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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Communication—Dissolution and Passivation of a Ni-Cr-Fe-Ru-Mo-W High Entropy Alloy by Elementally Resolved Electrochemistry

The formation and dissolution of the passive film on a novel NiCrFeRuMoW high entropy alloy in H 2 SO 4 was measured with atomic emission spectroelectrochemistry. Passivation involved Cr and Ru enrichment, whether passivation was achieved spontaneously at open circuit or by potentiostatic step into the passive domain. The other elements, Ni, Fe, Mo, and W, dissolved at rates approximately proportional to their bulk concentrations. The air formed film showed only a Cr enrichment. Excellent corrosion resistance properties were indicated by spontaneous passivation, the low passive current in the polarization curve, and the absence of detectable elemental dissolution during the active peak.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Mitigating Transition Metal Dissolution from Mn-rich Cathodes: Influence of Processing and Testing Methods

Manganese-rich oxides continue to gain interest with respect to the development of Earth-abundant options for lithium-ion cathodes. Of the unique challenges that hinder the respective performance of various classes of such materials, manganese dissolution still stands as a common theme. The work herein explores Li 3 PO 4 as a robust surface protection layer on a prototypical, Co-free, manganese-rich cathode in the way of a lithium- and manganese-rich oxide. The study highlights the critical importance of synthesis and processing in realizing optimal performance of a given surface treatment by comparing sol-gel and atomic layer deposition methods. Furthermore, cycling protocols are emphasized as a critical factor in adequately gauging the efficacy of surface protection strategies to mitigate manganese dissolution and the subsequent electrochemical consequence. Optimized Li 3 PO 4 coatings coatings on lithium- and manganese-rich cathode particles are shown to greatly mitigate capacity fade, impedance rise, pore/void formation and mechanical damage during long-term cycling.

Mallick, Subhadip [Argonne National Laboratory (AN↗
DOE OSTI2017

Transition metal dissolution, ion migration, electrocatalytic reduction and capacity loss in Lithium-ion full cells

Continuous operation of full cells with layered transition metal (TM) oxide positive electrodes (NCM523) leads to dissolution of TM ions and their migration and incorporation into the solid electrolyte interphase (SEI) of the graphite-based negative electrode. These processes correlate with cell capacity fade and accelerate markedly as the upper cutoff voltage (UCV) exceeds 4.30 V. At voltages ≥ 4.4 V there is enhanced fracture of the oxide during cycling that creates new surfaces and causes increased solvent oxidation and TM dissolution. Despite this deterioration, cell capacity fade still mainly results from lithium loss in the negative electrode SEI. Among TMs, Mn content in the SEI shows a better correlation with cell capacity loss than Co and Ni contents. As Mn ions become incorporated into the SEI, the kinetics of lithium trapping change from power to linear at the higher UCVs, indicating a large effect of these ions on SEI growth and implicating (electro)catalytic reactions. Lastly, we estimate that each Mn II ion deposited in the SEI causes trapping of ~10 2 additional Li + ions thereby hastening the depletion of cyclable lithium ions. Using these results, we sketch a mechanism for cell capacity fade, emphasizing the conceptual picture over the chemical detail.

25 ENERGY STORAGE↗
DOE OSTI2021

Synthesis, antibiotic structure–activity relationships, and cellulose dissolution studies of new room-temperature ionic liquids derived from lignin

Ionic liquids (ILs) are promising pretreatment solvents for lignocellulosic biomass, but are largely prepared from petroleum precursors. Benzaldehydes from depolymerized lignin, such as vanillin, syringaldehyde, and 4-methoxy benzaldehyde, represent renewable feedstocks for the synthesis of ionic liquids. We herein report syntheses of novel lignin-derived ionic liquids, with extended N -alkyl chains, and examine their melting points, cellulose dissolution capacities, and toxicity profiles against Daphnia magna and E. coli strain 1A1. The latter organism has been engineered to produce isoprenol, a drop-in biofuel and precursor for commodity chemicals. The new N,N -diethyl and N,N -dipropyl methyl benzylammonium ILs were liquids at room temperature, showing 75–100 °C decreased melting points as compared to their N,N,N -trimethyl benzylammonium analog. Extension of N -alkyl chains also increased antibacterial activity threefold, while ionic liquids prepared from vanillin showed 2- to 4-fold lower toxicity as compared to those prepared from syringaldehyde and 4-methoxybenzaldehyde. The trend of antibacterial activity for anions of lignin-derived ILs was found to be methanesulfonate < acetate < hydroxide. Microcrystalline cellulose dissolution, from 2 to 4 wt% after 20 min at 100 °C, was observed in all new ILs using light microscopy and IR spectroscopy. Ionic liquids prepared from H-, S- and G-lignin oxidation products provided differential cytotoxic activity against E. coli and D. magna , suggesting these compounds could be tailored for application specificity within a biorefinery.

09 BIOMASS FUELS↗
DOE OSTI2020

The dissolution of simulant vitrified intermediate level nuclear waste in young cement water

Abstract It is pertinent to the safety case for geological disposal in the UK that the behaviour of vitrified wastes in proximity to cementitious materials is understood. In this study, vitrified simulant intermediate level nuclear waste (ILW) was subject to dissolution in a synthetic cement water solution to simulate disposal conditions. Results show that the presence of alkali / alkaline earth elements in the cementitious solution can be favourable, at least in the short-term, leading to lower dissolution rates associated with incorporation of these elements into the altered layer of the glass.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Uranium Anodic Dissolution under Slightly Alkaline Conditions Progress Report Full-Scale Demonstration with DU Foil

Argonne National Laboratory (Argonne) is developing an alternative method for digesting irradiated low enriched uranium (LEU) foil targets to produce 99 Mo in neutral/alkaline media. This method consists of the electrolytic dissolution of irradiated uranium foil in sodium bicarbonate solution, followed by precipitation of base-insoluble fission and activation products, and uranyl-carbonate species with CaO. The addition of CaO is vital for the effective anion exchange separation of 99 MoO 4 2- from the fission products, since most of the interfering anions (e.g., CO 3 2- ) are removed from the solution, while molybdate remains in solution. An anion exchange is used to retain and to purify the 99 Mo from the filtrate. The electrochemical dissolver has been designed and fabricated in 304 stainless-steel (SS), and tested for the dissolution of a full-size depleted uranium (DU) target, wrapped in Al foil. Future work will include testing with low-burn-up DU foil at Argonne and later with high-burn-up LEU foils at Oak Ridge National Laboratory.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Full-Scale Testing of the Ambient Pressure, Acid-Dissolution Front-End Process for the Current Mo-99 Recovery Processes

The Global Threat Reduction Initiative (GTRI) Conversion Program is actively developing technologies for the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The conversion of conventional HEU targets to LEU for 99 Mo production requires approximately five times the uranium in a target to maintain the 99 Mo yield on a per target basis. Under GTRI, Argonne is developing two front-end options for current 99 Mo production processes to allow the use of LEU-foil targets. In both processes, the aim is to produce a front-end product that is compatible with current 99 Mo purification operations and will provide the same or higher yield of 99 Mo for the same number of irradiated targets. The two front-end processes under development as part of this project are (1) the dissolution of irradiated LEU foil (up to 250 g in a single batch) and nickel fission recoil barrier in nitric acid at ambient pressure and (2) the electrochemical dissolution of LEU foil in a series of steps that produces an alkaline (basic) solution feed for 99 Mo purification. This report describes the results of performance tests and design optimization of the ambient pressure, nitric acid dissolver system. The design, fabrication, and performance test planning for this system are described in more detail in previous reports (Jerden et al. 2011a,b, 2012). Full-scale demonstrations of both of the front-end processes using irradiated uranium foils are planned to be performed at Oak Ridge National Laboratory this fiscal year.

Jerden, James↗
DOE OSTI2012

Design, Fabrication, and Testing of an Acid-Dissolution Front-End Process for Current Mo-99 Recovery Processes: Optimization for Full-Scale Demonstration in Hot-Cell Facility

The National Nuclear Security Administration’s Global Threat Reduction Initiative Conversion Program (GTRI) develops technology necessary for converting civilian nuclear facilities that use high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The conversion of conventional HEU targets to LEU for the production of Mo-99 for medical use requires approximately five times the amount of uranium per target to maintain current Mo-99 yields. Under GTRI, Argonne National Laboratory (Argonne) is developing two frontend options for the current Mo-99 recovery processes to allow the use of LEU-foil targets: (1) an ambient pressure, nitric acid dissolution process; and (2) an electrochemical, basic dissolution process. The goal in developing both frontend options is to produce a product that will be compatible with current purification operations and that will, with the same number of targets irradiated, provide the same or a higher yield of Mo-99 at the end of processing. This work was done under the Mo-99 International Production Support, FY 2012 work package: 21.2.94.2.6 LEU Target – Foreign Production Development, ANL Activity 1.

Jerden, James↗
DOE OSTI2013

Full-Scale Testing of the Ambient Pressure, Acid-Dissolution Front-End Process for the Current 99 Mo Recovery Processes

The Global Threat Reduction Initiative (GTRI) Conversion Program is actively developing technologies for converting civilian facilities that use high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The conversion of conventional HEU targets to LEU for the production of 99 Mo production requires approximately five times the uranium in a target to maintain the 99 Mo yield on a per-target basis. Under GTRI, Argonne National Laboratory (Argonne) is developing two frontend options for current 99 Mo production processes to allow the use of LEU-foil targets. In both processes, the aim is to produce a frontend product that is compatible with current 99 Mo purification operations and will provide the same or a higher yield of 99 Mo for the same number of irradiated targets. The two frontend processes under development as part of this project are (1) the dissolution of irradiated LEU foil (up to 250 g in a single batch) and nickel fission recoil barrier in nitric acid at ambient pressure; and (2) the electrochemical dissolution of LEU foil in series of steps that produces an alkaline (basic) solution feed for 99 Mo purification. This report describes results from performance tests and design optimization of the ambient pressure, nitric-acid-dissolver system. The design, fabrication, and performance test planning for this system are described in more detail in previous reports (Jerden et al. 2011a,b, 2012). Full-scale demonstrations of both of the frontend processes using irradiated uranium foils are planned to be performed at Oak Ridge National Laboratory this fiscal year.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Intermediate Scale Laboratory Testing to Understand Mechanisms of Capillary and Dissolution Trapping during Injection and Post-Injection of CO 2 in Heterogeneous Geological Formations

Carbon Capture and Storage (CCS) represents a technology aimed to reduce atmospheric loading of CO 2 from power plants and heavy industries by injecting it into deep geological formations, such as saline aquifers. A number of trapping mechanisms contribute to effective and secure storage of the injected CO 2 in supercritical fluid phase (scCO 2 ) in the formation over the long term. The primary trapping mechanisms are structural, residual, dissolution and mineralization. Knowledge gaps exist on how the heterogeneity of the formation manifested at all scales from the pore to the site scales affects trapping and parameterization of contributing mechanisms in models. An experimental and modeling study was conducted to fill these knowledge gaps. Experimental investigation of fundamental processes and mechanisms in field settings is not possible as it is not feasible to fully characterize the geologic heterogeneity at all relevant scales and gathering data on migration, trapping and dissolution of scCO 2 . Laboratory experiments using scCO 2 under ambient conditions are also not feasible as it is technically challenging and cost prohibitive to develop large, two- or three-dimensional test systems with controlled high pressures to keep the scCO 2 as a liquid. Hence, an innovative approach that used surrogate fluids in place of scCO 2 and formation brine in multi-scale, synthetic aquifers test systems ranging in scales from centimeter to meter scale developed used. New modeling algorithms were developed to capture the processes controlled by the formation heterogeneity, and they were tested using the data from the laboratory test systems. The results and findings are expected to contribute toward better conceptual models, future improvements to DOE numerical codes, more accurate assessment of storage capacities, and optimized placement strategies. This report presents the experimental and modeling methods and research results.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

Waste Treatment of Acidic Solutions from the Dissolution of Irradiated LEU Targets for 99-Mo Production

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99 Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99 Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is the calcination of the uranium waste from 99 Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99 Mo production.

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

Chlorite, Biotite, Illite, Muscovite, and Feldspar Dissolution Kinetics at Variable pH and Temperatures up to 280 C

Summary Sheet silicates and clays are ubiquitous in geothermal environments. Their dissolution is of interest because this process contributes to scaling reactions along fluid pathways and alteration of fracture surfaces, which could affect reservoir permeability. In order to better predict the geochemical impacts on long-term performance of engineered geothermal systems, we have measured chlorite, biotite, illite, and muscovite dissolution and developed generalized kinetic rate laws that are applicable over an expanded range of solution pH and temperature for each mineral. This report summarizes the rate equations for layered silicates where data were lacking for geothermal systems.

15 GEOTHERMAL ENERGY↗
DOE OSTI2020

AX Farm and A Farm Third Retrieval Technology Chemical Dissolution Options Evaluation

In response to DOE-ORP direction to prepare a Practicability Evaluation for Waste Tank AX-102, and to address the Washington State Department of Ecology recommendation to evaluate the use of acid reagents for chemical dissolution, an Integrated Project Team (IPT) consisting of retrieval, process chemistry, corrosion, closure and laboratory subject matter experts was assembled to evaluate chemical dissolution options. This document provides a summary of these efforts and is intended to aid in the evaluation and third retrieval technology decision process not only for AX-102, but for all the AX and A Farm tanks, since the tanks contain similar waste and are flat-bottomed with uneven bottom plates.

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

Molybdenum Recovery from Filters Used in Large-scale Dissolution of Sintered Mo-disks

99 Mo production with linear accelerators can be achieved via the bremsstrahlung photonuclear reaction 100 Mo(γ,n) 99 Mo or the neutron capture reaction 98 Mo(n,γ) 99 Mo. For commercial producers, maximum recovery of enriched 98 Mo and 100 Mo target material is critical for sustaining an economic production cycle. During the peroxide dissolution of Mo metal disks and subsequent conversion to K 2 MoO 4 , several grams of Mo can be lost during filtration from solution when several hundred grams of sintered Mo disks are processed. This investigation shows that 5–8 g Mo is routinely retained on the filter units, but it can be almost fully recovered using aqueous washes. Washing can be done immediately and incorporated into the dissolution procedure, or it can occur several months after the initial filtration process to decrease processing time.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

The Dissolution of Cr and Fe at 850°C in FLiNaK and FLiBe

Understanding the solubility of common alloying elements, such as Cr and Fe, in fluoride salts, can provide valuable insights into the degradation process of alloys in these salts and assist in modeling the attack observed in static and flowing conditions. Previous isothermal dissolution experiments at 550°-750°C in FLiNaK showed little effect of time or temperature on the amount of dissolution, which was not expected. Therefore, the purpose of this milestone was to extend those observations to 850°C where Ni-based alloys have been developed with 100+X better creep strength than the historic Hastelloy N alloy developed for molten salt reactors more than 50 years ago. Exposures were conducted in the same commercial FLiNaK and FLiBe salts used in the lower temperature exposures.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development of a bench-scale dissolution concept for the direct extraction of nuclear fuel

Current used nuclear fuel reprocessing efforts utilize a hydrometallurgical approach in which the UNF is dissolved in hot nitric acid followed by solvent extraction into an organic solvent to harvest target nuclides. Previous studies have shown that the dissolution and loading process could be combined into a single organic dissolution/extraction step, producing loaded organic in a single step process. This single step process also includes the advantage of selectively targeting key nuclides in the dissolution while leaving undesirable constituents as part of the undissolved solids. This process is referred to hereafter as direct extraction. Ongoing research from multiple national labs has proven the effectiveness of this technique at research scale. Therefore, potential methods to implement direct extraction at both bench and industrial scale have been developed. The key features of potential dissolver system designs were identified via the team at Pacific Northwest National Laboratory and several designs based on industrial counterparts were assessed for feasibility. This report summarizes the advantages and disadvantages of multiple methods, concluding with a path forward to create multiple unique dissolver designs. The first design will be a single stage recirculating eductor mixer. The second design recommendation is a stator rotor static mixing flow loop design. Each dissolver could be utilized separately, simultaneously, or in series to answer questions surrounding reaction kinetics including residence time, provide a proof of concept for targeted extractions of specific nuclides, and inform needs for industrial scale implementation

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Recycling of PET by Dissolution-Purification-Recovery (DPR) Process Using Bioderived Solvent

Polyethylene terephthalate (PET) waste continues to accumulate at staggering rates, with the majority still routed to landfills due to limitations in current recycling methods. While mechanical recycling is widely implemented, it struggles to effectively remove colorants, additives, and harmful contaminants - resulting in discolored, degraded materials with limited reuse potential. In this presentation, we demonstrate a scalable dissolution-purification-recovery (DPR) process designed to selectively reclaim high-purity PET from post-consumer bottle flake. The method utilizes a food grade bio-based solvent for targeted PET dissolution, followed by activated carbon treatment for impurity removal and antisolvent-induced precipitation for polymer recovery. The resulting recycled PET exhibits complete decolorization, significantly reduced metal contamination, and minimal loss in molecular weight. These performance metrics indicate strong suitability for remanufacturing into new bottles, offering a circular alternative to traditional mechanical recycling pathways.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗