Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “URANIUM DIOXIDE”

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.

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

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

In-pile thermal conductivity of uranium dioxide at low burnup

In-pile thermal conductivity of uranium dioxide (UO 2 ) was investigated at low burnup levels (<0.2 kWd/kg-UO 2 ) to elucidate the concurrent effects of the fission damage and the thermal recovery. In-pile experiments were performed in the Halden reactor. The fuel centerline and the cladding temperatures were measured during the experiments, and the power level of specially designed test rodlets were also monitored. The uncertainty of the pellet-cladding gap's thermal resistivity was minimized with a liquid-metal bond, while comparable measurements were made in reference test rodlets without liquid metal bonding. The experimental data were analyzed using the inverse heat transfer approach. The heat conduction equation was solved by applying the measured temperature of the cladding as a boundary condition, and the calculated fuel centerline temperature was compared with the measured fuel centerline temperature to determine the relative matrix resistivity. Results showed that the in-pile matrix resistivity, the phonon-lattice interaction terms of the thermal conductivity, was ~1.5–2.5 times higher than its unirradiated value due to fission-induced damage. Finally, the in-pile annealing caused a significant recovery of the matrix resistivity, and the amount of recovery increased from ~20 to ~33%, while the annealing temperature increased from 700 to 1,000 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2020

Generation of the Thermal Scattering Law of Uranium Dioxide with Ab Initio Lattice Dynamics to Capture Crystal Binding Effects on Neutron Interactions

Scattering of thermal neutrons and Doppler broadening of epithermal neutron resonances in uranium and its compounds may be sensitive to crystal binding. The thermal scattering law (TSL) for uranium dioxide, which captures crystal binding effects, has been reevaluated for ENDF/B-VIII.0. Phonon spectra were generated using ab initio lattice dynamics for the paramagnetic phase and validated against experiment. Improved agreement with the Debye-Waller coefficient as a function of temperature is found relative to the spectrum used for the ENDF/B-VII.1 evaluation. The TSL was generated using the phonon expansion method within the NJOY nuclear data processing package and was found to be in reasonable agreement with inelastic neutron scattering measurements. The present evaluation predicts a reduction in the inelastic scattering cross section relative to ENDF/B-VII.1 and a total scattering cross section consistent with neutron transmission experiments

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Charge influence on point defect interactions with xenon bubbles in uranium dioxide

The interaction of xenon (Xe) bubbles with small self-defects in uranium dioxide (UO2) has been studied using molecular statics simulations. The results show that the pressure and size of bubbles have a minimal impact on the heterogeneity of the interaction between a charge neutral Xe bubble and charged point defects, while the local charge distribution around the bubble overwhelmingly determines the interaction strength and critical interaction distance. The charge effect on the interactions between Xe bubbles and defects was further confirmed by assessing the point defect interaction energies with negatively or positively charged Xe bubbles. The Xe density, or pressure, has a much smaller effect on the interaction energy between charged bubbles and small defects at least for low and medium pressure bubbles. The interactions between charged Xe bubbles and point defects basically follow the Coulomb electrostatic interaction law, which is independent of the empirical potentials employed for the Xe–UO2 system in this work.

Yang, L. (ORCID:0000000322166071)↗
NASA NTRS1973

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Mcdonald, G. E.↗
DOE OSTI2022

Induced Ferromagnetism in Epitaxial Uranium Dioxide Thin Films

Abstract Actinide materials have various applications that range from nuclear energy to quantum computing. Most current efforts have focused on bulk actinide materials. Tuning functional properties by using strain engineering in epitaxial thin films is largely lacking. Using uranium dioxide (UO 2 ) as a model system, in this work, the authors explore strain engineering in actinide epitaxial thin films and investigate the origin of induced ferromagnetism in an antiferromagnet UO 2 . It is found that UO 2+ x thin films are hypostoichiometric ( x <0) with in‐plane tensile strain, while they are hyperstoichiometric ( x >0) with in‐plane compressive strain. Different from strain engineering in non‐actinide oxide thin films, the epitaxial strain in UO 2 is accommodated by point defects such as vacancies and interstitials due to the low formation energy. Both epitaxial strain and strain relaxation induced point defects such as oxygen/uranium vacancies and oxygen/uranium interstitials can distort magnetic structure and result in magnetic moments. This work reveals the correlation among strain, point defects and ferromagnetism in strain engineered UO 2+ x thin films and the results offer new opportunities to understand the influence of coupled order parameters on the emergent properties of many other actinide thin films.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Suppressed thermal conductivity in hyperstoichiometric uranium dioxide controlled by phonon lifetimes

Thermal transport in nuclear fuels used for nuclear energy applications is directly tied to performance and reliability. Uranium dioxide (UO 2 ), one of the most important nuclear fuels, can accumulate excess oxygen atoms as interstitial defects, which significantly impacts thermal transport properties. In this study, thermal conductivities and inelastic neutron scattering (INS) measurements on UO 2+ x were performed at low temperatures (2–300K). The thermal conductivity of UO 2+ x (x = 0, 0.03, 0.04, and 0.11) is significantly suppressed compared to UO 2 except near the Néel temperature T N = 30.8 K, where it is independent of x. INS measurements demonstrate that the heat capacities and phonon group velocities of UO 2 and UO 2.08 are similar, and the suppressed thermal conductivity results from smaller phonon lifetimes. These perceptions advance our understanding of thermal transport properties in advanced nuclear fuels and guide safe and economic utilization of nuclear energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Multielectronic and multiatomic effects in the U O 4,5 x-ray absorption spectroscopy of uranium dioxide

In order to explain all of the spectral features observed in the U O 4,5 x-ray absorption spectrum of uranium dioxide (UO 2 ), it is necessary to include both multielectron effects and multiatomic effects. The 5d (core hole)-5f (electron) angular momentum coupling that gives rise to the giant resonance has been treated within ligand field density functional theory, and the electron scattering that generates the extended x-ray absorption fine structure has been included via the spectral simulation program FEFF: both within a UO 8 fluorite cluster picture. An atomic model is insufficient to explain all of the observed spectral features.

36 MATERIALS SCIENCE↗
DOE OSTI2021

U M subshell X-ray emission spectroscopy of uranium dioxide: the effect of excitation energy

There are very great differences between the non-resonant U M subshell X-ray Emission Spectra of Uranium Dioxide generated with low and high-energy excitation. Additionally, these different spectra are each supported by theoretical constructs. It will be shown that the differences between these experimental spectra can be reconciled and the utility of both theoretical approaches confirmed, including the use of FEFF in the low energy and high-energy regimes.

36 MATERIALS SCIENCE↗
NASA NTRS2014

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Hickman, Robert↗