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Effects of Borate and Organics on U(VI) Solubility in WIPP Brine
The solubility of uranium (VI) in Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities. This research was performed by the Los Alamos National Laboratory Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP). The WIPP Actinide Source Term Program (ASTP) did not develop a model for the solubility of actinides in the VI oxidation state. The solubility of UO 2 2+ , in the absence of WIPP specific data, is presently set to be equal to a conservatively high 1 mM within the WIPP Performance Assessment (PA) for all expected WIPP conditions (SOTERM, 2019) as selected at the recommendation of the Environment Protection Agency (EPA) (EPA, 2005). According to the current WIPP chemistry model assumptions and conditions, the expected pC H+ is about 9.5 and controlled by MgO buffering CO 3 2- . The goal of this study is to perform screening experiments that account for the contributions of organics and borate on uranium solubility. In this report, the solubility of U(VI) was determined at pC H+ 9 WIPP brine in the absence or presence of borate and organics at under-saturation approach. Experiments were equilibrated for about 135 days. Organic compounds present in WIPP waste can form strong complexes with actinides and can affect the oxidation states of actinides. The organic compounds addressed in WIPP performance assessment include EDTA (Ethylenediaminetetraacetic Acid), oxalate, citrate, and acetate (SOTERM, 2019). These data quantify the effects of WIPP-relevant concentrations of borate and organics effects on the solubility of U(VI) to challenge the predictions of the WIPP actinide model and inform decisions and recommendations made in the upcoming recertification of the WIPP (CRA-2024). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Experimental Strategy to Challenge Actinide Solubility Predictions” and designated LCO-ACP-26. All data reported were obtained under the LANL-CO Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD) (QAPD, 2017).
Ra-226 extraction and purification from simulated uranium tailings
Growing demand for Ac-225 has outstripped current production, increasing interest in alternative production routes. Furthermore, these often rely on irradiations of Ra-226, yet Ra-226 supplies are also limited. Uranium tailings provide an abundant source of Ra-226, though separation from these complex matrices remains challenging. A laboratory-scale process was developed to recover Ra-226 from simulated U-tailings. After acid leaching, Ra remained adsorbed on refractory oxides and silicates. An additional CaCl₂ leach of the residue raised recovery to ~ 64%. Subsequent purifications, including precipitation, ion-exchange, and extraction chromatography (TK101), achieved chemical yields up to 73 ± 7%, producing 7.68 ± 0.1 µCi of high-purity Ra-226.
Exploring the Complex Chemistry of Uranium within Molten Chloride Salts
Harsh environments represent a unique opportunity to explore new frontiers in chemistry while developing novel tools to meet global needs. Exploring the chemistry of uranium within molten salts is a key example. Actinide chemistry within the highly ionic environment of a molten salt is poorly understood, particularly in the presence of common salt impurities or without active oxidation state control. Delving into this chemistry can provide new insight into actinide and f-electron interactions. Furthermore, expanding our chemical knowledge can also enable advances in and deployment of molten salt reactors or molten salt recycle schemes. Both molten salt applications aim towards providing green, reliable, and equitable energy as well as critical materials for the world. Here the utilization of visible absorbance and Raman spectroscopies to understand and quantify U within chloride-based salt eutectics is discussed. Furthermore, machine learning techniques in the form of chemometric modelling are developed and described, providing advanced analytical tools to quantify and characterize the U present. In conclusion, these tools are then leveraged to monitor and explore the dynamic fundamental chemistry of U within chloride-based salt melts.
Stabilization of Photocatalytically Active Uranyl Species in a Uranyl–Organic Framework for Heterogeneous Alkane Fluorination Driven by Visible Light
When photoactivated, the uranyl ion is a robust oxidant capable of abstracting hydrogen atoms from nonactivated C–H bonds. Yet, the highly reactive singly reduced [U V O 2 ] + intermediate is unstable with respect to disproportionation to the uranyl dication and insoluble tetravalent uranium phases, which limits the usage of uranyl ions as robust photocatalysts. Herein, we demonstrate that photoactivated uranyl ions can be stabilized by immobilizing and separating them spatially in a uranyl–organic framework heterogeneous catalyst, NU-1301. The visible-light-photoactivated uranyl ions in NU-1301 demonstrated longer-lived U(V) and radicals than those in homogeneous counterparts, as evidenced by X-ray photoelectron spectroscopy and time-dependent electron paramagnetic resonance, leading to higher turnovers and enhanced stability for the fluorination of nonactivated alkanes.
MURR LEU structural and thermal hydraulics analyses: Part I – Preliminary irradiation thermo-mechanical behavior
The University of Missouri Research Reactor (MURR) is expected to be converted from highly enriched uranium (HEU, ≥ 20 wt% U-235) U-Al x dispersion fuel to low-enriched uranium (LEU, < 20 wt% U-235) with U-10Mo monolithic fuel. This work introduces high-fidelity irradiation thermo–mechanical (T-M) analysis of the MURR LEU focusing on changes in coolant channel gap thickness. Three-dimensional (3D) finite element (FE) models were developed to simulate the irradiation T-M behavior of the MURR LEU element with all 23 curved fuel plates, the two side plates, and the combs. It was shown that channel gap thickness changes were influenced not only by plate thickness variations due to fuel swelling and creep but also by the radial displacement of consecutive MURR LEU plates. Modeling the fuel element assembly captured side plate displacements, which were shown to reduce radial fuel plate displacements towards the convex side. The maximum local radial displacement in the element was predicted at the end of life (EOL) as 23.7 mil (602.0 µm) on the lateral centerline of plate 23 towards the convex side. The maximum stripe-averaged reduction in channel gap thickness, particularly relevant for thermal hydraulics (TH) safety analysis, was calculated as 15.9 mil (403.9 µm) in single-side heated channel 24 (the outermost channel). These results account for the thermal resistance from the oxide build-up on cladding surfaces which was shown to be up to 0.82 mil (20.8 µm) thick. It was demonstrated that accounting for oxide layer thermal resistance led to a 10 °C higher peak fuel temperature and a 4.4 mil (111.8 µm) greater maximum local radial displacement. The impact of the calculated channel gap thickness changes on the MURR LEU TH safety analysis is evaluated in Part II.
Alternate Methods for Cleaning Zirconium Plate
The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.
Comparison of Industrialized Late 20th Century Flowsheets for Reprocessing Used Nuclear Fuel
This study identified and compared three flowsheets for reprocessing used nuclear fuel (UNF) industrialized in plants in the United Kingdom (the Thermal Oxide Reprocessing Plant), France (UP2-800/UP3) and Japan (the Rokkasho Reprocessing Plant). The study also identified the major implications for a plant in the United States if it were initiated. All flowsheets employed the established Plutonium Uranium Reduction Extraction (PUREX) solvent extraction technology to separate uranium and plutonium from UNF dissolved in nitric acid. However, differences in the approaches to managing iodine-129, tritium and technetium were identified in the flowsheets. A US plant would also need to separate krypton-85 as well as iodine-129 and tritium for immobilization and disposal.
Degree of Weathering of H-Chondrites from Frontier Mountain, Antarctica
One of the factors that determines the survival time of meteorites on the Earth's surface is the rate of weathering. For meteorites from hot deserts, a clear correlation is found between the degree of weathering and the terrestrial age, but for Antarctic meteorites this correlation is weak or even lacking. The lack of a clear correlation can partly be attributed to the two-stage history of many Antarctic meteorites, which spend part of their terrestrial residence time in the ice before they are exposed on the ice. Recently, it was found that for Lewis Cliff (LEW) meteorites local conditions on the ice play an important role in the weathering process. This work focuses on weathering effects in ordinary chondrites from Frontier Mountain (FRO), North Victoria Land. Although most FRO meteorites were classified as weathering category A or B, many are contaminated with terrestrial uranium, deposited from meltwater. This suggests that weathering plays a more significant role than the qualitative A-B-C weathering index indicates. We therefore determined the degree of weathering more quantitatively, by deriving the amount of oxidized metal from the concentrations of Fe and Ni in the nonmagnetic fraction of 23 H-chondrites and 1 L-chondrite. The results will be compared with those of LEW meteorites and will be discussed in terms of terrestrial age and location of find on the ice.
Degree of Weathering of H-Chondrites From Frontier Mountain, Antarctica
One of the factors that determines the survival time of meteorites on the Earth's surface is the rate of weathering. For meteorites from hot deserts, a clear correlation is found between the degree of weathering, and the terrestrial age, but for Antarctic meteorites this correlation is weak or even lacking. The lack of a clear correlation can partly be attributed to the two-stage history of many Antarctic meteorites, which spend part of their terrestrial residence time in the ice before they are exposed on the ice. Recently, it was found that for Lewis Cliff (LEW) meteorites local conditions on the ice play an important role in the weathering process. This work focuses on weathering effects in ordinary chondrites from Frontier Mountain (FRO), North Victoria Land. Although most FRO meteorites were classified as weathering category A or B, many are contaminated with terrestrial uranium, deposited from meltwater. This suggests that weathering plays a more significant role than the qualitative A-B-C weathering index indicates. We therefore determined the degree of weathering more quantitatively, by deriving the amount of oxidized metal from the concentrations of Fe and Ni in the nonmagnetic fraction of 23 H-chondrites and 1 L-chondrite The results will be compared with those of LEW meteorites and will be discussed in terms of terrestrial age and location of find on the ice.
Analysis of neptunium oxides produced through modified direct denitration
Production of neptunium-237 ( 237 Np) target materials for plutonium-238 ( 238 Pu) radioisotope thermoelectric generators (RTGs) for deep space exploration requires advanced chemistry and engineering development. Currently, the domestic Pu-238 Supply Program at Oak Ridge National Laboratory produces neptunium dioxide (NpO 2 ) for target material using a modified direct denitration (MDD) flowsheet. Although the chemistry, reaction mechanisms, and product characteristics of MDD are well understood for uranium, corresponding studies of the neptunium system are still needed to continue optimization of target material properties, production equipment design, and production flowsheets. Here, the objective of this work is to characterize crystalline phases, morphology, surface texture, and particle size of NpO 2 produced via MDD reactions. Solid-phase characterization techniques, including powder X-ray diffraction (pXRD) and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS), were employed to achieve this objective. Subsequent data processing using the Morphological Analysis for Material Attribution (MAMA) software was performed to analyze particle morphology and size. Broadly, the powders were found to contain a mixture of NpO 2 and Np 2 O 5 after denitration with a variety of morphologies. After high-firing, the product was found to be NpO 2 with a typical polycrystalline oxide morphology and a grain size ranging from 0.72 to 0.94 µm. These analyses provide knowledge on the reaction pathway for a non-traditional NpO 2 synthesis method and offer additional unique insight into production-scale environments for transuranic materials.
Uranium sorption to organic matter and long-term accumulation in a pristine alpine wetland
Understanding the controls on uranium (U) mobility in the environment is key to improve the management of sites contaminated by U mining activities. Previous research has shown that natural or engineered wetlands are particularly able to scavenge high amounts of U(VI) and U(IV) under noncrystalline forms. However, questions remain on the respective roles of sorption and reduction processes in the removal of U from running waters in wetlands, as well as on the long-term stability of U storage. Here, we performed a series of geochemical, isotopic ($δ$ 238 U, ( 234 U/ 238 U)), microscopic (SEM-EDXS, EPMA) and spectroscopic (µ-XRF, µ-XAS, XANES and EXAFS at the U L 3 and M 4 -edges and Fe K-edge) investigations to determine the modes of U accumulation and assess U mobility in a natural exceptionally U-enriched (up to 5000 µg/g) wetland on the shore of Lake Nègre (Mediterranean Alps, France). Uranium (VI) was largely dominant in the two studied soil cores, except a few samples containing as much as ~50 % U(IV). At the particle scale, U is associated to a variety of organic constituents of the soil matrix with a homogenous oxidation state. Bulk EXAFS spectroscopy at the U L 3 -edge shows that U is mostly mononuclear, with dominant monodentate binding to organic moieties (C neighbors at ~3.45 Å). An additional minor fraction of U under polymeric forms is inferred from wavelet (CCWT) analysis of the EXAFS data. These observations are reinforced by 1 M bicarbonate extractions that result in the dissolution of 82–96 % of total U, including putative polymeric species. At the wetland scale, similar or slightly fractionated isotopic ratios ($δ$ 238 U) between the wetland-feeding creek waters and the wetland soils are observed, supporting the idea that U(VI) sorption on organic matter is the primary U scavenging mechanism. Furthermore, it indicates that partial U(VI) reduction to U(IV) occurs as a second step, after sorption. Analysis of U decay chain disequilibria in the cores as a function of depth suggests that U accumulation in this wetland has lasted for several thousand years. In conclusion, we propose that the wetland acts as an active reactor where U has been massively accumulating for ~14500 years, especially as U(VI) forms associated to organic matter and is further partly exported to the lake through soil erosion.
Radiolytically reworked Archean organic matter in a habitable deep ancient high-temperature brine
Abstract Investigations of abiotic and biotic contributions to dissolved organic carbon (DOC) are required to constrain microbial habitability in continental subsurface fluids. Here we investigate a large (101–283 mg C/L) DOC pool in an ancient (>1Ga), high temperature (45–55 °C), low biomass (10 2 −10 4 cells/mL), and deep (3.2 km) brine from an uranium-enriched South African gold mine. Excitation-emission matrices (EEMs), negative electrospray ionization (–ESI) 21 tesla Fourier-transform ion cyclotron resonance mass spectrometry (FT-ICR MS), and amino acid analyses suggest the brine DOC is primarily radiolytically oxidized kerogen-rich shales or reefs, methane and ethane, with trace amounts of C 3 –C 6 hydrocarbons and organic sulfides. δ 2 H and δ 13 C of C 1 –C 3 hydrocarbons are consistent with abiotic origins. These findings suggest water-rock processes control redox and C cycling, helping support a meagre, slow biosphere over geologic time. A radiolytic-driven, habitable brine may signal similar settings are good targets in the search for life beyond Earth.
UFCC Calibration Measurements for UO2 Samples
The uranium UF 6 coincidence counter (UFCC) was designed to measure 5A/B cylinders containing UF6 in the passive neutron self-interrogation mode. The purpose of the present report is to extend the UFCC capability to other sample types such as oxides and liquids that might be found during verification activities. A 252 Cf source was introduced into the detector to make the active mode measurements possible, because only the UF 6 samples have adequate neutron emission for self-interrogation. A customized HDPE insert was designed to position the 252 Cf source under the UO 2 sample cans. The sample enrichments varied from depleted to 93% to establish the shape of the calibration curve and the measurement uncertainties.
Advanced Fuel Cycle Cost Basis Report: Supporting Document 7 Presentation: Du and RU Disposal Costs [Slides]
Worldwide DU and RU in various chemical forms are some of the largest legacy products of the nuclear industry (defense & power) in both mass and volume. Chemical forms include U metal or alloy, UF6, UO2, UO3, U3O8, and UF4. Most of this material is now in above ground storage. Due to chemical stability and low water solubility oxides are the preferred form for safe storage and ultimate disposal. Long term U disposal in large amounts presents more of a potential radioactivity problem than its conversion and temporary storage. Now a near term problem with freshly-mined uranium ore, radon emanation will eventually be a long-term problem for both DU and RU dispositioned in large quantities at a specific location.
Selection and Characterization of Surrogate Material for Developing Zr Barrier Coatings for Uranium Nitride Particles
In support of UC 1-x N x fuel kernels (~ 800 micrometer diameter) development work ongoing at Oak Ridge national laboratory (ORNL), where the fuel kernels will be embedded inside a possible zirconium based matrix, Argonne National Laboratory (ANL) is contributing to help develop a robust metal diffusion barrier coating. Zirconium nitride (ZrN) and Zirconium oxide (ZrO2) were selected as possible diffusion barrier candidates, and the atomic layer deposition (ALD) has been chosen as a candidate technology to deposit the film. Development of such a barrier coating with the ALD process entails significant number of trials to study relevant parameters. Therefore, to prevent needless wastage, instead of coating directly over fuel kernels, a surrogate candidate with similar dimensions and physical properties were been selected and coated to develop the diffusion barrier coating.
UFCC Calibration Measurements for Bulk U 3 O 8 Powder Samples
The uranium UF 6 coincidence counter (UFCC) was designed to measure 5A/B cylinders containing UF 6 in the passive neutron self-interrogation mode. The purpose of the present report is to extend the UFCC capability to other sample types such as oxides and liquids that might be found during verification activities. A 252 Cf source was introduced into the detector to make the active mode measurements possible, because only the UF 6 samples have adequate neutron emission for self-interrogation. The sample enrichments varied from 0.72% to 91% to establish the shape of the calibration curve and the measurement uncertainties. The measurements included the singles, doubles, and triples rates. The time correlations between the 252 Cf source and the induced fission reactions reduced the statistical error in the doubles and triples rates to be small enough for useful measurements in 5-10 minutes.
Formation of Clustered DNA Damage after High-LET Irradiation: A Review
Radiation can cause as well as cure cancer. The risk of developing radiation-induced cancer has traditionally been estimated from cancer incidence among survivors of the atomic bombs in Hiroshima and Nagasaki. These data provide the best estimate of human cancer risk over the dose range for low linear energy transfer (LET) radiations, such as X- or gamma-rays. The situation of estimating the real biological effects becomes even more difficult in the case of high LET particles encountered in space or as the result of domestic exposure to particles from radon gas emitters or other radioactive emitters like uranium-238. Complex DNA damage, i.e., the signature of high-LET radiations comprises by closely spaced DNA lesions forming a cluster of DNA damage. The two basic groups of complex DNA damage are double strand breaks (DSBs) and non-DSB oxidative clustered DNA lesions (OCDL). Theoretical analysis and experimental evidence suggest there is increased complexity and severity of complex DNA damage with increasing LET (linear energy transfer) and a high mutagenic or carcinogenic potential. Data available on the formation of clustered DNA damage (DSBs and OCDL) by high-LET radiations are often controversial suggesting a variable response to dose and type of radiation. The chemical nature and cellular repair mechanisms of complex DNA damage have been much less characterized than those of isolated DNA lesions like an oxidized base or a single strand break especially in the case of high-LET radiation. This review will focus on the induction of clustered DNA damage by high-LET radiations presenting the earlier and recent relative data.