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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2021

The effects of low- Z shielding on uranium isotope discrimination using the time-emission profiles of long-lived delayed neutrons

Characteristic radiation signals can often be obscured or distorted by relatively small amounts of shielding, as is the case for highly-enriched uranium (HEU). Methods that can detect and identify special nuclear material (SNM) even in the presence of shielding are therefore of high value in many nuclear security and nonproliferation applications. It has been previously shown that the buildup and decay time profiles of long-lived delayed neutron emission from induced fission are unique and can be used to successfully discriminate between uranium isotopes and infer enrichment without the need for a calibration standard. Because the long-lived delayed neutron precursors exhibit decay times on the order of tens of seconds, their time profiles are not sensitive to the delays associated with scattering or diffusion in shielding materials, which occur on much shorter time scales. The use of capture-based neutron detectors can also mitigate the effects of shielding, because unlike recoil-based detectors, they do not have a low-energy detection threshold, so that neutrons that lose their energy in the shielding are still detectable. In this work, the effects of low-Z shielding on the discrimination of uranium isotopes using delayed neutron time profile measurements are investigated for the first time. Furthermore, the delayed neutron buildup and decay time profiles of polyethylene-shielded HEU and depleted uranium objects are measured using two different capture-gated detector designs. The recorded profiles show good agreement with the shapes predicted from six-group delayed neutron models even with the shielding in place, and the enrichment of the HEU and DU objects was estimated accurately to within 5%. While this method shows potential for discriminating and characterizing extreme cases of enrichment when a moderate amount of shielding is present, simulation results suggest that significant limitations exist for intermediate enrichment levels due to distortions to the delayed neutron time profile caused by neutron reflection in the shielding.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Strategies toward the Synthesis of Advanced Functional Sorbent Performance for Uranium Uptake from Seawater

A polymer fiber-based adsorbent (AF1) composed of acrylonitrile and itaconic acid functional groups was synthesized by a radiation-induced graft polymerization technique onto hollow-gear shaped polyethylene (PE) fibers. Investigation of the optimum reaction parameters for the conversion of grafted cyano moieties into amidoxime groups was conducted by the reaction with hydroxylamine at different temperatures and periods in a variety of aqueous and organic solvents. The 13 C CP/MAS spectra of amidoximated AF1 adsorbent fibers from water–methanol and dimethyl sulfoxide (DMSO) revealed that both the cyclic imide dioxime and open-chain amidoxime were formed through the reaction. The conversion from amidoxime to imide dioxime was found to occur slowly and gradually with the increasing reaction time. The quicker diffusion of DMSO as compared to that of water–methanol, in the grafted trunk PE fiber, resulted in faster kinetics of the amidoximation reaction. The uranium adsorption capacity of the amidoximated AF1 samples was determined after (i) 24 h contact with sodium-based brine spiked with uranium and (ii) 56 days of exposure to filtered seawater (Sequim Bay, WA, USA) in flow-through columns. Finally, the uranium extraction performance of the adsorbents after exposure to filtered seawater was consistent with the laboratory screening results, and the amidoximated AF1 samples (in DMSO at 70 °C for 3 h) exhibited the highest 56 day uranium adsorption capacity (5.04 ± 0.15 g U/kg-ads) with faster adsorption kinetics compared to the pristine AF1 adsorbents.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Amidinate Supporting Ligands Influence Molecularity in Formation of Uranium Nitrides

Uranium nitride complexes are attractive targets for chemists as molecular models for the bonding, reactivity, and magnetic properties of next-generation nuclear fuels, but these molecules are uncommon and can be difficult to isolate due to their high reactivity. Here, we describe the synthesis of three new multinuclear uranium nitride complexes, [U(BCMA) 2 ] 2 (μ-N)(μ-κ 1 :κ 1 -BCMA) (7), [(U(BIMA) 2 ) 2 (μ-N)(μ-N i Pr)(K 2 (μ-η 3 :η 3 -CH 2 CHN i Pr)] 2 (8), and [U(BIMA) 2 ] 2 (μ-N)(μ-κ 1 :κ 1 -BIMA) (9) (BCMA = N,N-bis(cyclohexyl)methylamidinate, BIMA = N,N-bis(iso-propyl)methylamidinate), from U(III) and U(IV) amidinate precursors. By varying the amidinate ligand substituents and azide source, we were able to influence the composition and size of these nitride complexes. 15 N isotopic labeling experiments confirmed the bridging nitride moieties in 7–9 were formed via two-electron reduction of azide. The tetra-uranium cluster 8 was isolated in 99% yield via reductive cleavage of the amidinate ligands; this unusual molecule contains nitrogen-based ligands with formal 1–, 2–, and 3– charges. Additionally, chemical oxidation of the U(IV) precursor U(N 3 )(BCMA) 3 yielded the cationic U(V) species [U(N 3 )(BCMA) 3 ][OTf]. Magnetic susceptibility measurements confirmed a U(IV) oxidation state for the uranium centers in the three nitride-bridged complexes and provided a comparison of magnetic behavior in the structurally related U(III)-U(IV)-U(V) series U(BCMA) 3 , U(N 3 )(BCMA) 3 , and [U(N 3 )(BCMA) 3 ][OTf]. At 240 K, the magnetic moments in this series decreased with increasing oxidation state, i.e., U(III) > U(IV) > U(V); this trend follows the decreasing number of 5f valence electrons along this series.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Theoretical analysis of uranium-doped thorium dioxide: Introduction of a thoria force field with explicit polarization

Thorium dioxide is used industrially in high temperature applications, but more insight is needed into the behavior of the material as part of a mixed-oxide (MOX) nuclear fuel, incorporating uranium. We have developed a new interatomic potential model including polarizability via a shell model, and commensurate with a prominent existing UO{sub 2} potential, to conduct configurational analyses and to investigate the thermophysical properties of uranium-doped ThO{sub 2}. Using the GULP and Site Occupancy Disorder (SOD) computational codes, we have analyzed the distribution of low concentrations of uranium in the bulk material, where we have not observed the formation of uranium clusters or the dominance of a single preferred configuration. We have calculated thermophysical properties of pure thorium dioxide and Th{sub (1−x)}U{sub x}O{sub 2} which generated values in very good agreement with experimental data.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Chemical speciation correlated with microstructural heterogeneity of interdicted uranium materials

Two uranium powders seized by law enforcement in Victoria, Australia, have been characterized by established nuclear forensic methods in a previously published study. Here, in this work, the results of further characterization by a scanning transmission x-ray microscope (STXM) operating in the soft x-ray regime are reported. STXM images are used to estimate the elemental distribution in micrometer-scale particles of each powder, and oxygen K-edge absorption spectra are used to determine the chemical state of uranium. The results of the current study are consistent with the previous analysis; the first powder is found to be a potassium-uranium hydrate, while the second powder is determined to be a mixture of uranium oxides primarily consisting of UO 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Uranium Hydride Nucleation and Growth Model FY'16 ESC Annual Report

Uranium hydride corrosion is of great interest to the nuclear industry. Uranium reacts with water and/or hydrogen to form uranium hydride which adversely affects material performance. Hydride nucleation is influenced by thermal history, mechanical defects, oxide thickness, and chemical defects. Information has been gathered from past hydride experiments to formulate a uranium hydride model to be used in a Canned Subassembly (CSA) lifetime prediction model. This multi-scale computer modeling effort started in FY’13, and the fourth generation model is now complete. Additional high-resolution experiments will be run to further test the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

FY16 Status Report for the Uranium-Molybdenum Fuel Concept

The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Risk Assessment and Management of Uranium Mill Tailings Legacies - 17508

exploitation activities took place at 61 sites. The great majority of these mining sites are located in central-east Portugal, including the Urgeirica mine that was the most important uranium deposit exploited in the country. The mining activities at this site generated 1.6 million m{sup 3} of radioactive wastes disposed in an open air area originating the largest radioactive tailings pile of the country. The environmental remediation of the Urgeirica site was one of the main priorities of the national plan for the rehabilitation of former uranium mining sites. Site remediation, that included the mine industrial area with several support infrastructures and a processing plant, was executed between 2006 and 2008 (1. phase) and between 2008 and 2014 (2. phase). Nevertheless, the environmental rehabilitation of the Portuguese uranium legacies sites is not yet concluded. One of the last largest mining complexes, the Quinta do Bispo open pit, is scheduled to be rehabilitated only in 2020. This work presents a study to evaluate the environmental impacts from the uranium mill tailings legacies before and after the remediation works, focusing on two different approaches: a deterministic methodology for dose assessment and a probabilistic methodology with the Monte Carlo method to estimate the incremental lifetime risk. Probabilistic distributions of the input parameters were used in the risk calculation approach. The methodology for dose and risk assessments may be applied to other sites that are still under intervention and the results may be used to prioritize the remediation stages. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Uranium mining and norm, a North American perspective - 15545

All soils and rocks contain naturally occurring radioactive materials (NORM). Many ores and raw materials contain relatively elevated levels of natural radionuclides and processing such materials can further increase the concentrations of natural radionuclides, sometimes referred to as technologically enhanced naturally occurring radioactive materials (TENORM). Examples of NORM minerals include uranium ores, monazite (a source of rare earth minerals), and phosphate rock used to produce phosphate fertilizer. Such activities have the potential to result in above background radiation exposure to workers and the public. Following a brief review of the sources and exposure from NORM in these varied industries, this paper will then present an overview of uranium mining and recovery in North America including discussion on the methods currently being used, production of NORM materials and wastes associated with these uranium recovery methods and the radiological composition of the NORM wastes produced by uranium recovery. The paper includes descriptions of the methods and associated NORM produced by both conventional (underground, open pit) and in situ recovery (ISR) mining methods and identifies the responsible government agencies in the US and Canada assigned the authority to regulate and control these NORM materials. (authors)

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

High Energy X-ray Applications for the Characterization of Metallic Uranium Alloys

Uranium-based metallic alloys are both candidates for fuels in high performance research and test reactors, as well as in fast spectrum reactors as nuclear transmutation fuels that can be utilized to burn long-lived minor actinides and fission products. Uranium-molybdenum (UMo) alloys are under development for the use in research and test reactors. Whereas, uranium-zirconium (U-Zr) alloys represent one of the types of fuels that are under consideration for use within the fast spectrum reactors. Both of these types of metallic fuels have been studied for over five decades in various types of reactors. Many of the previous studies had focused on performance-based aspects. However, recent advances in characterization tools and techniques, as well as short-term (low fluence) neutron irradiation facilities have contributed to a more in-depth understanding of nuclear fuel and material behavior following irradiation. This knowledge is crucial to the enhancement of current modeling and simulation techniques, particularly with respect to parameterization and validation. This paper focuses on the use of high energy x-rays to characterize metallic nuclear fuel. The specimens were composed of depleted uranium (DU) alloyed with either Zr or Mo, where DU-xZr (x = 10, 15, or 20 wt.%) and DU-yMo (y = 7 or 10 wt.%). The geometries of the specimens were 3 mm TEM disks that were ∼200 μm in thickness. The fuel alloys were initially prepared via an arc melting process. The samples were irradiated at the Advanced Test Reactor at Idaho National Laboratory (INL) using the hydraulic rabbit system. The hydraulic rabbit system allows for well-controlled, short term irradiations (less than one cycle), hence resulting in low fluences. The U-based alloys were irradiated to doses of 0.01 and 0.1 dpa. The irradiation temperature of the specimens ranged from 150 deg. C to 800 deg. C. Synchrotron characterization techniques can be used to provide insight into material microstructure ranging a multitude of spatial scales, from the atomic scale to the macroscale. High-energy x-rays at the National Synchrotron Light Source II (NSLS II) and the Advanced Photon Source (APS) have been utilized to understand the microstructural changes in these metallic alloys, specifically with respect to phase identification, phase fractions, chemical bonding, strain, and dislocation density through the use of diffraction and X-ray Absorption Spectroscopy (XAS). These experiments utilized a robot for specimen exchange during the diffraction measurements, allowing for remote handling of radioactive materials. This is an example of another recent advancement in characterization techniques that enables research with nuclear materials. Detailed results of the analyses of the microstructural changes are pending.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Integrated Methodology for the Impact Assessment of an Uranium Mill Contaminated Site - 19031

This work describes the development of mathematical models for radionuclides distribution and transport in the environment. The applicability of the models involves an integrated approach to assess the quantitative dispersion of the radionuclides in different environmental compartments when released from uranium tailings or from other low-activity mining wastes storages. For every compartment the important pathways of contamination and the radionuclides of major concern are identified and selected, based on their chemical, physical, radiological and toxic properties. Each compartment represents an environmental media: air, soil, water, vegetation and biota. Depending on the characteristics of the sources different types of possible releases are evaluated using phenomenological models. Models of environmental release, dispersion, transport and fate within each environmental compartment, as well as models of transfer between them, are applied to a specific site. The model outputs are the radionuclides activities for each one of the environmental compartments, where relevant exposure end-points are previously identified. The methodology was applied to a specific Uranium Mill Site, referred here as the 'Mill'. The site comprises approximately 3.2 km and includes two inactive mills, former ore stockpiles areas, a partially reclaimed tailings pond disposal area and a current tailings pond disposal area. The Mill site began operations in the late 1950's extracting uranium ore by alkaline leaching. Later on, a new facility using acid leaching for extracting uranium begun to operate in 1980. The mill operated first continuously and after intermittently until its closure, in March 2006. The data required for the model input was acquired from different sources: direct measurements from this particular site and available from multiple site investigation reports (in particular, environmental data for groundwater, soil, surface water, sediments and biota); other, were selected from published data. For the unknown data, assumptions were made. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Hydration of uranium residues contained in enriched UF{sub 6} cylinders - 14393

The transport of uranium hexafluoride (UF{sub 6}) is carried out in large steel cylinders in accordance with the requirements provided in standards ISO 7195 and ANSI N14.1. During filling and extraction operations, some air in-leakage in cylinders is possible and its moisture content results in the hydrolysis of UF{sub 6} and the creation of non-volatile uranium residues, in particular UO{sub 2}F{sub 2} complexes with H{sub 2}O and HF. The main purpose of this paper is to justify, for criticality safety assessment of enriched UF{sub 6} transport cylinders, a bounding hydration ratio H/U of 6 for uranium residues, based on the operating conditions of UF{sub 6} cylinders in fuel facilities. Moreover, the criticality safety assessment of UF{sub 6} cylinders transport should also consider water ingress during the immersion test of accident conditions of transport. Indeed, this water ingress creates new hydrated residues and its moderation impact on initial residues must be considered. The last part of this paper shows that this water ingress does not challenge the bounding hydration of uranium residues determined in the first part, due to the presence of UF{sub 6} in excess. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Proven Uranium Recovery Solutions for the Nuclear Fuel Manufacturing Market - 16060

During the Nuclear Fuel Manufacturing process a number of by-products and residues are generated from various sources. In recent years, Westinghouse has characterised and processed around 50,000 drums of residue comprising of numerous different types generated over 60 years from the manufacture of numerous types of fuel for greater than 50 reactors. These include, AGR, Magnox (uranium metal fuel), PWR, fast reactors and various test reactors. Different organisations have different definitions and names for each of these categories but essentially they fall into one of the following categories: Clean Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Dirty Scrap (U{sub 3}O{sub 8}, UO{sub 2}, UO{sub 3}, UF{sub 4}, powders, pellets), Gd Contaminated Scrap (U{sub 3}O{sub 8}, UO{sub 2}, powders, pellets), Ash, Sludge, Combustible / soft residues (e.g. ventilation filters, gloves, plastic bags), Contaminated metals (including contaminated scrap zirconium tubes), Materials arising from decommissioning. Routes have been developed across the industry for a number of the categories listed above, however in numerous examples, routes have not been established and materials have been stored creating a significant legacy. Westinghouse understands the issues regarding legacy uranic material and the financial and regulatory impact of storage. Interim storage of materials may seem like a low cost option in the short term; however ongoing costs for facilities, manpower and equipment, without a permanent and final solution can result in significant lifetime material management costs. Recovery of uranium and sending materials for safe disposal of enriched uranium residues is not easy because of the criticality risk both during processing and at a disposal site. Therefore quantities of materials than can be sent via permitted disposal routes are very limited The development of regulations often means that finding solutions is becoming more difficult. It is believed that timely resolution is ultimately the most cost effective solution. In addition, the intrinsic value of recycling this material is also high and in some cases can mitigate the cost of processing. (authors)

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

Assessment of Controlling Processes for Field-Scale Uranium Reactive Transport under Highly Transient Flow Conditions

This paper presents the results of a comprehensive model-based analysis of a uranium tracer test conducted at the U.S Department of Energy Hanford 300 Area (300A) IFRC site. A three-dimensional multi-component reactive transport model was employed to assess the key factors and processes that control the field-scale uranium reactive transport. Taking into consideration of relevant physical and chemical processes, the selected conceptual/numerical model replicates the spatial and temporal variations of the observed U(VI) concentrations reasonably well in spite of the highly complex field conditions. A sensitivity analysis was performed to interrogate the relative importance of various processes and factors for reactive transport of U(VI) at the field-scale. The results indicate that multi-rate U(VI) sorption/desorption, U(VI) surface complexation reactions, and initial U(VI) concentrations were the most important processes and factors controlling U(VI) migration. On the other hand, cation exchange reactions, the choice of the surface complexation model, and dual-domain mass transfer processes, which were previously identified to be important in laboratory experiments, played less important roles under the field-scale experimental condition at the 300A site. However, the model simulations also revealed that the groundwater chemistry was relatively stable during the uranium tracer experiment and therefore presumably not dynamic enough to appropriately assess the effects of ion exchange reaction and the choice of surface complexation models on U(VI) sorption and desorption. Furthermore, it also showed that the field experimental duration (16 days) was not sufficiently long to precisely assess the role of a majority of the sorption sites that were accessed by slow kinetic processes within the dual domain model. The sensitivity analysis revealed the crucial role of the intraborehole flow that occurred within the long-screened monitoring wells and thus significantly affected both field-scale measurements and simulated U(VI) concentrations as a combined effect of aquifer heterogeneity and highly dynamic flow conditions. Overall, this study, which provides one of the few detailed and highly data-constrained uranium transport simulations, highlights the difference in controlling processes between laboratory and field scale that prevent a simple direct upscaling of laboratory-scale models.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Spatial Engineering Direct Cooperativity between Binding Sites for Uranium Sequestration

Preorganization is a basic design principle used by nature that allows for synergistic pathways to be expressed. Herein, a full account of the conceptual and experimental development from randomly distributed functionalities to a convergent arrangement that facilitates cooperative binding is given, thus conferring exceptional affinity toward the analyte of interest. The resulting material with chelating groups populated adjacently in a spatially locked manner displays up to two orders of magnitude improvement compared to a random and isolated manner using uranium sequestration as a model application. This adsorbent shows exceptional extraction efficiencies, capable of reducing the uranium concentration from 5 ppm to less than 1 ppb within 10 min, even though the system is permeated with high concentrations of competing ions. The efficiency is further supported by its ability to extract uranium from seawater with an uptake capability of 5.01 mg g -1 , placing it among the highest-capacity seawater uranium extraction materials described to date. The concept presented here uncovers a new paradigm in the design of efficient sorbent materials by manipulating the spatial distribution to amplify the cooperation of functions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Reduction of Carbonyl Groups by Uranium(III) and Formation of a Stable Amide Radical Anion

In this work, methyl benzoate, N , N -dimethylbenzamide, and benzophenone were reduced by U III [N(SiMe 3 ) 2 ] 3 resulting in uranium(IV) products. Reduction of benzophenone lead to U IV [OC ∙ Ph 2 )][N(SiMe 3 ) 2 ] 3 , ( 1.1 ) which forms the dinuclear complex, [N(SiMe 3 ) 2 ] 3 U IV (OCPhPh-CPh 2 O)U IV [N(SiMe 3 ) 2 ] 3 ( 1.2 ), through coupling of the ketyl radical species upon crystallization. Reaction of N , N -dimethylbenzamide with U III [N(SiMe 3 ) 2 ] 3 resulted in U IV [OC ∙ (Ph)(NMe 2 )][N(SiMe 3 ) 2 ] 3 ( 2 ), a uranium(IV) compound and the first example of a charge-separated amide radical. In the case of methyl benzoate, the reduction resulted in U IV (OMe)[N(SiMe 3 ) 2 ] 3 ( 3 ) and benzaldehyde as the reduced organic fragment. Compound 2 showed the ability to act as a uranium(III) synthon in its reactivity with trimethylsilyl azide, a reaction that yielded U V (=NSiMe 3 )[N(SiMe 3 ) 2 ] 3 . Additionally, 2 was reduced with potassium graphite resulting in [U(μ-O)[O=C(NMe 2 )(Ph)][N(SiMe 3 ) 2 ] 2 ] 2 ( 4 ), a dinuclear uranium compound bridged by oxo ligands. Reduction of 2 in the presence of 15-crown-5 afforded isolation of the mono-oxo compound, [(15-crown-5) 2 K][UO[N(SiMe 3 ) 2 ] 3 ] ( 5 ). Lastly, the results expand the reduction capabilities of U III complexes and demonstrate a strategy for isolating novel metal-stabilized radicals.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Rapid paper spray mass spectrometry characterization of uranium and exemplar molecular species

The ability to detect and quantify the presence of specific inorganic elements and complexes is essential for environmental monitoring and nuclear safeguards applications. In this work, paper spray ionization mass spectrometry (PSI–MS) was used for the rapid chemical and isotopic characterization of trace inorganic species collected on cotton swipe substrates. Here, the direct analysis of cotton swipes using this ambient ionization technique led to fast sample analysis that retained original chemical information of the source material with minimal sample preparation. Mass spectra were collected on an atmospheric pressure ionization, high–resolution mass spectrometer for solutions containing uranyl acetate, uranyl chloride, uranyl nitrate, and uranyl tri– n –butylphosphate (uranyl–TBP) complexes. Gadolinium nitrate was used as an internal standard for the quantitative analysis of uranium. To demonstrate the ability to characterize inorganic contaminants in the presence of uranium, a multi–element inorganic standard containing U, Bi, Pb, Cd, Fe, and Zn was deposited onto cotton substrates and directly analyzed without purification. All elements doped on the cotton substrate were detected with strong signal–to–noise (S/N) ratios (S/N~1000 for UO 2 + on multi–element doped swipes) and high integrated intensities (>10 5 counts) from collection periods of approximately 1 min. Limits of detection were determined to be approximately 94 ng for UO 2 + and uranyl acetate through the measurement of ppb level solutions. The rapid analysis of uranium and other inorganic–containing samples while still retaining original chemical information (e.g. uranyl complexation) was demonstrated. Qualitative detection and speciation were achieved in less than 1 min of analysis. For uranium isotopic quantitation, longer accumulations (>15 min) can be sustained to improve the accuracy of minor 235 U isotopic abundance measurements to approximately 1% error.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗