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DOE OSTI2023

Upcycling and urban mining for nanomaterial synthesis

Responsible processing of complex waste streams, especially through resource recovery and upgrading, is essential to lessen the environmental and societal impacts of materials production and disposal. Upcycling waste into materials that have higher value than their original form, and urban mining, the recovery of metals from electronic waste, are two recent strategies in waste management. Sustainable cradle-to-grave manufacturing means recovering resources that would otherwise contaminate the planet. Here, we discuss upcycling of plastic, rubber, and related carbon waste streams into graphene, as well as the recovery of precious metals and rare earth elements, while removing toxic heavy metals, all using flash Joule heating. Furthermore, this article provides a brief overview of the current state of upcycling and urban mining as it applies to materials conversion, recovery, or nanomaterial synthesis, with a concluding perspective highlighting the considerations, opportunities, and challenges that accompany upcycling and urban mining research.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2020

Pitch Variation Experiments in Water-Moderated Square-Pitched U(6.90)O 2 Fuel Rod Lattices with Fuel to Water Volume Ratios Spanning 0.08 to 0.67

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

RCT Continuing Training (Q2 2021) [Slides]

This quarter’s training assignment is split into two sections. Section one will review the fundamental elements of removing items from controlled areas. This section discusses each decision point and the required actions to determine if an item is suitable for release. Section two briefly reviews Area Radiation Monitors, commonly referred to as ‘ARMs’, and discusses the required actions to take in the event of an ARM alarm.” Report includes slides and transcript.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2023

Molybdenum Sleeve Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.55 cm Pitch)

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM-096, LEUCOMP-THERM-097, LEU-COMP-THERM-101, and LEU-COMP-THERM-102. The purpose of these experiments was to measure the effects of molybdenum in nearly-critical systems. The molybdenum was introduced into the fuel arrays as tubular sleeves that surrounded some of the fuel rods in the fuel arrays measured. Four hundred molybdenum tubes nominally 12.7 mm outside diameter, 498 mm long, with 0.762 mm wall thickness were provided for the experiments by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). Small polyethylene adapters at each end of the tubes were used to center each tube on a fuel rod in the assembly. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 1261 fuel rod positions on a pitch of 0.610 in (1.5494 cm) in a series of 20 hexagonal rings surrounding the central fuel rod. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The five critical experiments in this series were performed in August through December 2022, in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility. Case 1 had no molybdenum sleeves, Case 2 had 208 molybdenum sleeves clustered at the center of the array, Case 3 had 397 molybdenum sleeves clustered at the center of the array, Case 4 had 175 molybdenum sleeves in the central position and in five alternating hexagonal rings, and Case 5 had 331 molybdenum sleeves in the central position and in seven alternating hexagonal rings. All five critical experiments are judged to be acceptable as benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Tantalum Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.02 cm Pitch)

The United States (US) Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (SNL). The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water moderated and — reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM 096, LEU- COMP-THERM-097, LEU-COMP-THERM-101, LEU-COMP-THERM-102, and LEU-COMP-THERM-111. Applied interests in neutron-absorbing elements at epithermal and intermediate neutron energy ranges at DOE’s Hanford Site and Savannah River Site were identified through DOE Office of Environmental Management (EM) workshops and needs surveys performed in conjunction with the National Nuclear Security Administration/Nuclear Criticality Safety Program. In response, a team lead by Oak Ridge National Laboratory (ORNL) performed a study evaluating the use of 7uPCX for testing the epithermal cross sections of materials at SNL. This marked the inception of the set of experiments described here. The purpose of these experiments was to measure the effects of tantalum in nearly-critical systems. The tantalum was introduced into the fuel arrays as experiment rods within a central test region. The central test region was designed to target the epithermal neutron energy range by providing a dry cylindrical cavity constructed of aluminum and lined with 0.04 in (0.1016 cm) thick cadmium sheet. The central test region has an outer diameter of 3.75 in (9.525 cm) length of 31 in (78.74 cm) and can hold 85 tantalum rods nominally 0.25 in (0.635 cm) outside diameter and 31.25 in (79.375 cm) long. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 2016 fuel rod positions on a pitch of 0.4 in (1.016 cm) in a series of 21 hexagonal rings positioned around the central test region. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The eight cases in this experiment series were performed in 2024 in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility by varying the number of fuel rods in the critical assembly. The personnel involved in conducting these experiments included David Ames, Elijah Lutz, Beth Hanson, Patrick Ward, Jason Soares, and Augie Chapa from SNL, as well as Mathieu Dupont and B.J. Marshall from ORNL. Case 1 had no tantalum experiment rods. Case 2 had seven tantalum rods in the center of the central test region. Case 3 had 18 tantalum rods in the third hexagonal ring of the central test region. Case 4 had 19 tantalum rods in the center of the central test region. Case 5 had 30 tantalum rods in the outer most positions of the central test region. Case 6 had 37 tantalum rods in the center of the central test region. Case 7 had 61 tantalum rods in the center of the central test region. Case 8 had 85 tantalum rods filling all positions in the central test region. All eight critical experiments are judged to be acceptable as benchmark experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Method for separation of rare earth elements from coal ash using supercritical carbon dioxide

Embodiments of the present disclosure generally relate to the recovery and extraction of rare earth elements. More specifically, embodiments of the disclosure relate to methods for separating rare earth elements from coal, coal by-product(s), and/or coal-derived product(s). In an embodiment, a method of removing rare earth elements from a coal-derived product is provided. The method generally includes introducing supercritical CO 2 to the coal ash to form a first mixture, introducing a first acid to the first mixture to form a second mixture, and removing a first composition from the second mixture, the first composition comprising the one or more rare earth elements.

Fan, Maohong↗

DOE OSTI2022

Mercury Removal from Contaminated Water by Wood-Based Biochar Depends on Natural Organic Matter and Ionic Composition

Biochars can remove potentially toxic elements, such as inorganic mercury [Hg(II)] from contaminated waters. However, their performance in complex water matrices is rarely investigated, and the combined roles of natural organic matter (NOM) and ionic composition in the removal of Hg(II) by biochar remain unclear. Here, we investigate the influence of NOM and major ions such as chloride (Cl – ), nitrate (NO 3 – ), calcium (Ca 2+ ), and sodium (Na + ) on Hg(II) removal by a wood-based biochar (SWP700). Multiple sorption sites containing sulfur (S) were located within the porous SWP700. In the absence of NOM, Hg(II) removal was driven by these sites. Ca 2+ bridging was important in enhancing removal of negatively charged Hg(II)-chloro complexes. In the presence of NOM, formation of soluble Hg-NOM complexes (as seen from speciation calculations), which have limited access to biochar pores, suppressed Hg(II) removal, but Cl – and Ca 2+ could still facilitate it. The ability of Ca 2+ to aggregate NOM, including Hg-NOM complexes, promoted Hg(II) removal from the dissolved fraction (<0.45 μm). Hg(II) removal in the presence of Cl – followed a stepwise mechanism. Weakly bound oxygen functional groups in NOM were outcompeted by Cl – , forming smaller-sized Hg(II)-chloro complexes, which could access additional intraparticle sorption sites. Therein, Cl – was outcompeted by S, which finally immobilized Hg(II) in SWP700 as confirmed by extended X-ray absorption fine structure spectroscopy. We conclude that in NOM containing oxic waters, with relatively high molar ratios of Cl – : NOM and Ca 2+ : NOM, Hg(II) removal can still be effective with SWP700.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2021

Time dependent zero valent iron oxidation and the reductive removal of pertechnetate at variable pH

Elemental iron Fe 0 is a promising reductant for removal of radioactive technetium-99 (Tc) from complex aqueous waste streams that contain sulfate, halides, and other inorganic anions generated during processing of legacy radioactive waste. The impact of sulfate on the kinetics of oxidation and reduction capacity of Fe 0 in the presence of Tc has not been examined. We investigated the oxidative transformation of Fe 0 and reductive removal of TcO 4 - in 0.1 M Na 2 SO 4 as a function of initial pH (i.e., pH i 4, 7, and 10) under aerobic conditions up to 30 days. Tc reduction was the fastest at pH i 7 and slowest at pHi 10 (Tc reduction rate pHi 7 > 4 > 10). Aqueous fraction of Tc was measured at 0.4% at pH i 7 within 6 h, whereas ≥ 97% of Tc was removed from solutions at pH i of 4 and 10 within 24 h. Solid phase characterization showed that magnetite was the only oxidized crystalline phase for the first 6 h regardless of initial pH. Lepidocrocite was the most abundant oxidized product for pH i 10 after 5 days, but was not observed at pH of 4 or 7.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Evaluating and Enhancing Iron Removal via Filterable Iron Precipitates Formation during Coal-Waste Bioleaching

Iron removal via jarosite precipitate formation is a commonly used technique in various hydrometallurgical processes. Excess iron removal often becomes essential to an overall metal recovery circuit. This is particularly important to processes involving iron-bearing minerals. A technique, which involved the use of pyrite to generate acid for leaching, for iron removal is critical to enabling the process. Iron removal using CaO or similar reagents is expensive and often results in lost product. In the present study, various compounds that facilitate jarosite formation, namely Na2SO4, NH4OH, KCl, and KOH, were utilized and their effect in precipitation was observed. Visual Minteq assisted simulations were run in order to evaluate favorable conditions for iron removal. Morphology and elemental composition of precipitates were analyzed using scanning electron microscopy equipped with energy-dispersive X-ray spectroscopy, and the phase purity was identified using X-ray diffraction analysis.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Non–linear bonding trends in maleonitrile-1,2–dithiolate complexes of the transuranium actinides

The trivalent actinides are produced in the nuclear fuel cycle during power production and provide the largest long-term radiation dose in used nuclear fuel. It is ideal for these elements to be removed from used nuclear fuel for disposal and a necessity for fuel recycling. A key challenge to this is the similarity of chemical behavior of the trivalent actinides to the lanthanides that are also present as fission products in used fuel. Thus far, some of the most effective separations of actinides from lanthanides utilise chelating agents containing sulfur moieties such as dithiophosphinates that selectively bind to actinide ions because of a greater bond covalency relative to lanthanide ions. Typically, greater differences between actinide and lanthanide ions are observable the more ligands and chelators bonds have a covalent character. Here, a series of complexes of the trivalent actinides Np(III) through Cf(III) (excluding Bk(III)) with maleonitrile-1,2-dithiolate (mnt 2– ) are synthesized along with their lanthanide counterparts (La(III) – Nd(III), Sm(III) – Gd(III), Dy(III)), in order to characterize the nature of chemical bonds with these metal ions and a polarizable, non-innocent, sulfur-donor ligand. The metal-sulfur bonds in these complexes trend shorter than measured for lanthanides with equivalent ionic radii. However, particularly large deviations are observed in the neptunium and plutonium complexes in both structure and bonding, resulting in a nonlinear bond length trendline for the actinide series. Density Functional Theory (DFT) calculations with Quantum Theory of Atoms in Molecules (QTAIM) and Natural Bond Order (NBO) analyses indicate that for the neptunium and plutonium complexes, the presence of increased 5f-orbital participation, energy degeneracy of the metal and ligand orbitals, and the structure packing result in shortened M–S bonds. The stabilization of the energy of the 5f-orbitals and the decrease in f-contribution to bonding orbitals in the later actinides results in structural properties more similar to the lanthanide complexes.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Effect of geometry and operational parameters on the dehumidification performance of a desiccant coated heat exchanger

Solid desiccant dehumidification systems are an alternative to vapor-compression-based air dehumidification systems. They use solid desiccant materials to adsorb air moisture in space cooling. DCHE can integrate the cooling component in the heat exchanger to achieve higher dehumidification capacity. Comprehensive numeric modeling would be necessary to assist in the design and operation of the DCHE under development to achieve maximum performance. Here, we provide the details of a heat and mass transfer model developed for this purpose. The model aims to cover a gap in existing models by independently modeling heat transfer on the solid desiccant, fin, and tube. The model results were compared with an experimental reference for validation. The air outlet humidity and temperature results for dehumidification and regeneration showed a deviation lower than 15% from the experiment. The validated model was used to perform a parametric study for a series of design and operating conditions. The parametric analysis showed that an increase of 25% in desiccant layer thickness and thermal contact resistance between solid elements reduces moisture removal by 9.1% and 1%, respectively. These results indicate the need to independently model the desiccant layer.

42 ENGINEERING↗

DOE OSTI2024

Correlation of Fabrication Methods and Enhanced Wear Performance in Nanoporous Anodic Aluminum Oxide with Incorporated Molybdenum Disulfide (MoS 2 ) Nanomaterials

Wear performance is integral to component longevity, minimizing industrial waste and excess energy costs in a wide variety of applications. Anodized aluminum oxide (AAO) has many beneficial properties leading to its wide use across industries as a surface treatment for many aluminum components, but the wear properties of the coating could be improved significantly. Here, we used an electrochemical method to incorporate molybdenum disulfide (MoS 2 ), a nanomaterial used as a dry lubricant, to modify alloys of aluminum during AAO preparation. Using Raman spectroscopy and tribological scratch measurements, we thoroughly characterized the structure and wear behavior of the films. The MoS 2 deposition procedure was optimal on aluminum 5052 anodized in higher acid concentrations, with friction coefficients at around 0.05 (~10× better than unmodified AAO). Changing anodization conditions to produce harder films with smaller pores led to worsened wear properties, likely because of lower MoS 2 content. Studying a commercial MoS 2 /AAO film of a different Al alloy (7075) showed that a heat treatment step intended to fully convert all deposited MoS x species to MoS 2 can adversely affect wear in some alloys. While Al 6061 and 1100 produced films with worse wear performance compared to Al 5052 or 7075, our results show evidence that acid cleaning after initial anodization likely removes residual alloying elements, affecting MoS 2 incorporation. This study demonstrates a nanomaterial modified AAO film with superior wear characteristics to unmodified AAO and relates fabrication procedure, film structure, and practical performance.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Process Development of the Vaporizing Foil Actuator Welding Technique

The industrial focus on continuous improvement for products is leading design engineers to consider multi-material design concepts more frequently. The design concepts use the advantages of each material while simultaneously minimizing the drawbacks. It also allows for the overall weight of the products to be reduced. Some traditional joining methods, though, struggle to join dissimilar material combinations. A technique known as vaporizing foil actuator welding, which was developed around ten years ago, is a solid-state joining method that has the potential to overcome the barriers to dissimilar material joining. The biggest drawbacks of vaporizing foil actuator welding are that it is still exclusively used in laboratory settings and early in the technology development process. As such, there is not enough confidence in the process for it to be transitioned to manufacturing environments yet. This work begins by analyzing the current state of the technology and identifying a roadmap to achieve a transition to manufacturing. The analysis found that increased confidence in the technology requires sample to sample repeatability to be improved. Beyond that, the process also needs to be fully automated before it can be considered as a viable technique for mass production. In order to further develop the technique, a fully automated work cell was constructed. By removing the human element from the process, this cell helped identify iii the aspects which contribute to process variability. The positioning of the vaporizing foil actuator was identified as the most important factor in process stability. A small set of samples were made in the automated cell with the addition of one component epoxy adhesive around the perimeter of the weld. These samples showed improved process repeatability compared with the samples made without adhesive. This indicates that small amounts of adhesive may be critical in using the process in industrial settings. Additionally, work was performed with manual positioning and adhesive dispensing. This work found that a hybrid joint behaves like traditional weld bonding, where the addition of adhesive makes the joint stronger than if either of the two joining methods were used on their own. To realize this performance with vaporizing foil actuator welding, though, the adhesive needs to be protected from burning during the rapid compression of air during the impact process.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Bioleaching to produce clean loblolly pine for thermochemical conversion

Lignocellulosic biomass contains inorganic elements that could induce slagging, ash fouling, and emission of both corrosive and harmful gases during thermochemical conversion. Water leaching pretreatment can remove most of the water-soluble elements, while it has limited industrial applicability due to its ineffectiveness in removing water-insoluble elements. Bioleaching is an alternative pretreatment method which has not been fully studied. In this work, bioleaching by Aspergillus niger strains NRRL 2001, NRRL 3122, and NRRL 567 was conducted to pretreat loblolly pine biomass with added water. Further, the removal of inorganic elements (K, Ca, Mg, and S) by bioleaching with A. niger NRRL 2001 was compared with chemical leaching with citric acid, HCl, and NaOH. It was shown that the pH reduction during bioleaching greatly improved K and Mg leaching to the point comparable to acid leaching, whereas Ca was precipitated by oxalic acid produced by the fungus. Bioleaching was also conducted in manually separated needles and chips components of loblolly pine. Glucose was added at different levels to assist bioleaching. At higher glucose levels, gluconic acid and citric acid were produced during bioleaching, leading to the increase of K, Ca, and Mg removal rates by more than 35%, 180%, and 390%, respectively, from the feedstock as compared with water leaching. Overall, bioleaching greatly improved removal of K, Ca, Mg, but not S from loblolly pine feedstocks as compared with water leaching.

09 BIOMASS FUELS↗

DOE OSTI2023

Conclusions from 3 years of continuous capture plant operation without exchange of the AMP/PZ-based solvent at Niederaussem – insights into solvent degradation management

A many times heard mantra of solvent degradation management in amine-based post combustion capture is “keep the solvent clean” to minimize solvent consumption. It is assumed that the amine losses would decrease by the removal of metals, degradation products, and reactive trace components which are captured from the flue gas, like NO 2 (as potentially driving components of the amine degradation besides dissolved O 2 ). However, this theoretical hypothesis – based on results from laboratory experiments typically generated with fresh amines – disregards the complexity of the solvent matrix, interaction of potential metal catalysts with degradation products and oxidizing agents, and specific chemical requirements which must be fulfilled before a degradation mechanism can proceed. Degradation of the solvent CESAR1 (aqueous solution of 3.0 molar 2-amino-2-methylpropan-1-ol (AMP) and 1.5 molar piperazine (PZ)) is investigated in a unique long-time test campaign (testing time up to now 40 months; 24/7 operation) without replacement of the solvent inventory at the capture pilot plant at the lignite-fired power plant in Niederaussem. Three solvent management strategies with different effect mechanisms are investigated and evaluated: (a) removal of only anionic compounds and trace elements (within 75 days solvent inventory treated two times) and anionic as well as cationic compounds and trace elements (114 days, inventory treated four times) from the solvent by ion exchange, (b) adsorptive removal of trace elements from the solvent by active carbon in 35% of the operating time, and (c) removal of >80% NO 2 by flue gas pretreatment with thiosulfate/sulfite solution (dosing for 2,000 h). The results of the testing program clearly show that “solvent cleanliness” is not a well-defined parameter and that results from laboratory tests, tests without fully representative industrial flue gasses, and short-term testing of monoethanolamine cannot be generalized for other solvents and industrial application. Furthermore, these results showcase that specific degradation management considering solvent, capture plant and flue gas quality is reasonable. Overshooting efforts for solvent management are contra-productive and produce unnecessary waste streams, efficiency losses and costs.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Removal of Bond Sodium from Full-Length Unirradiated Fermi-1 Blanket Elements and Assembly via Melt-Drain-Evaporate Process

Equipment was designed, fabricated, tested, and operated in an inert atmosphere radiological glovebox at Idaho National Laboratory to demonstrate the removal of bond sodium from full-length unirradiated Fermi-1 radial blanket elements and an entire radial blanket assembly using a Melt-Drain-Evaporate process. A series of three runs was performed with individual and multiple radial Fermi-1 blanket elements, and a fourth run was conducted with an entire Fermi-1 radial blanket assembly. After each run, the depleted uranium alloy slugs in every element slid out of its cladding, mechanically exhibiting the effectual absence of bond sodium. Quantitative and further qualitative analyses of the treated Fermi-1 material revealed the substantive, if not complete, absence of sodium metal in blanket element components. Indeed, entire columns of depleted uranium alloy slugs and associated cladding from select treated elements were separately contacted with water to react with residual sodium metal on the element surfaces, forming hydrogen gas. Samples of the resultant gas were analyzed to quantify the hydrogen concentration, which correlated to a residual sodium metal content for the treated element. Accordingly, no detectable sodium metal (<7 µg) was found on the surfaces of multiple depleted uranium alloy slugs after removal from its cladding. Detectable sodium metal, ranging from 14 to 30 µg, was found on the surfaces of one separated column of slugs and two separated full-length cladding segments. Each element originally contained ~25 g of bond sodium, resulting in quantitative sodium metal removal efficiencies of =99.9998%. The remaining separated uranium alloy slugs and cladding segments from the four runs were piece-wise immersed in alcohol while videorecording possible gas bubble formation as an additional qualitative analysis for residual sodium metal on treated surfaces, which identified the predominant absence of sodium reactivity on the balance of treated Fermi-1 blanket material. The results of this demonstration substantiate a path forward for treatment and disposal of 34 metric tons heavy metal of irradiated sodium-bonded Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt

Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Sequential removal of cation/H + exchangers reveals their additive role in elemental distribution, calcium depletion and anoxia tolerance

Multiple Arabidopsis H + /Cation exchangers (CAXs) participate in high-capacity transport into the vacuole. Previous studies have analysed single and double mutants that marginally reduced transport; however, assessing phenotypes caused by transport loss has proven enigmatic. Here, in this study, we generated quadruple mutants (cax1-4: qKO) that exhibited growth inhibition, an 85% reduction in tonoplast-localised H + /Ca transport, and enhanced tolerance to anoxic conditions compared to CAX1 mutants. Leveraging inductively coupled plasma mass spectrometry (ICP-MS) and synchrotron X-ray fluorescence (SXRF), we demonstrate CAX transporters work together to regulate leaf elemental content: ICP-MS analysis showed that the elemental concentrations in leaves strongly correlated with the number of CAX mutations; SXRF imaging showed changes in element partitioning not present in single CAX mutants and qKO had a 40% reduction in calcium (Ca) abundance. Reduced endogenous Ca may promote anoxia tolerance; wild-type plants grown in Ca-limited conditions were anoxia tolerant. Sequential reduction of CAXs increased mRNA expression and protein abundance changes associated with reactive oxygen species and stress signalling pathways. Multiple CAXs participate in postanoxia recovery as their concerted removal heightened changes in postanoxia Ca signalling. This work showcases the integrated and diverse function of H + /Cation transporters and demonstrates the ability to improve anoxia tolerance through diminishing endogenous Ca levels.

36 MATERIALS SCIENCE↗