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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analysis of solid uranium particulates on cotton swipes with an automated microextraction-ICP-MS system

Here, an automated microextraction method coupled to an inductively coupled plasma – mass spectrometer (ICP-MS) was developed for the direct analysis of solid uranium particulates on the surface of cotton swipes. The microextraction probe extracts particulates from the sample surface, in a flowing solvent, and directs the removed analyte to an ICP-MS for isotopic determination. The automated system utilizes a mechanical XY stage that is software controlled with the capability of saving and returning to specific locations and a camera focused to the swipe surface for optimal viewing of the extracted locations (i.e., material present). Here, particulates (n = 135) were extracted and measured by ICP-MS, including 35 depleted uranyl nitrate hexahydrate (UN) (used for mass bias corrections), 50 uranyl fluoride (UO 2 F 2 ), and 50 uranyl acetate (UAc) particulates. Blank extractions were performed on the cotton swipes between triplicate sample analyses. Between each swipe extraction, the probe was sent between two wells containing 10% and 5% HNO 3 to clean the probe head and to eliminate any analyte carryover between particulates. The measured 235 U/ 238 U and 234 U/ 238 U isotope ratios for the UO 2 F 2 particulates were 0.00725(8) and 0.000054(4), a percent relative difference (% RD) of –0.041% and –1.7% from the reference isotope ratios determined in-lab through multi-collector ICP-MS analysis of dissolved aliquots of the U material. The UAc samples had a measured 235 U/ 238 U isotope ratio of 0.00206(7), a –0.96% relative difference from the reference value of 0.00208(1). The 234 U/ 238 U and 236 U/ 238 U isotope ratios were 0.000008(1) and 0.000031(4), –5.1% RD and –4.3% RD, respectively. The automated sample stage enabled seamless and rapid particle analysis, leading to a significant increase in throughput versus what was previously possible. Additionally, the saved location capability reduced user sampling error as sampling locations were easily stored and recalled. Analysis of U particles on the swipe surface – including blanks, mass bias, and triplicate extractions – was completed in less than an hour without any sample preparation necessary.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Direct isotopic analysis of solid uranium particulates on cotton swipes by microextraction-ICP-MS

Direct isotope ratio analysis of solid uranium particulates on cotton swipes was achieved using a solution-based microextraction technique, coupled to a quadrupole inductively coupled plasma - mass spectrometer (ICP-MS). This microextraction-ICP-MS methodology provides rapid isotopic analysis which could be applicable to nuclear safeguards measurements. Particulates of uranyl nitrate hexahydrate (UO 2 (NO 3 ) 2 ·6H 2 O) and uranyl fluoride (UO 2 F 2 ) ranging from 6 μm to 40 μm in length were transferred to cotton swipes with a particle manipulator. The microextraction probe then delivers a 5% nitric acid (HNO3) solvent onto the swipe surface to extract the uranium species. The extracted sample is then delivered to the ICP-MS for isotopic determination. The majority of uranium signal (~99% and ~94% for UO2(NO3)2·6H2O and UO 2 F 2 , respectively) was detected in the first 15 s extraction, while subsequent extractions on the same location had low or no U signal, suggesting near complete removal of the solid uranium compounds from the swipe surface. Ten samples (for each of the uranium compounds), were analyzed for their isotopic composition. For UO 2 (NO 3 ) 2 ·6H 2 O, the determined isotope ratios resulted in a % relative difference (% RD) from the referenced isotope ratios of 0.97, 1.0, and 7.3% for 234 U/ 238 U, 235 U/ 238 U, and 236 U/ 238 U, respectively. The % RD of the UO 2 F 2 isotope ratios were 1.9 and 0.60% for 234 U/ 238 U and 235 U/ 238 U, respectively. The preliminary limits of detection were determined to be 0.002, 0.4, and 60 pg for 234 U, 235 U and 238 U, respectively This work demonstrates that microextraction ICP-MS is a rapid and sensitive method that could directly determine uranium isotope ratios of UO 2 (NO 3 ) 2 ·6H 2 O and UO 2 F 2 particulates on cotton swipes.

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

A new HALEU isotopic certified reference material

Candidate certified reference material (CRM) U196 was prepared by gravimetrically mixing and dissolving NNSA NRMP CRM 112-A (natural uranium metal) and CRM 116-A (highly enriched 93.2% 235 U metal) to achieve a high-assay low-enriched uranium (HALEU) isotopic abundance reference material. CRM U196 will be certified for the 233 U/ 238 U, 234 U/ 238 U, 235 U/ 238 U, and 236 U/ 238 U ratios. Gravimetrically determined uranium isotope amount ratios and values derived from the isotope data were calculated from the buoyancy corrected mass data and from certified values for the metal CRMs used to create the solution. The 235 U isotope amount fraction (·100) and expanded uncertainty is 19.5846 ± 0.0028. For attribute verification, total evaporation and modified total evaporation isotope ratio measurements were performed. In conclusion, the reference material is to be distributed as a unit containing 5 mg uranium as dry uranyl nitrate.

High-assay low-enriched uranium↗
DOE OSTI2022

Irradiation-induced amorphization of UO 2 films prepared by spraying-assisted combustion synthesis

Spraying-assisted combustion synthesis with uranyl nitrate – acetylacetone – 2-methoxyethanol solutions was used to prepare UO 2 films on an aluminum alloy substrate. The tuning of the spraying parameters and annealing temperatures allowed the preparation of UO 2 films with thicknesses varying from 10–300 nm and 5–10 nm UO 2 grain size. High-resolution electron microscopy and X-ray photoelectron spectroscopy showed that increasing the annealing temperature promotes Mg diffusion from the substrate into the films. The incorporation of Mg reduced the overall crystallinity of the films. The irradiation with Ar 2+ ions (1.7 MeV energy and a fluence of 2 × 10 16 ions/cm 2 ) did not degrade the quality of the films. However, the Mg content significantly influenced the irradiation-induced restructuring of the UO 2 films. Irradiated films with low or no Mg content exhibit high crystallinity, and the UO 2 /Al interfacial layer becomes highly porous. Films with higher Mg content are mostly amorphized after irradiation. The origin of irradiation-induced amorphization was related to the formation of Mg y U 1-y O 2±x solid solutions. Chemically complex, pore-free, and amorphous Mg-Al-O film/substrate interfacial layers enable continuous Mg diffusion during irradiation. As a result, the gradual increase in Mg amounts triggers irradiation-induced precipitation of a crystalline MgO-rich phase within the amorphous films.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Absolute Branching Ratios of Prominent $γ$ Rays Following $α$ Decay of 233 U

The absolute $γ$-ray branching ratios (emission probabilities) for 19 prominent peaks in the $α$ decay of 233 U are reported with energies ranging from 97 to 366 keV. The branching ratios were obtained from $α-γ$ coincidence measurements, normalized to the $α$ singles. The high-purity 233 U material (> 99.98%) was separated from its decay products before preparation of a 4.74-mg/cm 2 uranyl nitrate foil with a 6-mm diameter. Furthermore, this is the first report on the absolute normalization of the branching ratios since 1984 and it is the only measure by $α-γ$ coincidences, where the $α-γ/α$ ratio removes several potential systematic biases. Overall, the results are consistent with the adopted literature, and they validate the quantities most relevant to the non-destructive assay of fissile 233 U material, which is relevant to the thorium fuel cycle and associated safeguards.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Irradiation-enhanced Interactions at UO 2 /Al 2 O 3 /Al Interfaces

Combustion synthesis is used to prepare thin UO 2 films on aluminum alloy substrates. This simple preparation method involves electrospraying uranyl nitrate + acetylacetone + 2-methoxyethanol solution on the substrate, followed by a short annealing at 350 or 550 °C. The irradiation of films with a 40 Ar 2+ ion beam (energy of 1.7 MeV and fluences of 7.7 × 10 16 and 1.3 × 10 17 ions/cm 2 ) is conducted to investigate irradiation-induced restructuring processes. High-resolution transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS) investigations show that the annealing temperature significantly influences the crystallinity and stability of materials during irradiation. A small amount of Mg in the alloy substrate diffuses into the amorphous Al 2 O 3 interfacial layer between the film and the substrate. Local thermal spikes from the incoming ions facilitate the irradiation-induced mixing of immiscible Al 2 O 3 and UO 2 for the materials prepared at 350 °C. This mass diffusion produces relatively large cavities at the interface. Selective diffusion of a more significant amount of Mg for the materials prepared at 550 °C suppresses the mixing of the Al 2 O 3 interlayer with the film but forms Mg y U 1–y O 2±x solid solutions during irradiation. Local thermal heating triggers the precipitation of a discontinuous crystalline MgO layer close to the film surface. As a result, the enhanced and selective diffusion of Mg into the film makes the materials prepared at 550 °C more robust and mechanically stable than those prepared at 350 °C.

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

Irradiation-Driven Restructuring of UO 2 Thin Films: Amorphization and Crystallization

Combustion synthesis in uranyl nitrate–acetylacetone–2-methoxyethanol solutions was used to deposit thin UO 2 films on aluminum substrates to investigate the irradiation-induced restructuring processes. Thermal analysis revealed that the combustion reactions in these solutions are initiated at ~160 °C. The heat released during the process and the subsequent brief annealing at 400 °C allow the deposition of polycrystalline films with 5–10 nm UO 2 grains. The use of multiple deposition cycles enables tuning of the film thicknesses in the 35–260 nm range. Irradiation with Ar 2+ ions (1.7 MeV energy and a fluence of up to 1 × 10 17 ions/cm 2 ) is utilized to generate a uniform distribution of atomic displacements within the films. X-ray fluorescence (XRF) and alpha-particle emission spectroscopy showed that the films were stable under irradiation and did not undergo sputtering degradation. X-ray photoelectron spectroscopy (XPS) showed that the stoichiometry and uranium ionic concentrations remain stable during irradiation. The high-resolution electron microscopy imaging and electron diffraction analysis demonstrated that at the early stages of irradiation (below 1 × 10 16 ion/cm 2 ) UO 2 films show complete amorphization and beam-induced densification (sintering), resulting in a pore-free disordered film. Prolonged irradiation (5 × 10 16 ion/cm 2 ) is shown to trigger a crystallization process at the surface of the films that moves toward the UO 2 /Al interface, converting the entire amorphous material into a highly crystalline film. This work reports on an entirely different radiation-induced restructuring of the nanoscale UO 2 compared to the coarse-grained counterpart. The preparation of thin UO 2 films deposited on Al substrates fills an area of national need within the stockpile stewardship program of the National Nuclear Security Administration and fundamental research with actinides. Here, the method reported in this work produces pure, robust, and uniform thin-film actinide targets for nuclear science measurements

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

Occupational Contamination with a Highly Enriched Uranium Solution

This medical case report describes the first reported instance of occupational skin contamination with a uranyl nitrate solution containing highly enriched uranium. The report provides an overview of the unique medical treatment and management considerations in such a case. Herein, internal dose assessment is covered in detail. The discussion covers key points regarding uranium characteristics, chemical and radiological damage to body tissues from HEU exposure, and resources available for assistance with a case of radiological contamination. This information adds to the limited medical literature on this topic and provides a valuable reference for medical personnel when dealing with this uncommon problem.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2023

Challenges in correlating oxygen stable isotope ratios of hydrates on uranium ore concentrates to process waters

Exchange of oxygen stable isotopes (δ 18 O values) between precipitation waters and uranium oxides is governed by thermodynamics or kinetics. It has been assumed that meteoric waters can be related to precipitation waters in uranium ore concentrates and their calcination and reduced uranium oxide products. With this assumption, the δ 18 O values of uranium materials could provide forensic signatures that identify the production history and geolocation of nuclear materials. To further exploit the potential of δ 18 O values in nuclear material analysis, this study examines the oxygen stable isotope exchange in two UOCs, magnesium diuranate (MDU) and sodium diuranate (SDU). MDU and SDU were synthesized from solutions of uranyl nitrate hexahydrate using precipitation waters with unique oxygen isotope compositions. The structures of the MDU and SDU were analyzed using powder X-ray diffraction (p-XRD) and thermal mass loss curves, while the δ 18 O values of waters generated during thermal decomposition were analyzed using a thermogravimetric analyzer coupled to an isotope ratio infrared spectrometer (TGA-IRIS). By p-XRD, the MDU was uniform and amorphous across all syntheses with residual crystalline material incorporated as a minor component. Combined with the TGA results, all of the MDU is likely amorphous MgU 2 O 7 ·3H 2 O with MgO impurities present throughout. In contrast, the SDU synthesis resulted in multiple phases with many samples exhibiting crystalline phases including a combination of Na(UO 2 ) 4 O 2 (OH) 5 ·5H 2 O and Na 2 (UO2) 6 O 4 (OH) 6 ·8H 2 O with a Na 2 U 2 O 7 minor phase. A small fraction of the SDU samples were amorphous with no crystalline XRD peaks observed. Mass loss curves of the SDU samples revealed that the amorphous samples contained inclusions of similar crystalline phases compared to the crystalline materials. The uniformity of the MDU samples enabled highly reproducible measurements of δ 18 O values of the water vapor yielded for two dehydration events at 170 °C and 500 °C. In contrast, the multiphase composition of the SDU samples resulted in poor reproducibility in δ 18 O values. In conclusion, neither system revealed any correlation between the δ 18 O values of precipitation water, and the waters released during dehydration of the UOCs.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Applications of Alternate Actinide Calibration on Inductively Coupled Plasma Mass Spectrometry (ICPMS) Analysis of H-Canyon Highly Enriched Uranium Legacy Material at the Savannah River Site (SRS)

The Savannah River National Laboratory (SRNL) uses Inductively Coupled Plasma Mass Spectrometry (ICPMS) to support numerous programs at the Savannah River Site (SRS). As part of an objective to prepare high assay low enriched uranium (HALEU), the SRS H-Canyon needed analysis of legacy uranyl nitrate that is stored at its Outside Facilities (HOF). With the application of a standard practice for alternate actinide calibration on ICP-MS, SRNL has successfully demonstrated its capability to characterize the uranium/actinide composition of the HOF tanks that are intended for consolidation prior to preparation of the HALEU. A comparison of the uranium analytical results between recently upgraded SRNL ICP-MS and Inductively Coupled Plasma Optical Emission Spectrometry (ICP-OES) instrumentation shows the alternate actinide calibration is a valid complementary method when used in conjunction with other analytical techniques (Davies-Gray Potentiometric Titration). The methods, instrumentation, and ongoing developments on ICP-MS at SRNL will be discussed.

Jones, Mark A.↗
DOE OSTI2024

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

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

Rapid Oxidation of Uranium Steel Alloys: Combustion Synthesis

This project explored the use of combustion synthesis as a rapid, high-temperature method for oxidizing uranium-bearing steel alloys. Traditional laboratory-scale synthesis methods often fail to replicate the thermal and kinetic conditions experienced by real-world particulates, particularly those formed under rapid quenching or high-temperature scenarios. Combustion synthesis offers a promising alternative by enabling fast, localized heating and flexible precursor selection. A series of targeted experiments were conducted using a U 2 NiCrFe 4 alloy as the precursor. The alloy was oxidized using combustion synthesis reactions fueled by uranyl nitrate and glycine, achieving peak temperatures exceeding 1,200 °C. Postreaction analysis using scanning electron microscopy (SEM), elemental mapping, and Raman spectroscopy revealed the formation of iron-based oxides, with limited but detectable evidence of uranium oxide phases such as UO 2 . The results indicate that under the rapid reaction and cooling conditions of combustion synthesis, iron oxides form preferentially, but uranium oxide formation is kinetically limited. These findings validate combustion synthesis as a viable method for simulating the oxidation behavior of uranium steels in extreme environments and lay the groundwork for future studies aimed at enhancing uranium oxide formation through higher temperatures or modified precursor compositions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Application of Raman Spectroscopy to Determine Uranium Content in ADUN Solution

The work presented in this report is part of the ongoing efforts to address the nuclear material control and accounting needs for advanced reactor fuel fabrication facilities. This work was supported by the Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy Office of Nuclear Energy‘s Nuclear Fuel Cycle and Supply Chain program. The activities and engagements under the MPACT program are designed to support a robust US civilian nuclear energy enterprise. In the work described in this report, we supported MPACT objectives by developing measurement techniques that could be used for material accounting and process monitoring and by working with industry partners to identify existing gaps and areas for improvement. Oak Ridge National Laboratory has been working with commercial tristructural isotropic (TRISO) fuel fabricators such as Standard Nuclear to develop technology for rapid and cost-effective uranium content assessment. This work has focused on demonstrating advanced measurement techniques (e.g., Raman spectroscopy) that can be used for rapid, reliable, and cost-effective routine measurements of uranium content in feed solutions and liquid waste streams as well as for monitoring in-line process measurements and product streams. Specifically, this report explores techniques for accurately determining uranium content in acid-deficient uranyl nitrate (ADUN) solutions and detecting low uranium concentrations in ammonia solutions. Developing such measurement techniques will benefit TRISO fuel fabrication facilities, facilities involved in other parts of the fuel cycle that require online monitoring of aqueous solutions, and potentially molten salt fuel reactors. This work supports developing Raman spectroscopy procedures to determine uranium concentrations in ADUN solutions, which are used as feedstock in the sol–gel process for creating TRISO fuel. Some additional benefits of using Raman spectroscopy for uranium quantification in fabrication facilities include enabling online monitoring of the chemical process, which would provide near real-time feedback; eliminating the need for sample transfers, preparation, or dilution; providing nondestructive measurements; and user friendliness. In this fiscal year, FY25, we determined the identity of the unknown Raman band at approximately 853 cm−1 that was discovered in ADUN Raman spectra in FY24, created calibration curves and determined uranium concentrations of two ADUN solutions, and compared the Raman results to results obtained from inductively coupled plasma mass spectrometry and Davies–Gray titration. Furthermore, we have identified focus areas for experimentation in future fiscal years. A key result is that the accuracy of using Raman can provide accuracy comparable to destructive analysist techniques, With a well-developed calibration curve, using standards and a large number of samples (more than five samples), uncertainty on the order of 1%–3% is achievable. Given that the uncertainties achieved by Raman spectroscopy were on the order of the uncertainties achieved using ICP-MS, we conclude that with a well-developed procedure Raman spectroscopy can be used to determine uranium concentrations in ADUN solutions for NMC&A applications. The benefits of such an approach are that the time and effort will be less than that of comparable destructive analysis techniques, with approximately the same level of technical expertise. This will be attractive to operators of fuel fabrication facilities as it will lower costs and improve efficiencies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Plasma Flow Reactor Investigation into Nucleation of Uranium Oxide on Silica Substrates to Better Understand Fallout Formation

Understanding particulate formation in nuclear debris is critical for predicting fallout transport after a nuclear event. Improved characterization of fallout formation and transport could lead to better guidance for emergency response in a post-detonation scenario. By analyzing how U-oxides nucleate onto different forms of SiO 2 (crystalline and amorphous), we can gain insight into how entrained environmental materials may incorporate into fallout. In this experiment a Plasma Flow Reactor (PFR), was used to replicate the high temperature and extreme flow conditions in a nuclear fireball. Uranyl nitrate was injected into the plasma, and PFR-generated Uranium oxides were allowed to condense onto amorphous (nanoparticles of varying sizes) or crystalline (quartz) SiO 2 substrates. SiO 2 substrates were characterized before and after U-oxide deposition using Scanning Electron Microscopy (SEM) based techniques (i.e. EDS) in order to characterize how U-oxides may nucleate onto these substrates. After a collection time of 4 minutes at an RF coil distance of 25cm, both the amorphous SiO 2 nanoparticles and the crystalline structures demonstrated dendritic nucleation of U-oxide species as identified by SEM/EDS.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Spectroscopic investigation of cation effects in U(VI)-NO3- complexation in aqueous solutions

Understanding and manipulating uranyl speciation in aqueous solutions is critical for advancing chemical separation, sensing, and understanding environmental transport of uranyl. We report on the significant enhancement in the complexation of uranyl with nitrate in aqueous solutions containing quaternary ammonium cations leading to the formation of anionic complexes. We base this on the comparative study of the effect of monovalent cations (Na+, Li+, NH4+, and N(CH3)4+) on the complexation equilibria of uranyl-nitrate in solutions, probed by time-resolved laser induced fluorescence spectroscopy (TRLFS). Lifetime-corrected spectra, obtained by extrapolating the time-resolved spectra to t = 0, were used to study speciation to mitigate the effects of variations in the fluorescence lifetimes that depend on extraneous factors such as dynamic quenching. We demonstrate that the lifetime-corrected spectra can be used to determine uranyl speciation in aqueous solutions where the mono-nitrate complex forms, and in acetonitrile with N(CH3)4NO3 where di-nitrate and tris-nitrato species are formed. Aqueous solutions containing N(CH3)4+ are shown to promote the formation of higher complexes of uranyl compared to other inorganic nitrate salts based on the higher redshift in the spectra, poor fits to the two-component model, and the higher apparent formation constants. Comparing the trends in uranyl speciation in the presence of N(CH3)4+ in aqueous and acetonitrile solutions, it is proposed that the quaternary ammonium cations (quats) promote the formation of anionic complexes of uranyl by the ion-association mechanism. These results provide a basis for designing quat-assisted separation systems that target anionic actinide species in aqueous solutions.

Nayak, Srikanth (ORCID:0000000302135796)↗
DOE OSTI2025

Uranyl fluorescence in acidic solution: quenching effects by tetramethylammonium (TMA + )

The quenching of uranyl luminescence by various cation species was studied in aqueous media at low pH. Solutions with different nitrate salts, held at constant uranyl nitrate, nitric acid, and ion concentration, were tested to examine the quenching effects of the cations from the nitrate salts. Alkali metal (Li + , Na + , Rb + ) and quaternary ammonium cations (NH 4 + ), (CH 3 ) 4 N + (TMA + ), (C 2 H 5 ) 4 N + (TEA + ) were investigated. Solutions containing TMA + reduced the lifetime of uranyl fluorescence significantly more than the other cations. Uranyl emission spectra also showed that TMA + increased the complex formation between uranyl and nitrate ions. Fluorescence decay lifetime measurements for most solutions yielded values between 1.4–1.9 μs at 20 °C, while 1.8 M TMA + reduced the lifetime of uranyl fluorescence to 0.6 μs. Decay rate versus concentration data (Stern–Volmer plots) indicated a dynamic quenching process with increasing fluorescence decay rates at higher cation concentrations for Li + , TMA + , and TEA + . The temperature dependencies of the decay rates and the kinetics in D 2 O were also examined.

Persinger, Thomas D. [Argonne National Laboratory ↗