV) form. Here we describe subsurface geochemical conditions that result in low uranium concentrations in an alluvial aquifer with naturally occurring uranium in soils and sediments in the presence of complexing ligands under oxidizing conditions. Groundwater was saturated with respect to calcite and contained calcium (78–90 mg/L) with elevated levels of carbonate alkalinity (291–416 mg/L as HCO 3 -). X-ray adsorption near edge structure (XANES) spectroscopy identified that sediment-associated uranium was oxidized as a uranium(VI) form (85%). Calcite was the predominant mineral by mass in the ultrafine fraction in uranium-bearing sediments (>16 mg/kg). Furthermore, groundwater geochemical modeling indicated calcite and/or a calcium-uranyl-carbonate mineral such as liebigite in equilibrium with groundwater. The δ 13 C (0.57‰ ± 0.15‰) was indicative of abiotic carbonate deposition. Thus, solid-phase uranium(VI) associated with carbonate is likely maintaining uranium(VI) groundwater levels below the maximum contaminant level (MCL; 30 µg/L), presenting a deposition mechanism for uranium attenuation rather than solely a means of mobilization.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Trace Actinide Signatures of a Bulk Neptunium Sample

In nuclear forensic analyses, measurements of actinide elements in a sample can assist with identifying interdicted or unknown materials. While these radiochemical signatures have been extensively investigated in uranium materials, less is known about bulk neptunium samples. This paper describes the measurement of trace actinide concentrations and isotopic profiles in a 237 Np oxide sample. Uranium, plutonium, americium, and curium concentrations and isotopic profiles in the sample were determined and deemed potentially useful for distinguishing different sources of 237 Np. Furthermore, several different potential radiochronometry systems were also investigated; discordant results indicate that the Np sample was never completely purified of other actinide elements, or that subsequent contamination of the sample occurred. Few prior studies of neptunium materials have been reported, and these data suggest that trace actinide constituents could provide unique signatures to identify material out of regulatory control.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Thermal Process Technology for Nuclear Applications - 20379

ANSTO's Synroc technology has been developed to provide a safe, secure matrix for the immobilization and final disposal of radioactive waste. Synroc technology will be used to manage radioactive wastes from the production of the radioisotope Molybdenum-99 (Mo-99). This paper shall outline various stages of the process development with specific reference to the thermal treatment technology of calcination. Calcination is a key step in the Synroc process [1-2].The rotary thermal processing system includes: an advanced heating element design for increased robustness and ease of remote operation and maintenance, an enhanced modular design of components for ease of remote maintenance in a hot cell and in compliance with hot cell radioactive environment requirements for safety, reliability and maintainability. In addition to thermal treatment of waste from nuclear medicine production, this technology provides solutions for a variety of nuclear materials processing applications including sintering UO{sub 2} pellets for reactor fuel rods, oxidation of UO{sub 2} pellets, swarf, and powder to U{sub 3}O{sub 8}, de-nitration of Uranyl nitrate and hydrofluorination of UO{sub 2} pellets. The paper will also discuss thermal processing solutions for a range of nuclear applications. (authors)

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

Assembly of Uranyl Peroxides from Ball Milled Solids

Mechanochemistry enables transformations of highly insoluble materials such as uranium dioxide or the mineral studtite [(UO 2 )(O 2 )(H 2 O) 2 ]·(H 2 O) 2 into uranyl triperoxide compounds that can subsequently assemble into hydroxide-bridged uranyl peroxide dimers in the presence of lithium hydroxide. Dissolution of these solids in water yields uranyl peroxide nanoclusters including U 24 , Li 24 [(UO 2 )(O 2 )(OH)] 24 . Insoluble uranium solids can transform into highly soluble uranyl peroxide phases in the solid state with miniscule quantities of water. Furthermore, such reactions are potentially applicable to uranium processing in the front and back end of the nuclear fuel cycle.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

When Every Drum is a Win - Tackling a Thirty-Five Thousand Drum Legacy - 20064

Cameco Corporation's (Cameco) Port Hope Conversion Facility (PHCF) was previously owned by the federal Crown Corporation, Eldorado Nuclear, which held a significant inventory of legacy waste material at the time Cameco was formed in 1988. As a result, Cameco was granted an allocation of 150,000 cubic metres of space in the Long-Term Waste Management Facility (LTWMF) located in the Municipality of Port Hope (MPH), which opened to the receipt of Cameco material in June 2018. Cameco is currently undertaking a major site cleanup and renewal of its Port Hope Conversion Facility (PHCF) which is known as the Vision in Motion (VIM) project. Over its operating history, Eldorado accumulated an inventory of over 35,000 drums of accumulated waste that was primarily stored at two offside warehouse locations. In 2017 and 2018, the project focused on repackaging two well characterized legacy wastes (magnesium fluoride slag and depleted uranium titanium oxide) from one of the two offsite locations. At the conclusion of this work, approximately 15,000 drums of these wastes had been repackaged and disposed of at the LTWMF. There remained approximately 1000 drums from this location that either could not be repackaged safely or were different waste types and required further verification and/or processing. The buildings at this location were scheduled for demolition in early 2019, which triggered the development of a process for triaging legacy drums in August 2018, which allowed for preparation and shipment of approximately half of these drums to the LTWMF by March 2019, with the remainder moved to the second offsite warehouse location in February 2019 for further characterization which is ongoing. This paper will discuss key lessons learned as the drummed legacy waste disposal inventory has been reduced to approximately half of the initial inventory. This will include the development of an alternate packaging process; prioritization of characterization activities and selection of techniques where minimal inventory information is available; determination of next step(s) for each drum as it is assessed; key safety considerations; and how to make inroads into an overwhelming task while under public and regulatory scrutiny. In a relatively short period of time, significant progress has been made to organize and gather information about the legacy waste, update the inventory records and determine the most appropriate pathways (i.e. LTWMF disposal, disposal at another appropriate facility, site storage until future processing and/or disposal). Since 2017 the volume of the decades-old legacy waste inventory has decreased significantly. With the majority of the known materials safely disposed of at the LTWMF, every drum removed from the endless rows of 20,000 pyramidally-stacked waste drums with limited history that is safely characterized, prepared and shipped is considered a win. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗