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At least 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Discovery of davemaoite, CaSiO 3 -perovskite, as a mineral from the lower mantle

Calcium silicate perovskite, CaSiO 3 , is arguably the most geochemically important phase in the lower mantle, because it concentrates elements that are incompatible in the upper mantle, including the heat-generating elements thorium and uranium, which have half-lives longer than the geologic history of Earth. We report CaSiO 3 -perovskite as an approved mineral (IMA2020-012a) with the name davemaoite. The natural specimen of davemaoite proves the existence of compositional heterogeneity within the lower mantle. Furthermore, our observations indicate that davemaoite also hosts potassium in addition to uranium and thorium in its structure. Hence, the regional and global abundances of davemaoite influence the heat budget of the deep mantle, where the mineral is thermodynamically stable.

58 GEOSCIENCES↗
DOE OSTI2022

Thermal Neutron Scattering Law Evaluation for Zirconium Carbide and Critical Mass Calculations

Zirconium carbide (ZrC) is a candidate material for use in advanced high temperature reactors, including space nuclear thermal propulsion applications. Thermal neutron scattering laws (TSLs) are generated for carbon bound in ZrC, C(ZrC), and zirconium bound in ZrC, Zr(ZrC), using ab initio lattice dynamics methods. These evaluations are to be submitted for inclusion in ENDF/B-VIII.1 and use the incoherent approximation for inelastic scattering as well as the new mixed elastic scattering treatment. The application of disordered alloy theory is introduced to appropriately capture the isotopic composition of Zr and C in the elastic scattering cross section. Localized higher energy vibrations in the C(ZrC) phonon density of states that are separated from lower energy modes result in quantized oscillations in the inelastic contributions to the TSL with a significant likelihood of large energy down-scattering and up-scattering interactions, where the latter increases in probability with temperature. The quanta of energy transfer during neutron thermalization is substantially greater than classically expected within the thermal neutron energy range. MC21 critical mass calculations of ZrC mixtures with high-enriched uranium demonstrate an impact of the TSLs when compared to free-gas treatment for 235 U concentrations less than 0.2 g/cm 3 . Additional MC21 critical mass calculations with homogenous mixtures of ZrC and reactor-grade graphite also demonstrate sensitivity to the ZrC TSL for thermal spectrum driven fission systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Gandolinium poison solubility testing for the downstream impacts from accelerated basin de-inventory

The Accelerated Basin De-inventory (ABD) Program at the Savannah River Site (SRS) is designed to accelerate the de-inventory of L-Basin and accelerate the Spent Nuclear Fuel (SNF) Disposition mission. Spent fuel will be dissolved in H-Canyon without recovery of uranium. The dissolver solutions will be temporarily stored, pH-adjusted to excess hydroxide (which will facilitate precipitation of metal oxides/hydroxides), transferred to the Concentration, Storage, and Transfer Facility (CSTF), and subsequently immobilized in the Defense Waste Processing Facility (DWPF) during planned sludge batch campaigns. ABD accelerates basin closure, significantly reduces programmatic risk, and greatly reduces the lifecycle budget requirements for the site by eliminating the need for a SNF drying and packaging capability. The ABD approach represents a significant change to the clean-up approach for the SRS. However, the increased fissile loading in sludge batches, due to the dissolver solutions, requires investigation to ensure fissile limits are efficiently and safely managed; higher fissile loadings in the glass are projected to be two to three times higher than the current fissile concentration limit of 897 g/m 3 and will be addressed in a future report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Flowsheet for the Neutralization of Accelerated Basin De-inventory (ABD) Material

Under the Accelerated Basin De-inventory (ABD) program H-Canyon will be dissolving aluminum spent nuclear fuel (ASNF) and then neutralizing that solution without performing head-end strike or uranium recovery operations. After dissolution in H-Canyon the material will be neutralized to a free hydroxide concentration of either 0.6 M or 1.2 M to meet the waste acceptance criteria for transfer to the SRS Concentration, Storage and Transfer Facility (CSTF). In order to ensure the material could be successfully neutralized and transferred, SRNL completed experiments with simulated H-Canyon dissolver solutions. Two bounding simulants were developed based on the expected compositions of ASNF to be dissolved, one containing only Gd as the neutron poison and one containing Fe as an additional poison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Synthesis of UO 2 nanoparticles via coulometric titration: influence of electrolytes and ligands on synthesized particle properties

Here, this study investigates the controlled synthesis of uranium oxide nanoparticles (UO 2 NPs) via coulometric titration under ambient conditions, focusing on the impact of electrolyte anions, salt concentrations, and strongly complexing ligands on particle formation and properties. Using dilute solutions of mineral acids (HNO 3 , HCl, and HClO 4 ), we demonstrate that the choice of electrolyte anion affects particle size, polydispersity, and surface charge. Particle characterization using dynamic light scattering and transmission electron microscopy shows that particles synthesized in perchlorate are largest and show a high degree of polydispersity. In comparison, particles synthesized in chloride are smaller and more uniform in size. Synthesis in nitrate yields a wide variety of particle sizes, with the major fraction (>65 %) having a smaller size than that obtained in the other two electrolytes. The presence of more strongly coordinating ligands such as sulfate and acetate modulate hydrolysis and condensation reactions, with sulfate suppressing nanoparticle formation across a wide concentration range ([SO 4 2- ] > 5 mM) and acetate enabling stable colloidal suspensions at [HOAc] ≤ 0.1 M. Synthesis in concentrated electrolytes (2 M NaNO 3 , NaCl, or NaClO 4 ) accelerates reaction kinetics but introduces challenges, as particles showed increased polydispersity and aggregation, and were more prone to oxidation. Electrolyte effects on actinide oxide nanoparticle formation are discussed and a short comparison to established nanoparticle syntheses is drawn. This work underscores the importance of tailoring synthesis parameters to achieve desired nanoparticle properties, providing valuable insights for optimizing UO 2 NP production for various applications.

Actinide hydrolysis↗
DOE OSTI2026

Uranium Hexafluoride Hydrolysis Reaction Dynamics from Cryogenic Layering, FTIR Spectroscopy, and Isotopic Substitution

The first direct evidence that the hydrolysis reaction of uranium hexafluoride (UF 6 ) follows multiple competing pathways which are driven by the ratio of water to UF 6 , temperature, and isotopic composition is presented. Using temperature dependent infrared spectroscopy, it is shown the hydrolysis can be prevented at temperatures below 150 K, and that water-rich environments promote the formation of uranium oxyfluoride intermediates. Spectral shifts reveal isomeric transitions and the growth of polymeric species, with reaction reversibility observed at high water concentrations. Additionally, controlled heating rates affect the emergence of intermediates. The final particulate product consistently forms as uranyl fluoride hydrate, though its morphology and spectral signature vary with reaction conditions and annealing. These findings help clarify long-standing uncertainties surrounding UF 6 hydrolysis.

Chemical reactions↗
NASA NTRS1966

Stabilization of Tungsten-Uranium Dioxide Composites Under Thermal Cycling Conditions

Uranium losses during thermal cycling of tungsten - uranium dioxide composites to 2500° C in flowing purified hydrogen for specimens initially containing 35 volume percent uranium dioxide were reduced by the use of thorium dioxide, calcium oxide, or yttrium oxide as additives in solid solution with the uranium dioxide. The effectiveness of the additives decreased in the order yttrium oxide, calcium oxide, and thorium dioxide. Stabilization of uranium dioxide by calcium oxide or yttrium oxide can be explained in part on the basis of the introduction of oxygen vacancies into the fluorite lattice and the associated lowered partial molar free energy of oxygen. The difference in the effectiveness of calcium oxide and yttrium oxide is discussed, and a possible explanation of the observed behavior is proposed. Of the concentrations of yttrium oxide tested (i.e., 2. 5, 5, and 10 mole percent), 10 mole percent of yttrium oxide was the most effective in pre­ venting loss of uranium from composities.

Gluyas, Richard E.↗
DOE OSTI2022

Restoration Insights Gained from a Field Deployment of Dithionite and Acetate at a Uranium In Situ Recovery Mine

Mining uranium by in situ recovery (ISR) typically involves injecting an oxidant and a complexing agent to mobilize and extract uranium in a saturated ore zone. This strategy involves less infrastructure and invasive techniques than traditional mining, but ISR often results in persistently elevated concentrations of U and other contaminants of concern in groundwater after mining. These concentrations may remain elevated for an extended period without remediation. Here, we describe a field experiment at an ISR facility in which both a chemical reductant (sodium dithionite) and a biostimulant (sodium acetate) were sequentially introduced into a previously mined ore zone in an attempt to establish reducing geochemical conditions that, in principle, should decrease and stabilize aqueous U concentrations. While several lines of evidence indicated that reducing conditions were established, U concentrations did not decrease, and in fact increased after the amendment deployments. We discuss likely reasons for this behavior, and we also discuss how the results provide insights into improvements that could be made to the restoration process to benefit from the seemingly detrimental behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Minimum Detectable Intakes and Doses for Uranium Bioassays—Comparison between Alpha Spectrometry and ICP-MS

Naturally occurring uranium complicates monitoring for occupational exposures. There are several retroactive methods that can be used to monitor for occupational exposures, with benefits and drawbacks to each. Analysis of uranium in urine by mass spectrometry and alpha spectrometry is compared, and methods of determining an occupational exposure are presented. Furthermore, the minimum detectable concentrations from each analysis and a method for intake determination based on the analytical results are compared for various solubility types and mixtures. Mass spectrometry with radiochemical separation was found to be the most sensitive analysis for detecting occupational exposures to anthropogenic mixtures based on minimum detectable doses calculated from the proposed method for intake determination.

bioassay↗
NASA NTRS1985

Uranium distribution and 'excessive' U-He ages in iron meteoritic troilite

Fission tracking techniques were used to measure the uranium distribution in meteoritic troilite and graphite. The obtained fission tracking data showed a heterogeneous distribution of tracks with a significant portion of track density present in the form of uranium clusters at least 10 microns in size. The matrix containing the clusters was also heterogeneous in composition with U concentrations of about 0.2-4.7 ppb. U/He ages could not be estimated on the basis of the heterogeneous U distributions, so previously reported estimates of U/He ages in the presolar range are probably invalid.

Fisher, D. E.↗
DOE OSTI2023

Part II: Predicting performance of $\mathrm{DOWEX 21K}$ resin for remediation of comingled contaminants in groundwater

The selectivity of ion exchange (IX) resins for aqueous contaminant removal can be impacted by changing concentrations of competing natural groundwater ions. In a two-part investigation, the Hanford Site 200 West Area pump-and-treat (P&T) facility in Washington State, USA is used as a case study to evaluate the performance of two IX resins for groundwater treatment: Purolite® A532E for pertechnetate (TcO 4 - ) removal, explored in Part I, and DOWEX 21K (DOWEX) for uranium (U) removal. In Part II, DOWEX selectivity for U, as uranyl carbonate species, and uptake kinetics is quantified in a series of laboratory-scale aqueous batch experiments containing Hanford-relevant concentrations of competing anions nitrate (NO 3 - ), sulfate (SO 4 2- ), chloride (Cl - ), and carbonate (CO 3 2- ), as well as co-mingled contaminant TcO 4 - . Here the results demonstrate that DOWEX trimethylammonium functional groups are highly selective for U carbonate species (85–100 % uptake) under all conditions investigated. Only NO 3 - concentrations of 100 mM were shown to decrease U removal, with the extent (85–99 %) depending on competing anion concentrations present in solution. However, at the highest NO 3 - concentrations reported for groundwaters treated at the P&T facility (25 mM), the effect on U uptake is minimal. The batch sorption results are modeled to obtain chloride normalized equilibrium exchange coefficients (K) for predicting DOWEX performance: K SO4--/Cl- = 2.0, K NO3-/Cl- = 5.0, K HCO3-/Cl- = 1.5, K TcO4-/Cl- = 2,000, and K U/Cl- = 50,000. These K values predict little effect of current and future influent chemistries on U removal by DOWEX, where both uranyl carbonate species and TcO 4 - are removed such that effluent concentrations meet groundwater treatment requirements.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Characterizing major and trace element compositions in fallout melt glass from a near-surface nuclear test

We report the chemical and isotopic compositions of fallout melt glasses from nuclear tests contain a range of information constraining the physical conditions within the fireball and the mechanisms of fallout formation but historic studies tended to exclude the behavior of stable major and trace elements. Here, we present a large study specifically focused on major and trace element relationships within a population of macroscale fallout samples from a single event. We interpret these data to better constrain how fallout melt glass formation in near surface environments is influenced by that environment and demonstrate how major and trace element abundances can provide useful insights into chemical processes within the fireball. Data confirm that the uranium in the fallout glass population derives from two isotopically distinct endmembers: isotopically enriched uranium (presumably from the weapon), and natural composition uranium that may be a combination of anthropogenic and environmental materials from within the blast zone. The similarity between major and trace element concentrations in fallout and corresponding local soils from the event site confirm the local soils as the most probable source of entrained material into the fireball and the source of carrier material into which the bomb vapor was incorporated. The lack of correlation between major and trace element abundances with size indicates that volatility driven processes, such as condensation from the fireball, do not control the composition of macroscale fallout melt glass. Although the fallout has major and trace element chemical characteristics broadly similar to those of the local, associated soils, some systematic differences are observed between the two populations. Fallout melt glass is depleted in volatile elements such as K, Na, Tl and Pb, consistent with heating to temperatures above ~1000 °C for 3-10 s. This is supported by the results of laser heating experiments performed on rhyolitic soil at temperatures (1600-2200 °C) and timescales (1-120 s) that are broadly relevant to fallout formation conditions. Relative enrichments of metals such as Cu and Co do not correlate with the abundance of uranium, suggesting that fallout also records input of near field anthropogenic materials. Our observations suggest that major chemical features can be related to processing in the fireball and used to inform the thermal-chemical evolution of the system. Ultimately, these data are consistent with a fallout formation mechanism that involves rapid melting of surface materials to form carrier material melts with minor incorporation of bomb vapor and a degree of volumetric volatile loss due to heating.

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

Probing air-water interfaces of dibutyl phosphoric acid (HDBP) aqueous solutions using vibrational sum frequency generation (vSFG) spectroscopy

Liquid-liquid extraction is a separation technique implemented in a wide variety of areas, achieving particular success in both the nuclear and biomedical fields. In this work, vibrational sum frequency generation spectroscopy (VSFG) and surface tension measurements were used to investigate the adsorption of dibutyl phosphate (DBP) at air-aqueous interfaces to simulate liquid-liquid systems relevant to the Plutonium Uranium Redox Extraction (PUREX) Process. The objective of this work is to establish qualitative relationships between changes in the bulk aqueous phase concentrations of DBP and its concentration and structure at air-liquid interface as probed with VSFG. Nitric acid concentration and solution ionic strength were varied to examine their effect on the interfacial DBP.. Introduction of DBP into neat water resulted in reduction of the VSFG spectral intensity in the dangling O-H region (3680 – 3800 cm -1 ) but large increase in the H-bonded O-H stretch frequency region (3000 – 3500 cm-1) and the appearance of the CH 3 symmetric stretch and CH 3 Fermi resonance peaks at ~ 2880 and 2945 cm -1 , respectively, indicating DBP at the air-water interface. The intensity of the C-H strecth peaks increased as DBP concentration increased from 0.24 to 32 mM, accompanied by a decreasing surface tension values. At fixed DBP concentration, the addition of either or both of HNO 3 and NaNO 3 to an ionic strength of 1 M or 3 M led to significant reduction of the O-H VSFG peaks and enhancement of the C-H peaks. The origins of these experimental observations are attributed to both the increased HDBP molecules partitioning and adsorption to the interface and the protonation of the interfacial DBP- molecules.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
NASA NTRS2020

Toward an In-Depth Material Model for Cermet Nuclear Thermal Rocket Fuel Elements

The development and qualification of nuclear thermal propulsion (NTP) fuel element technologies would be aided by an in-depth model of material response and failure modes at operating conditions. Integrated computational materials engineering techniques have the potential to provide such a model, as demonstrated here through three case studies focused on a tungsten-uranium mononitride cermet fuel. The first case focuses on the erosion of tungsten (W, also named wolfram), a nominal coating/cladding material, in hot hydrogen. Ab initio techniques are used to calculate erosion rates and thermal expansion at NTP operating conditions. The second focuses on the stability of uranium mononitride (UN) fuels at high temperature and in the presence of hydrogen. Phase diagram techniques reveal potential instabilities and decomposition pathways at high hydrogen concentrations. The third focuses on using microstructure information to predict high temperature mechanical response and failure of tungsten, used in refractory cermet materials. Combined finite element and discrete dislocation dynamics techniques provide mechanical properties in agreement with experimental methods. The integration of these techniques for an all-encompassing material model is discussed.

cermet↗
DOE OSTI2023

Part I: Predicting performance of Purolite A532E resins for remediation of comingled contaminants in groundwater

Ion exchange (IX) resins are used in pump-and-treat (P&T) facilities to remove soluble groundwater contaminants. However, natural anions present at concentrations orders of magnitude higher than contaminants can compete for IX sites and impact resin lifecycles. Here, the Hanford Site’s 200 West Area P&T facility (Washington State, USA) was selected as a case study because it currently uses two IX resins: Purolite® A532E (A532E) to remove pertechnetate (TcO 4 - ) and DOWEX 21K (DOWEX) to remove uranium from groundwater. Nitrate (NO 3 - ), sulfate (SO 4 2- ), chloride (Cl - ), and carbonate (CO 3 2- ) anions have been identified to potentially compete for A532E and DOWEX IX sites. Hanford-relevant anion groundwater concentrations were used to design a series of laboratory-scale batch experiments to evaluate the impact of competing anions on resin performance and potential kinetic effects. These data are then modeled to obtain Cl--normalized equilibrium exchange coefficients (K) needed to predict IX resin performance. The work is presented in two parts, with IX performance evaluated for A532E in Part I and DOWEX in Part II. Part I results demonstrate that TcO 4 - uptake is not impacted by NO 3 - , SO 4 2- , Cl - , CO 3 2- (as HCO 3 - ) and U(VI) carbonate anions, with K TcO4-/Cl- > 4,000, likely due to the high selectivity of A532E trihexylammonium sites for the large, weakly hydrated TcO 4 - anion. Other anion K values were K NO3-/Cl- = 20, K SO4--/Cl- = 0.2, K HCO3-/Cl- = 0.09, K U/Cl- = 370–1000. These K values provide conservative parameters for predicting A532E performance, and demonstrate that, under these test conditions, A532E will remove TcO 4 - from current and future influent streams to meet groundwater treatment objectives.

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

Trace Element Analyses of Micron-Size Particles and Statistical Determination of Minimum Detection Limits

Savannah River National Laboratory (SRNL) has developed expertise in producing homogeneous, ca. 1 m-diameter spherical particles of mixed-element components, wherein dopants can be varied from a trace constituent (ppm) to wt.% concentrations. The samples used for this work are nickel-doped cerium oxide microspheres produced by SRNL. They were initially selected as analogs for plutonium-doped uranium oxide particles and analyzed as part of a larger study to evaluate whether electron probe microanalyzers (EPMA) can be used to characterize nuclear materials as an alternative or complementary method to mass spectrometers. The five samples used in this study contained nominal compositions of 0, 0.004, 0.04, 0.4 and 4 wt.% Ni. They were analyzed by both an Agilent 7900 Q-ICP-MS at SRNL and the JEOL JXA8530F Plus EPMA at the University of Minnesota. In addition to EPMA results (calibrated with high-precision Q-ICP-MS analyses) suggesting that the EPMA could address outstanding nuclear material characterization needs, these samples 1) showcase the ability of the EPMA to quantify not just trace concentrations, but trace concentrations in microparticles (1 m diameter, Fig. 1), and 2) offer a unique opportunity to evaluate the methodology for assessing the minimum detection limits of EPMA analyses.

McSwiggen, Peter [JEOL USA, 11 Dearborn Road, Peab↗
DOE OSTI2020

Airborne Release Fractions from Surrogate Nuclear Waste Fires Containing Lanthanide Nitrates and Depleted Uranium Nitrate in 30% Tributyl Phosphate in Kerosene

Airborne contaminants from fires containing nuclear waste represent significant health hazards and shape the design and operation of nuclear facilities. Much of the data used to formulate DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities,” from the U.S. Department of Energy, were taken over 40 years ago. The objectives of this study were to reproduce experiments from Pacific Northwest Laboratories conducted in June 1973 employing current aerosol measurement methods and instrumentation, develop an enhanced understanding of particulate formation and transport from fires containing nuclear waste, and provide modeling and experimental capabilities for updating current standards and practices in nuclear facilities. A special chamber was designed to conduct small fires containing 25 mL of flammable waste containing lutetium nitrate, ytterbium nitrate, or depleted uranium nitrate. Carbon soot aerosols showed aggregates of primary particles ranging from 20 to 60 nm in diameter. In scanning electron microscopy, ~200-nm spheroidal particles were also observed dispersed among the fractal aggregates. The 200-nm spherical particles were composed of metal phosphates. Airborne release fractions (ARFs) were characterized by leaching filter deposits and quantifying metal concentrations with mass spectrometry. The average mass-based ARF for 238 U experiments was 1.0 × 10 -3 with a standard deviation of 7.5 × 10 -4 . For the original experiments, DOE-HDBK-3010-94 states, “Uranium ARFs range from 2 × 10 -4 to 3 × 10 -3 , an uncertainty of approximately an order of magnitude.” Therefore, current measurements were consistent with DOE-HDBK-3010-94 values. ARF values for lutetium and ytterbium were approximately one to two orders of magnitude lower than 238 U. Furthermore, metal nitrate solubility may have varied with elemental composition and temperature, thereby affecting ARF values for uranium surrogates (Yb and Lu). In addition to ARF data, solution boiling temperatures and evaporation rates can also be deduced from experimental data.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Hot Hydrogen Exposure of U x Zr 1-x C y Nuclear Fuel: The Influence of Composition and Density

Refractory carbide nuclear fuel has been one of the most promising fuel candidates for space nuclear propulsion due to its high melting point, temperature stability, and compatibility in a hot hydrogen environment. In this study, U x Zr 1-x C y fuel was produced by means of a carbothermic reduction process in different UC compositions including 5,10, 20, and 30 at.% UC in the fuel compound. The powder feedstock was consolidated via direct current sintering with densities up to 97% of the theoretical density. The samples with different U x Zr 1-x C y compositions were exposed to hot hydrogen at 2600 K for a cumulative time of 300 min. The samples were characterized by SEM, XRD, density, and measured for mass losses. The high-density samples displayed improved performance in hot hydrogen by minimizing porous sites and reducing areas of direct contact with hydrogen gas, leading to reduced mass losses. Variations in sample density proved to induce large changes in mass loss rates, increasing them up to 90%. The compositions with higher UC content reported the largest mass losses in the study. The loss of uranium occurred primarily at the surfaces exposed to the hot hydrogen where changes in the lattice constant confirmed losses exceeding 50% of the initial UC content in higher compositions, specifically to U 0.3 Zr 0.7 C y . XRD analyses revealed the presence of UH 3 in U 0.3 Zr 0.7 C y suggesting that metallic uranium formed inside the sample as a product of carbon losses. High-density U x Zr x-1 C y fuel with UC concentrations at or below 20 at.% UC exhibited stability and negligible density changes in a high temperature hydrogen environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗