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At least 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

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.

and nuclear chemistry↗
DOE OSTI2016

Petrochemical and Tectonogenesis of Granitoids in the Wuyo-Gubrunde Horst, Northeastern Nigeria: Implication for Uranium Enrichment

The Wuyo-Gubrunde Horst in the northeastern Nigeria consists of migmatite gneiss, unaltered, altered, and sheared porphyritic granites, pegmatites, aplites, basalts, and sandstone. Uranium has been reported in rhyolite, sheared rocks, and sandstone within the area. The petrogenesis of the granitoids and associated rocks in the area was evaluated in the light of new geochemical data, which showed that the U content of altered porphyritic granite is highest and hydrothermal-related. The granitoids are metaluminous, sub-alkaline, and S-type granite, and have evolved by partial melting of crustal material emplaced at moderate depth of 20–30 km in a syn-to late-collisional within-plate tectonic setting. The negative Eu/Eu* anomaly and high (La/Yb){sub N} ratio of the granitoids indicate magma fractionation. The low SiO{sub 2} (<53%) and high Fe{sub 2}O{sub 3} (10%) of the altered porphyritic granite compared to other similar rock units suggest pervasive alteration. The associated basalts are tholeiitic, emplaced within continental plate tectonic setting, and enriched in Ni, V, Nb, Sr, and light rare earth elements, and they have SiO{sub 2}, Fe{sub 2}O{sub 3}, V, Th, and Co contents that are similar to those of the altered porphyritic granites. The U occurrence in the Wuyo-Gubrunde Horst is believed to be sourced from the adjoining Bima sandstone in the Benue Trough, which locally contains carbonaceous zones with anomalously high concentrations of U. The Fe{sup 2+}/Fe{sup 3+} redox fronts formed by alteration of the iron-rich basalts provided the requisite geochemical barrier for U-bearing hydrothermal fluid, causing enrichment of U leached and mobilized from the sandstone through fractures in the rocks.

58 GEOSCIENCES↗
DOE OSTI2020

Reticular exploration of uranium-based metal—organic frameworks with hexacarboxylate building units

We report that the rational reticular design of metal–organic frameworks (MOFs) from building units of known geometries is essential for enriching the diversity of MOF structures. Unexpected and intriguing structures, however, can also arise from subtle changes in the rigidity/length of organic linkers and/or synthetic conditions. Herein, we report three uranium-based MOF structures—i.e., NU-135X (X = 0, 1, 2)—synthesized from trigonal planar uranyl nodes and triptycene-based hexacarboxylate ligands with variable arm lengths. A new chiral 3,6-connected nuc net was observed in NU-1350, while the extended versions of the ligand led to 3-fold catenated MOFs (NU-1351 and NU-1352) with rare 3,6-connected cml-c3 nets. Lastly, the differences in the topology of NU-1350 and NU-1351/NU-1352 could be attributed to the slight distortions of the shorter linker in the former from the ideal trigonal prism geometry to an octahedral geometry when coordinated to the trigonal planar uranyl nodes.

3,6-connected net↗
DOE OSTI2020

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO 2 ) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO 2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO 2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO 2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO 2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO 2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO 2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO 2 . Here, the predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO 2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

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

Effects of calcium and phosphate on uranium(IV) oxidation: Comparison between nanoparticulate uraninite and amorphous U IV –phosphate

The mobility of uranium in subsurface environments depends strongly on its redox state, with U IV phases being significantly less soluble than U VI minerals. This study compares the oxidation kinetics and mechanisms of two potential products of U VI reduction in natural systems, a nanoparticulate UO 2 phase and an amorphous U IV –Ca–PO 4 analog to ningyoite (CaU IV (PO 4 ) 2 ·1–2H 2 O). The valence of U was tracked by X-ray absorption near-edge spectroscopy (XANES), showing similar oxidation rate constants for U IV O 2 and U IV –phosphate in solutions equilibrated with atmospheric O 2 and CO 2 at pH 7.0 (k obs,UO2 = 0.17 ± 0.075 h -1 vs. k obs,U IV PO4 = 0.30 ± 0.25 h -1 ). Addition of up to 400 μM Ca and PO 4 decreased the oxidation rate constant by an order of magnitude for both UO 2 and U IV –phosphate. The intermediates and products of oxidation were tracked by electron microscopy, powder X-ray diffraction (pXRD), and extended X-ray absorption fine-structure spectroscopy (EXAFS). In the absence of Ca or PO 4 , the product of UO 2 oxidation is Na–uranyl oxyhydroxide (under environmentally relevant concentrations of sodium, 15 mM NaClO 4 and low carbonate concentration), resulting in low concentrations of dissolved U VI (<2.5 × 10 -7 M). Oxidation of U IV –phosphate produced a Na-autunite phase (Na 2 (UO 2 )PO 4 ·xH 2 O), resulting in similarly low dissolved U concentrations (<7.3 × 10 -8 M). When Ca and PO 4 are present in the solution, the EXAFS data and the solubility of the UVI phase resulting from oxidation of UO 2 and UIV–phosphate are consistent with the precipitation of Na-autunite. Bicarbonate extractions and Ca K-edge X-ray absorption spectroscopy of oxidized solids indicate the formation of a Ca–UVI–PO 4 layer on the UO 2 surface and suggest a passivation layer mechanism for the decreased rate of UO 2 oxidation in the presence of Ca and PO 4 . Interestingly, the extractions were unable to remove all of the oxidized U from partially oxidized UO 2 solids, suggesting that oxidized U is distributed between the interior of the UO 2 nanoparticles and the labile surface layer. Accounting for the entire pool of oxidized U by XANES is the likely reason for the higher UO 2 oxidation rate constants determined here relative to prior studies. In conclusion, our results suggest that the natural presence or addition of Ca and PO 4 in groundwater could slow the rates of U IV oxidation, but that the rates are still fast enough to cause complete oxidation of U IV within days under fully oxygenated conditions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Synthesis, characterization, and reactivity of the first uranium metallocene 1,2 bis(diphenylphosphino)acetylene complexes.

Two new uranium metallacyclopropenes, (C 5 Me 4 R) 2 U(η 2 -Ph 2 PC=CPPh 2 ) (R = Me, Et) were prepared by reducing the corresponding (C 5 Me 4 R) 2 UCl 2 complexes with KC 8 in the presence of 1,2-bis(diphenylphosphino)acetylene (Ph 2 P–C≡C–PPh 2 ). Both compounds were fully characterized by a combination of elemental analysis and multinuclear NMR, UV–visible–NIR, and IR spectroscopies. Differences in the electronic spectra of these novel compounds and the known (C 5 Me 5 ) 2 U(η 2 -Me 3 SiC=CSiMe 3 ) are discussed. Finally, also presented is the solid-state structure of (C 5 Me 4 Et) 2 U(η 2 -Ph 2 PC=CPPh 2 ), which reveals significant distortions of the coordinated 1,2-bis(diphenylphosphino)acetylene (Ph 2 P–C≡C–PPh 2 ) ligand.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Modelling and analysis of salt-convection effect on oxide reduction process for uranium oxides using smoothed particle hydrodynamics

Oxygen ion transport via molten salt continuum is a key kinetic feature of electrochemical reduction process for uranium oxides in molten salt. During the electrolysis, the evolution of oxygen ion concentration field in molten salt continuum appears from porous metallic layer to open bulk salt. Despite the important role affecting process throughput, its effect coupled with controlled salt movement was not adequately investigated. To capture the reactive oxygen ion transport phenomena, a multidimensional reduction model based on smoothed particle hydrodynamics (SPH) was combined with the advection-diffusion model. A pelletized oxide fuel bed capturing interconnected salt continuum is considered as a reactant material geometry, which allows forced convective delivery of molten salt through the reactant materials with a relatively low pressure head. Here, in this study, reduction time of a pelletized-fuel bed was quantified with respect to key design parameters of the reactant material configuration. The simulation results demonstrated the advantage of salt convection which effectively removes oxygen ion products from the reactant materials and consequently accelerates the reduction process. Moreover, the reduction-time correlation of a single pelletized fuel was developed as a function of relevant dimensionless parameters and was extended to the application for a pelletized-fuel bed. Finally, several insights and implications in the use of a pelletized oxide fuel were presented. The developed SPH-based framework enabled multi-physics analysis of governing dynamics in appreciating and screening candidate reactant material configurations in a cost-effective manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Reactions of atomic thorium and uranium cations with CF 4 studied by guided ion beam tandem mass spectrometry

The fundamental chemistry of the actinides continues to be an area of interest because of their use in nuclear fuel cycles. Here, the kinetic energy dependences of the reactions of thorium and uranium cations with CF 4 were studied using guided ion beam tandem mass spectrometry. The products observed include AnF x + (An = Th or U, x = 1-3) and CF y + (y = 1-3). Although most reactions are exothermic according to literature thermochemistry, all cross sections display strong endothermic features indicating a restriction in the reactivity at low energies. For both actinides, only the AnF + + CF 3 product channel is observed at low energies; however, its formation proceeds very inefficiently. Furthermore, these results indicate that the reactions of Th + and U + with CF 4 are not thermodynamically driven and suggest that the low-energy reactivity results from precise alignment of the reactants.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Why is weapons grade plutonium more hazardous to work with than highly enriched uranium?

Highly Enriched Uranium and Weapons grade plutonium have assumed positions of dominant importance among the actinide elements because of their successful uses as explosive ingredients in nuclear weapons and the place they hold as key materials in the development of industrial use of nuclear power. While most chemists are familiar with the practical interest concerning HEU and WG Pu, fewer know the subtleties among their hazards. In this study, a primer is provided regarding the hazards associated with working with HEU and WG Pu metals and oxides. The care that must be taken to safely handle these materials is emphasized and the extent of the hazards is described. The controls needed to work with HEU and WG Pu metals and oxides are differentiated. Given the choice, one would rather work with HEU metal and oxides than WG Pu metal and oxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Influence of hydrophilic groups and metal-ion adsorption on polymer-chain conformation of amidoxime-based uranium adsorbents

In this study, we focus on the influence of hydrophilic groups and metal-ion loading on adsorbent polymer conformation, which controls access to adsorption sites and may limit adsorption capacity. Gaining a better understanding of the factors that influence conformation may yield higher-capacity adsorbents. Polyamidoxime (PAO), deuterated-PAO polyacrylic acid diblock copolymers (d-PAO-b-PAA), and randomly configured copolymers (PAO-co-PAA) were synthesized and characterized by neutron reflectometry in air and D 2 O. For d-PAO-b-PAA, characterization was also performed after alkali conditioning and in simulated seawater. PAO and PAO-co-PAA, with similar molecular weight and grafting density, extended from 95-Å thickness in air to 180 and 280-Å in D 2 O, respectively. This result suggests that polymer swelling may cause the additional adsorption capacity observed when polymer hydrophilicity increases. Two d-PAO-b-PAA samples, A and B, with a d-PAO thickness of 55-Å swelled to 110-Å and 140-Å, respectively, with an overall thickness increase of ~160% in D 2 O. After alkali conditioning, molecular interactions increased the density of PAA near the PAO-PAA interface, while the d-PAO thickness only decreased by ~10 Å. The d-PAO thickness of both samples declined to ~90-Å after adsorption in simulated seawater due to polymer-chain crosslinking. In conclusion, these results are expected to aid in improving adsorbent synthesis to increase uranium capacity.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Reactive transport of uranium in fractured crystalline rock: Upscaling in time and distance

In this study, batch adsorption and breakthrough column experiments were conducted to evaluate uranium transport through altered material that fills fractures in a granite rock system at the Grimsel Test Site in Switzerland at pH 6.9 and 7.9. The role of adsorption and desorption kinetics was evaluated with reactive transport modeling by comparing one-, two-, and three-site models. Emphasis was placed on describing long desorption tails that are important for upscaling in time and distance. The effect of increasing pH in injection solutions was also evaluated. For pH 6.9, a three-site model with forward rate constants between 0.07 and 0.8 ml g –1 h –1 , reverse rate constants between 0.001 and 0.06 h –1 , and site densities of 1.3, 0.104, and 0.026 μmol g –1 for ‘weak/fast’, ‘strong/slow’, and ‘very strong/very slow’ sites provided the best fits. For pH 7.9, a three-site model with forward rate constants between 0.05 and 0.8 mL g –1 h –1 , reverse rate constants between 0.001 and 0.6 h –1 , and site densities of 1.3, 0.039, and 0.013 μmol g –1 for a ‘weak/fast’, ‘strong/slow’, and ‘very strong/very slow’ sites provided the best fits. Column retardation coefficients (R d ) were 80 for pH 6.9 and 10.3 for pH 7.9. Model parameters determined from the batch and column experiments were used in 50 year large-scale simulations for continuous and pulse injections and indicated that a three-site model is necessary at pH 6.9, although a K d -type equilibrium partition model with one-site was adequate for large scale predictions at pH 7.9. Batch experiments were useful for predicting early breakthrough times in the columns while column experiments helped differentiate the relative importance of sorption sites and desorption rate constants on transport.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Evolving radon diffusion through earthen barriers at uranium waste disposal sites

Field measurements of Rn-222 fluxes from the tops and bottoms of compacted clay radon barriers were used to calculate effective Rn diffusion coefficients (D Rn ) at four uranium waste disposal sites in the western United States to assess cover performance after more than 20 years of service. Values of D Rn ranged from 7.4 × 10 -7 to 6.0 × 10 -9 m 2 /s, averaging 1.42 × 10 -7 . Water saturation (S W ) from soil cores indicated that there was relatively little control of D Rn by S W , especially at higher moisture levels, in contrast to estimates from most steady-state diffusion models. Further, this is attributed to preferential pathways intrinsic to construction of the barriers or to natural process that have developed over time including desiccation cracks, root channels, and insect burrows in the engineered earthen barriers. A modification to some models in which fast and slow pathway D Rn values are partitioned appears to give a good representation of the data; 4% of the fast pathway was needed to fit the data regression. For locations with high S w and highest D Rn (and fluxes) at each site, the proportion of fast pathway ranged from 1.7% to 34%, but for many locations with lower fluxes, little if any fast pathway was needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium–10 wt% molybdenum fuel plate assembly

Residual stresses are expected in monolithic, aluminum clad uranium 10 wt% molybdenum (U–10Mo) nuclear fuel plates because of the large mismatch in thermal expansion between the two bonded materials. The full residual stress tensor of the U–10Mo foil in a fuel plate assembly was mapped with 0.1 mm resolution using high-energy (86 keV) X-ray diffraction. The in-plane stresses in the U–10Mo foil are strongly compressive, roughly -250 MPa in the longitudinal direction and -140 MPa in the transverse direction near the center of the fuel foil. The normal component of the stress is weakly compressive near the center of the foil and tensile near the corner. The disparity in the residual stress between the two in-plane directions far from the edges and the tensile normal stress suggest that plastic deformation in the aluminum cladding during fabrication by hot isostatic pressing also contributes to the residual stress field. A tensile in-plane residual stress is presumed to be present in the aluminum cladding to balance the large in-plane compressive stresses in the U–10Mo fuel foil, but cannot be directly measured with the current technique due to large grain size.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

A thermal desorption spectroscopy study of hydrogen trapping in polycrystalline α-uranium

The kinetics of hydrogen desorption from polycrystalline α-uranium (α-U) was examined using thermal desorption spectroscopy (TDS). The goal was to identify the major trap sites for hydrogen and their associated trap energies. In polycrystalline α-U six TDS adsorption peaks were observed at temperatures of 521 K, 556 K, 607 K, 681 K, 793 K and 905 K. In addition, the desorption was determined to be second order based on peak shape. The position of the first three peaks was consistent with desorption from UH3. To identify the trap site corresponding to the high temperature peaks the data were compared to a plastically deformed sample and a high purity single crystal sample. The plastically deformed sample allowed the identification of trapping at dislocations while the single crystal sample allow for the identification of high angle boundaries and impurities. Thus, with respect to the desorption energy associated with each peak, values between 12.9 and 26.5 kJ/mole were measured.

36 MATERIALS SCIENCE↗
DOE OSTI2017

A modified Embedded-Atom Method interatomic potential for uranium-silicide

Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). U-Si fuels benefit from higher thermal conductivity and higher fissile density compared to UO2. In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multiscale modeling efforts are underway to address this gap in knowledge. In this study, a semi-empirical Modified Embedded Atom Method (MEAM) potential is presented for the description of the U-Si system. The potential accurately describes not only the primary phase of interest (U3Si2), but also a variety of U-Si phases across the composition spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

In-situ neutron diffraction characterization of temperature dependence deformation in α-uranium

In-situ strain neutron diffraction measurements were conducted at temperature on specimens coming from a clock-rolled α-uranium plate, and Elasto-Plastic Self-Consistent (EPSC) modeling was employed to interpret the findings. The modeling revealed that the active slip systems exhibit a thermally activated response, while deformation twinning remains athermal over the temperature ranges explored (25-150 °C). The modeling also allowed assessment of the effects of thermal residual stresses on the mechanical response during compression. These results are consistent with those from a prior study of room-temperature deformation, indicating that the thermal residual stresses strongly influence the internal strain evolution of grain families, as monitored with neutron diffraction, even though accounting for these residual stresses has little effect on the macroscopic flow curve, except in the elasto-plastic transition.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Comparative chemical and structural analyses of two uranium dioxide fuel pellets

Here, a comparative investigation of the trace element, U and Pb isotopic compositions, and structural attributes of two uranium dioxide fuel pellets was conducted for nuclear forensic applications. One fuel pellet consists of natural UO 2 (0.71% 235 U), while the other contains enriched UO 2 (3.98% 235 U). Numerous impurities, including transition metals, lanthanides, and main group elements were detected in both fuel pellets with total trace element contents that vary between 9.77 and 11.37 μg/g. The relative abundances of the transition metal impurities in the natural UO 2 fuel pellet are linked to the source U ore and provide important insights into its provenance. X-ray fluorescence spectroscopy was utilized to investigate and compare the macro-(cm)-scale distribution of impurities within each of the fuel pellets. In addition, structural properties of the fuel pellets were examined using powder X-ray diffraction (PXRD) and Raman and infrared spectroscopy. Differences in the intensity of the T2 g U-O stretching vibrational mode were observed in the Raman spectra of the fuel pellets. Lastly, Pb isotope ratios were investigated to potentially differentiate between the natural and enriched UO 2 samples and are linked to the process history of these materials.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Controlled current-rate AC flash sintering of uranium dioxide

Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol -1 and 380 kJ mol -1 , respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO 2 . Furthermore, both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗