SEARCH · Engineering Papers

Results for “uranium”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 973 records · Page 54

All sources NASA NTRS DOE OSTI

DOE OSTI2019

Homoleptic Aryl Complexes of Uranium (IV)

The synthesis and characterization of sterically unencumbered homoleptic organouranium aryl complexes containing U-C σ-bonds has been of interest to the chemical community for over 70 years. Reported herein are the first structurally characterized, sterically unencumbered homoleptic uranium (IV) aryl-ate species of the form [U(Ar) 6 ] 2- (Ar=Ph, p-tolyl, p-Cl-Ph). Magnetic circular dichroism (MCD) spectroscopy and computational studies provide insight into electronic structure and bonding interactions in the U-C σ-bond across this series of complexes. Overall, these studies solve a decades-long challenge in synthetic uranium chemistry, enabling new insight into electronic structure and bonding in organouranium complexes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2015

Evaluating ligands for use in polymer ligand film (PLF) for plutonium and uranium extraction

We describe a new analyte extraction technique using Polymer Ligand Film (PLF). PLFs were synthesized to perform direct sorption of analytes onto its surface for direct counting using alpha spectroscopy. The main focus of the new technique is to shorten and simplify the procedure for chemically isolating radionuclides for determination through a radiometric technique. 4'(5')-di-t-butylcyclohexano 18-crown-6 (DtBuCH 18 C 6 ) and 2-ethylhexylphosphonic acid (HEH[EHP]) were examined for plutonium extraction. Di(2-ethyl hexyl) phosphoric acid (HDEHP) were examined for plutonium and uranium extraction. DtBuCH 18 C 6 and HEH[EHP] were not effective in plutonium extraction. HDEHP PLFs were effective for plutonium but not for uranium.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2025

Single particle trace element analysis of uranium oxide standards with laser-induced breakdown spectroscopy

Femtosecond Laser-Induced Breakdown Spectroscopy (LIBS) was used to analyze over 1000 individual particles of uranium oxide trace element standard CRM124. Intra-granular analysis showed elemental concentrations were consistent across different surface locations on each uranium grain. However, inter-granular comparison revealed large heterogeneity between different grains of the same sample. By comparing the distribution of elemental concentrations across the particle populations, we were able to correctly identify a manufacturing process involving wet chemistry, rather than a dry mixing of two powders to give an overall chemical average. In conclusion, this demonstrates the utility of LIBS as a large-scale single particle analysis tool for nuclear forensics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

RIMS analysis of isotopically tagged uranium particles with application to Intentional Forensics

Traditional nuclear forensics approaches leverage isotopic measurements for characterizing nuclear forensics signatures in fuel cycle materials. To aid in provenance assessment, isotopically perturbed transition metal taggants can be added to fuel cycle materials. We present a case study for examining natural uranium oxide powder that was tagged with isotopically perturbed Mo before and after irradiation using resonance ionization mass spectrometry. In conclusion, this method accurately and precisely measures taggant and uranium compositions, rapidly enabling the ability to clearly discriminate between tagged particles and other material with application for robust assessment of material provenance.

Forensic Anthropology↗

DOE OSTI2021

Phase Relationships in the Carbon–Titanium–Uranium System for Ultra-High Temperature Nuclear Fuels

Carbides of uranium have attracted interest as fuels for nuclear thermal propulsion (NTP) to drive deep-space exploration owing to their attractive thermal and neutronic properties. Optimization of NTP technology, however, requires ultra-high temperature reactor environments to maximize the ratio of thrust to propellant to achieve peak rocket engine efficiency. Incorporation of transition metals into uranium carbides offers a pathway to increase the melting point of carbide fuels to address the operational challenges posed by NTP. Here, a thermodynamic model has been developed to examine the phase relationships in the C–Ti–U system at NTP conditions. Calculated phase relationships at ultra-high temperatures predict a stable (U, Ti)C solid solution. Additionally, experimental work on C–Ti–U synthesis via arc melting has been carried out, providing data on phase constitution and compositions that can be used to further refine the thermodynamic model that has been developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies is a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium disposed of, land use per energy generated, and carbon emissions per energy generated. Finally, it is noted that the features of some SMR designs impact three main aspects of fuel cycle performance: (1) small cores which means high leakage (there is a radial and axial component), (2) no boron which means heterogeneous core and extensive use of control rods and BPs, and (3) single batch cores. But not all of the SMR designs have all of these traits. As a result, the approach used in this study is therefore a bounding case and not all SMRs may be affected to the same extent.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Determining the minimum required uranium carbide content for HTGR UCO fuel kernels

There are three important failure mechanisms that must be controlled in high-temperature gas-cooled reactor (HTGR) fuel for certain higher burnup applications are SiC layer rupture, SiC corrosion by CO, and coating compromise from kernel migration. All are related to high CO pressures stemming from free O generated when uranium present as UO 2 fissions and the O is not subsequently bound by other elements. Furthermore, in the HTGR UCO kernel design, CO buildup from excess O is controlled by the inclusion of additional uranium in the form of a carbide, UC x . An approach for determining the minimum UC x content to ensure negligible CO formation was developed and demonstrated using CALPHAD models and the Serpent 2 reactor physics and depletion analysis tool. Our results are intended to be more accurate than previous estimates by including more nuclear and chemical factors, in particular the effect of transmutation products on the oxygen distribution as the fuel kernel composition evolves with burnup.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Neutronic and thermal-hydraulic feasibility studies for High Flux Isotope Reactor conversion to low-enriched uranium silicide dispersion fuel

An iterative design process involving neutronic and thermal-hydraulic modeling and simulation has been employed to assess the feasibility of converting the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) silicide dispersion fuel. ORNL is funded by the National Nuclear Security Administration to evaluate HFIR conversion. Previous HFIR conversion studies focused on U-10Mo monolithic fuel; however, due to potential fabrication issues with the complex HFIR U-10Mo fuel design, ORNL is evaluating U 3 Si 2 -Al dispersion fuel as an alternative LEU fuel system.Fueled by 10.1 kg of HEU and operated at 85 MW, HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world. Retrofitting a compact, high-power density, HEU-based core with LEU is a challenging problem to solve, especially when considering the conversion requirements. Neutronic and thermal-hydraulic analyses were performed with Shift and HSSHTC, respectively, to predict reactor performance and thermal safety margins. A number of designs were proposed and evaluated using an iterative approach in an effort to show that reactor performance could match that obtained using HEU fuel and that thermal safety margins were adequate. Furthermore this study concludes that conversion of HFIR with U 3 Si 2 -Al LEU fuel is feasible if, among other requirements, the fuel meat region is centered and symmetric about the fuel plate thickness centerline, the active fuel zone length is increased from 50.80 cm to 55.88 cm, the proposed fabrication tolerances can be met, and the fuel can be qualified for HFIR conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A Low-Enriched Uranium (LEU) option for the conversion of FRM II

The Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) is Germany’s most powerful research reactor and uses Highly Enriched Uranium (HEU) fuel enriched at 93 %. The Technical University of Munich (TUM) operates the reactor and has been mandated to convert the FRM II to a lower-enrichment fuel. In preparation for the lower-enrichment fuel downselection, planned for 2023, TUM is evaluating multiple conversion scenarios. In this paper, it is demonstrated that a conversion of FRM II to Low-Enriched Uranium (LEU) is scientifically possible when using the novel monolithic U-10Mo fuel system, which is currently in the qualification process. As a research reactor, safety criteria, neutron flux distribution and cycle length are key metrics. To begin, neutronic and thermal-hydraulic models were created based on current technical drawings and relevant material properties. Specifically tailored computational methods and coupling schemes have also been developed to properly evaluate all relevant characteristics for the conversion of the FRM II reactor. Using these models and methods, a systematic parameter study was performed to explore the defined design space. The results of that study indicate that many LEU designs using monolithic U-10Mo appear to be viable to convert FRM II to LEU. One example of an LEU solution for the FRM II reactor that fulfills the stated safety, compatibility and scientific performance requirements is discussed in more detail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Uranium integral fission product yields for a spectrally-shaped 14.1 MeV neutron source at the National Ignition Facility

This paper describes the experimental results for an energy tuning assembly created to modify the National Ignition Facility deuterium–tritium fusion neutron source into a notional thermonuclear and prompt fission neutron spectrum, which has applications in integral measurements, nuclear data benchmarks, and radiation effects on microelectronics. The Monte Carlo neutron transport utilized MCNP5 to estimate the ETA-modified fluence using the ENDF-B/VIII.0 and IRDFF-II continuous energy nuclear data libraries, and SCALE Sampler was used to estimate the systematic nuclear data covariance using ENDF-B/VII.1 and IRDFF-II in a 252-group structure. The experiment fielded eight activation foils and a highly enriched uranium sample. This provided fifteen reaction channels that were used in a forward-fit comparison to the modeled results and to unfold the neutron spectrum using STAYSL. Gamma-ray spectrometry was performed on the activation and highly enriched uranium foils, and the reduced Χ 2 between the modeled and experimental values was 1.21. The results from the STAYSL unfold, reduced Χ 2 = 1.62, indicated that the modeled neutron spectrum was achieved and the systematic nuclear data uncertainty associated with the neutron transport and activation product cross sections was representative of the experiment. Integral cumulative fission product yield data were collected for 37 mass chains with a combination of gamma-ray spectrometry and radiochemical analysis. Fission product analysis was generally in agreement with two models using a semi-empirical fit and the General Observables of Fission code, with the exception of mass chains 88, 109, 111, 112, 113, 129, 139, 142, 144, 151, and 156.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Smoothed particle hydrodynamics modeling and analysis of oxide reduction process for uranium oxides

A common kinetic feature for oxide reduction chemical/electrochemical processes is oxygen transport via a porous metallic layer, which has been considered as a rate-determining step for reducing uranium oxides to metallic uranium. Accounting this kinetic behavior must involve the resolution of the moving reactive interface between shrinking oxide and expanding metal phases. This study presents a numerical model using smoothed particles hydrodynamics (SPH) to effectively deal with the evolution of the shrinking core reaction interface and oxygen transport via mass transfer of lithium oxide (Li 2 O) species in multiple mass transfer domains. We successfully validated the proposed model against a theoretical derivation for the oxide reduction process handling a shrinking oxide core with molten salt and metal ash medium on a simple planar geometry. Armed with successful validation results, the model examined a realistic reactant geometry to extend the arguments beyond the one-dimensional analyses, allowing the proposed model to apply to general application scenarios with multi-dimensional geometry. Here, this study demonstrated that the proposed model could simulate and evaluate an arbitrary reaction basket design without iterative experimental trials, which is prohibitive for a scaled high-temperature molten salt study in an inert environment. The construct potentially provides not only deep insights on multiphysics behaviors governing the process dynamics but also a robust framework for evaluating and screening candidate basket designs in the most cost-effective manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Uranium hydride corrosion. II Modeling spatially random growth using reaction and nucleation kinetics

A uranium hydride (UH 3 ) corrosion kinetics model has been developed, based on dimensionally restricted nucleation kinetics, which combines heretofore separate reaction and nucleation kinetics models into a single predictive construct. This theoretical framework accounts for the superficial and random pitting behaviour of UH 3 corrosion. The model can be used to generate the net corrosion surface area and volume fractions for a uranium specimen, generally accounting for its geometric form factor through the surface-to-volume ratio. Furthermore, the combined corrosion kinetics model enables the generation of quantities which may be directly compared to those obtained by non-destructive, non-contact evaluation methods.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Spatially-Resolved Analyses of Aerodynamic Fallout from a Uranium-Fueled Nuclear Test

The fiive silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x-ray spectroscopy. Several samples display compositional heterogeneity suggestive of incomplete mixing between major elements and natural U ( 238 U/ 235 U = 0.00725) and enriched U. Samples exhibit extreme spatial heterogeneity in U isotopic composition with 0.02 < 235 U/ 238 U < 11.84 among all five spherules and 0.02 < 235 U/ 238 U < 7.41 within a single spherule. Moreover, in two spherules, the 235 U/ 238 U ratio is correlated with changes in major element composition, suggesting the agglomeration of chemically and isotopically distinct molten precursors. Two samples are nearly homogenous with respect to major element and uranium isotopic composition, suggesting extensive mixing possibly due to experiencing higher temperatures or residing longer in the fireball. Linear correlations between 234 U/ 238 U, 235 U/ 238 U, and 236 U/ 238 U ratios are consistent with a two-component mixing model, which is used to illustrate the extent of mixing between natural and enriched U end members.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Effects of flow on uranium speciation in soils impacted by acidic waste fluids

Radioactive acidic liquid waste is a common byproduct of uranium (U) and plutonium (Pu) enrichment and recycling processes whose accidental and planned release has led to a significant input of U into soils and sediments across the world, including at the U.S. DOE's Hanford site (WA, USA). Because of the particularly hazardous nature of U, it is important to predict its speciation when introduced into soils and sediments by acidic waste fluids. Of fundamental importance are the coupled effects of acid-driven mineral transformation and reactive transport on U speciation. To evaluate the effect of waste-fluid residence time and co-associated dissolved phosphate concentrations on U speciation in impacted soils and sediments, uncontaminated surface materials (from the Hanford Site) were reacted with U-containing synthetic acidic waste fluids (pH 2) amended with dissolved phosphate concentrations in both batch (no flow) and flow-through column systems for 7–365 days. By comparing dissolved U behavior and solid phase speciation as a function of flow regimen, we found that the availability of proton-promoted dissolution products (such as Si) to sequester U into uranyl silicates was dependent on waste fluid-sediment contact time as uranyl silicates were not detected in short contact time flow-through systems but were detected in no-flow, long contact time, reactors. Moreover, the dominance of uranyl phosphate as neoprecipitate U scavenger (principally in the form of meta-ankoleite) in phosphate amended systems confirmed the importance of phosphate amendments for an efficient sequestration of U in the soils and sediments. Overall, our experiments suggest that the formation of uranyl silicates in soils impacted by acidic waste fluids is likely to be limited unless reaction products are allowed to accumulate in soil pores, highlighting the importance of investigating soil U speciation in flow-through, transport-driven systems as opposed to no-flow, batch systems. Finally, this study provides insights into uranium speciation and its potential changes under acidic conditions for better prediction of risks and subsequent development of efficient remediation strategies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Chemical durability and corrosion-induced microstructure evolution of compositionally complex titanate pyrochlore waste forms with uranium incorporation

Chemical durability of compositionally-complex rare-earth titanate pyrochlore solid solutions with or without uranium incorporation was investigated by semi-dynamic leaching testing. A Ti-enriched amorphous passivation film formed on the surface due to the surface alteration-reorganization mechanism with preferential release of the weakly oxygen-bonded rare earth elements. Elemental release was found to be strongly correlated with chemical disorder, with a negligible correlation with entropy. The release rate of the uranium gradually decreased with increasing numbers of components, implying a possibility of designing and optimizing waste form performance with enhanced chemical durability by controlling composition complexity and chemical disorder for effective nuclear waste management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Uranium nitride tristructural-isotropic fuel particle

A summary of activities around concept development, design, manufacturing, and early irradiation testing of a novel coated fuel particle, uranium nitride tristructural-isotropic fuel is provided. This fuel particle offers significantly higher uranium density over historic manifestations of coated fuel particles and may be more optimal for a range of advanced reactor applications. After reviewing the design process, steps involving production of this fuel form are discussed. Low burnup irradiation data on this fuel particle are now available, indicating that performance metrics have been met to date based on the original design expectations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Experimental assessment of antimony (Sb) in pure uranium for immobilizing fission product lanthanides

The use of fuel additives is one of the concepts to mitigate fuel cladding chemical interaction (FCCI) for metallic fuel because fission product lanthanides are expected to be immobilized by the additive. Antimony (Sb) has been discovered to be a good candidate in UZr fuel. The present study focuses on its mechanism for immobilization in pure uranium, U–4Sb alloy was fabricated to understand the Sb behavior, while U–4Sb–4Ce was fabricated to simulate the case when lanthanides are generated. Both of the as-cast and annealed samples were characterized by scanning electron microscope (SEM) and energy dispersive spectrometer (EDS). U–Sb precipitates are formed in U–4Sb alloy, while U–Sb and Ce–Sb were found in U–4Sb–4Ce alloy, thermal exposure does not change the Sb-precipitation morphologies or chemical composition. Furthermore, diffusion couple tests between those alloys and cladding materials (Fe or HT9) under 650 °C for 500 h were performed and analyzed using SEM/EDS. Diffusion couple tests demonstrate the reaction between cladding and uranium, while no reaction between Sb-precipitations and cladding materials was found.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗