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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.

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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Structure characterization and steam oxidation performance of U 3 Si 2 with Zr alloying additions

Research and development of high uranium density fuels supports the on-going effort to enhance reactor technology, safety, performance, and economics. Among high uranium density fuels, U 3 Si 2 is a favored fuel candidate due to its higher uranium density and improved thermal conductivity when compared to traditional UO 2 . While U 3 Si 2 demonstrates an improved fuel economy, it suffers from rapid degradation when exposed to water at T > 300°C. In order for U 3 Si 2 to be implemented in water cooled reactors, the water reaction must be understood and mitigated. Reported here is the performance of Zr alloying additions to U 3 Si 2 exposed to high temperature steam oxidation environments. Here, the investigation shows that homogenization heat treatment contributes to a delayed onset of oxidation and an improved homogeneity throughout the arc-melt fabricated alloyed compositions. Of the pre- and post-annealed samples, the post-annealed 7 vol% Zr addition exhibited the highest onset temperature of oxidation. A preliminary computational investigation to complement the experimental analysis assesses the solubility of Zr in U 3 Si 2 and its incorporation in the crystal lattice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Influence of point defect accumulation on in-pile thermal conductivity degradation: Fuel rod defect distribution and deviation between in-pile and post irradiation thermal conductivity

As nuclear fuel burn-up increases, its thermal conductivity degrades due to the accumulation of defects that lead to increased phonon scattering rates. This results in a rise in the centerline temperature of the fuel rod, whereby heat generation must be decreased to avoid undesired behavior such as fuel melting and extensive fission gas release. Fuel performance codes are utilized to optimize the fuel's operational conditions; while they are based on established physical principles, their empirical nature limits their predictive capabilities. Recently, an effort has been made to develop predictive fuel performance codes for commonly used nuclear fuels, as well as accelerated qualification of advanced nuclear fuels. In this report, we elaborate on the importance of careful analysis of point defects’ impact on thermal conductivity in fuel performance analysis. Further, a model is presented where point defect concentration is estimated based on Rate Theory modeling and used as input to the Klemens-Callaway model to calculate their contribution to the degradation of thermal conductivity in UO 2 under prototypical irradiation conditions. This analysis suggests that point defect concentration is significant at the rim of the fuel pellet and neglecting this leads to underestimation of the centerline temperature, which may have consequences on fission gas behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Finite temperature properties of uranium mononitride

Uranium mononitride (UN) is a promising nuclear fuel that combines the advantageous properties of readily used UO 2 and uranium alloys, such as high melting temperature and high uranium density, and thermal conductivity, respectively. A better understanding of UN behavior at operating temperatures can be obtained from finite temperature data, such as elastic properties. To get this information, ab initio molecular dynamics (AIMD) simulations were performed at five different temperatures using constant volume (NVT) and constant pressure (NPT) ensembles. Initially, the performance of PBE functional in reproducing experimental crystallographic properties and magnetic ordering is assessed. The finite temperature phonon dispersions are calculated using NVT simulation results, which show a softening of the phonon modes with increasing temperature. The NPT results are used to obtain the thermal expansion of UN and finite temperature electronic properties. The calculated thermal expansion is compared with our measurements using neutron diffraction. Additionally, the temperature dependent elastic properties of UN are evaluated using the strain-stress method in AIMD simulations, indicating that UN becomes softer and more compressible with increasing temperature. Also, the calculated Young’s modulus slope is in very good agreement with the experiment. The finite temperature heat capacity and electronic thermal conductivity are calculated from AIMD simulations, which are in better agreement with the experiment than the heat capacity and thermal conductivity calculated using the structures relaxed at 0K. Finally, the thermal diffusivity from AIMD has opposite temperature dependence compared to experimental results, which we argued comes from the underestimated electronic thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Evolution of dislocation loops and voids in post-irradiation annealed ThO 2 : A combined in-situ TEM and cluster dynamics investigation

The effect of isochronal annealing on the evolution of dislocation loop and void population in proton irradiated ThO 2 has been investigated. Post-irradiation annealing in other actinide oxides like UO 2 shows significant loop coarsening. ThO 2 samples were irradiated with 2 MeV protons up to a dose of 0.1 dpa at 600°C. Post-irradiation isochronal annealing was performed at 600, 800, 1000 and 1100°C for 1 h at each temperature using in-situ TEM. Only faulted 1/3<111> type dislocation loops were observed, and their sizes and distribution were characterized. Further, the population of self-interstitial atom (SIA) dislocation loops did not show any significant growth and coarsening. Additionally, nanometric voids were observed at annealing temperatures of 1000 and 1100°C. Using cluster dynamics (CD), we have studied the nucleation and growth of point defects and defect clusters, i.e., SIA prismatic dislocation loops and nanometric and sub-nanometric voids in proton irradiated ThO 2 . The CD model was further utilized to predict the growth and coarsening of loops and voids during isochronal annealing at the experimental and higher temperatures. The model did not predict significant SIA loop growth which closely corresponds to the TEM observations. CD predicted SIA loop coarsening is insignificant even at high annealing temperature of 1500°C because the model only considers the growth of defect clusters by absorption of like point defects, i.e., SIA loops absorb interstitials and voids absorb vacancies, and cannot account for their migration and coalescence due to elastic interaction. The CD model also predicts the evolution of nanometric voids having mean size within the error bounds of TEM observations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Optical vibrational spectroscopic signatures related to U 3 O 8 production processes

Uranium ore concentrates are materials found early within the nuclear fuel cycle and contain high concentrations of uranium in an easily transported form, making the concentrates a likely target for illegal diversion. These concentrates are typically converted to U 3 O 8 for further processing and, therefore, may lose specific physicochemical characteristics in determining the materials’ source and processing history. In this work, we explore the Raman spectra of eight oxide samples produced from various uranium ore concentrates and processing pathways to examine the presence of spectroscopic signatures relating to each sample's process history. Samples produced from amine extraction and dialkylphosphoric acid extraction processes show unique characteristics due to high concentrations of α-UO 3 , whereas samples calcinated from metallic diuranates do not form pure α-U 3 O 8 because of metallic ion inclusions. Pure α-U 3 O 8 oxide samples are obtained through calcination of ammonium diuranate, ammonium uranyl carbonate, and metastudtite intermediates. The Raman spectra of these oxide samples show close agreement with pristine α-U 3 O 8 spectra. However, deviations from the pristine spectra are observed in the 300–460 cm -1 spectral range. In conclusion, these deviations are unique identifying signatures that were likely created by lasting effects from the process history. Spectral center of mass calculations indicate grouping of samples based on processing history.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chemical interaction and compatibility of uranium mononitride and alumina forming austenitic stainless steel

Uranium mononitride (UN) and alumina forming austenitic (AFA) stainless steel are a potential fuel-cladding combination for the lead-cooled fast reactor (LFR). Chemical compatibility between UN and AFA steel needs to be verified before implementation in a nuclear reactor. Diffusion couple experiments at 823 K and 1023 K were conducted for nonirradiated UN samples in contact with as-cast (no thermally grown Al 2 O 3 ) and preoxidized (with thermally grown Al 2 O 3 ) AFA for 500 and 1000 h in an inert environment. Preoxidized AFA exhibited little to no interaction with all UN samples tested at both 823 K and 1023 K, displaying the stability and capability of the Al 2 O 3 layer to prevent chemical interaction and inter-diffusion with UN. Chemical interaction occurs between UN and as-cast AFA. At 1023 K, an aluminum and nitrogen rich phase (likely AlN) formed along the interface of as-cast AFA and UN samples. At 823 K the AlN phase was not prominently observed due to the reduced diffusivity of aluminum through AFA. The aluminum and nitrogen-enriched phase was also observed in a high temperature pressure-assisted test sample of UN and as-cast AFA thermally treated at 1373 K. Finally, in UN samples doped with a low weight percent of UO 2 (< 3 wt%), AlN was not detected along the interface at either temperatures, and an Al 2 O 3 layer likely formed along the interface and prevented further chemical interaction between UN and as-cast AFA.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Compatibility of Zircaloy-4 in eutectic PbSn and PbBiSn liquid metals

Here, a novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.

Bismuth↗
DOE OSTI2023

In-field alpha spectrometry system for uranium enrichment verification of uranium hexafluoride

A portable In-Field Alpha Spectrometry (IFAS) system was developed to support on-site uranium enrichment determination for samples taken at facilities working with gaseous UF 6 . It consists of a modified commercial alpha spectrometer, a vacuum pump, a laptop computer, and associated connectors and cables. The IFAS uses specially designed Single-Use Destructive Assay (SUDA) samplers to capture gaseous UF 6 and convert and store it as UO 2 F 2 in a thin-film zeolite. To examine the quantitative performance of the IFAS system a large number of replicate SUDA samples was acquired for natural uranium (NU, 0.72 wt%, 23 samples), low enriched uranium 1 (LEU1, 2.33 wt% 235 U, 10 samples), and low enriched uranium 2 (LEU2, 4.62 wt% 235 U, 10 samples). The standard deviation (SD) values for uranium enrichment determination with the IFAS for NU, LEU1, and LEU2, in 20-hour assays, were 0.037 wt% 235 U, 0.063 wt% 235 U, and 0.072 wt% 235 U, respectively. In addition, a separate analysis suggests that comparable SD values can be achieved in less time, approximately 8-10 hours. For context, these SD values are well below the International Atomic Energy Agency’s International Target Values for the determination of uranium enrichment in non-destructive assay measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Study of cluster ions produced from ToF-SIMS analysis of a U-6%Nb target

Cluster ions have been previously observed during time-of-flight secondary ion mass spectrometry (ToF-SIMS) analysis of metals and metal oxides. Furthermore, we have used ToF-SIMS to investigate cluster ions formed from the hydrocarbon-containing overlayer, the mixed U and Nb surface oxide, and underlying metal of a U-6 %Nb (U6Nb) target. In the overlayer, we observe U x O y + oxides and U-species likely containing hydrocarbons. In the surface oxide, we observe UO 2 +, U 2 O 2 +, U 3 O 5 +, and U 4 O 6 + as the most intense ions for each family of oxide ions containing x U atoms. Nb oxides for NbO 1-2 - were only observed in negative polarity. In contrast to the oxide, analysis of the underlying U6Nb alloy resulted in repeating units of U n +, U n (Nb)+, and U n (Nb 2 ) + ions for n = 3–11 as the highest intensity ions for each family of ions containing n U atoms. Nb n + clusters were not observed. U n +, U n (Nb)+, and U n (Nb 2 )+ clusters containing C- and O-species were observed and were likely produced from soluble C or O species or precipitates known to be present in U6Nb.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Material Interactions in Severe Accidents – Benchmarking the MELCOR V2.2 Eutectics Model for a BWR-3 MARK-I Station Blackout: Part I – Single Case Analysis

Here in this analysis, the two material interaction models available in the MELCOR code are benchmarked for a severe accident at a BWR under representative Fukushima Daiichi boundary conditions. This part of the benchmark investigates the impact of each material interaction model on accident progression through a detailed single case analysis. It is found that the eutectics model simulation exhibits more rapid accident progression for the duration of the accident. The slower accident progression exhibited by the interactive materials model simulation, however, allows for a greater degree of core material oxidation and hydrogen generation to occur, as well as elevated core temperatures during the ex-vessel accident phase. The eutectics model simulation exhibits more significant degradation of core components during the late in-vessel accident phase – more debris forms and relocates to the lower plenum before lower head failure. The larger debris bed observed in the eutectics model simulation also reaches higher temperatures, presenting a more significant thermal challenge to the lower head until its failure. At the end of the simulated accident scenario, however, core damage is comparable between both simulations due to significant core degradation that occurs during the ex-vessel phase in the interactive materials model simulation. A key difference between the two models’ performance is the maximum temperatures that can be reached in the core and therefore the maximum ΔT between any two components. When implementing the interactive materials model, users have the option to modify the liquefaction temperature of the ZrO 2 -interactive and UO 2 -interactive materials as a way to mimic early fuel rod failure due to material interactions. Through modification of the liquefaction of high melting point materials with significant mass, users may inadvertently limit maximum core temperatures for fuel, cladding, and debris components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Spark plasma sintering of fuel meats for U 3 O 8 based dispersion fuels

Research and test reactors often use dispersion-type fuel due to its increased thermal conductivity and burn-up capabilities compared to conventional fuel. Al-U 3 O 8 (aluminumtriuranium octaoxide) dispersion fuels have several advantages over their competitors, such as higher service temperature and better stability of oxygen stoichiometry. However, the two-step fabrication of dispersion fuel causes undesirable porosity in cold-pressed fuel meats that is preserved in co-extruded fuel plates. To combat this, spark plasma sintering (SPS) was used for the fabrication of Al-15, 20, and 30 vol% U 3 O 8 and 8 and 12 vol% Mo-U 3 O 8 fuel meats for the Al-U 3 O 8 time. The in situ SPS data was used to construct and validate Master Sintering Curves (MSCs) with accuracies in Al fuels at 0.02 g/cm 3 , and Mo fuels at 0.07 and 0.17 g/cm 3 . The as-sintered fuel meats were characterised using x-ray diffraction (XRD) and scanning electron microscopy (SEM) to understand chemical and physical changes following the SPS process. The pellets exhibited very high relative densities, the U 3 O 8 was observed to undergo reduction to UO 2 .

Aluminium↗
DOE OSTI2024

Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility

Uranium oxycarbide (UCO)-bearing tri-structural isotropic (TRISO) particle fuels are expected to be used in numerous U.S. commercial reactor applications within the next decade. Here, in this work, we reviewed historical particle fuel transient experiments to identify gaps in TRISO fuel performance transient testing. A BISON–Griffin modeling framework was then developed to conduct preliminary TRISO transient analyses and begin to address these gaps. The framework was demonstrated using limiting-case transient conditions from a prototypic high-temperature gas-cooled reactor (HTGR). It was then applied to develop a matrix of experiments that could be performed in the Transient Reactor Test Facility (TREAT) to (1) evaluate UCO-fueled particle performance at moderate and high heat rates, (2) assess whether historical testing involving UO 2 -fueled particles is applicable to modern UCO-fueled particles, (3) deconvolute the impacts of temperature and heat rate on particle transient response, and (4) collect the data needed for fuel performance model validation and/or further development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment

Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO 2 fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Oxidation of Dimethylsulfoxide in the Presence of Uranyl Salts

Here, we report here the oxidation of dimethyl sulfoxide in the presence of uranyl salts at temperatures typically considered ‘safe’ for its handling. Uranyl salts, even in the absence of light, appear to accelerate the well-known oxidative decomposition of DMSO in air. Sulfate, a product of the decomposition, was captured in the crystalline state as the uranyl sulfate phase, UO 2 (DMSO) 2 (SO 4 ).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Time-resolved formation of uranium and silicon oxides subsequent to the laser ablation of U 3 Si 2

The early-time kinetic behavior of species in a plume produced from laser ablation of U 3 Si 2 in an environment containing 2% O 2 is characterized using time-resolved absorption spectroscopy. The UO band around 593.55 nm and the SiO band around 230 nm is observed as well as various atomic and ionic uranium transitions. Temperatures and concentrations of these species are tracked and reported as well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Optical spectroscopy and modeling of uranium gas-phase oxidation: Progress and perspectives

We report that studies related to U gas-phase oxidation through plasma- and thermo-chemistry are important for many fields, including environmental monitoring, forensic analysis, debris analysis in a weapon detonation event, and nucleation physics. Recently, significant efforts have been made to understand the chemical pathways involved in the progression from U atoms to diatoms (UO) and polyatomic molecules (UxOy), employing optical spectroscopy tools and computational modeling. In many studies, laser ablation of U or a U-containing flow reactor are used as a highly resource-efficient, repeatable, tunable, and lab-scale testbed for studying gas-phase oxidation in U plasmas. The spectroscopic analysis of high-temperature gas-phase oxidation of U is challenging due to the congested U spectra, resolution limitations of instrumentation, and the numerous chemical reaction pathways possible. This article focuses on the current understanding and challenges related to studying U plasma chemistry, specifically U gas-phase oxidation and molecular formation, via optical spectroscopy of plasmas and associated computational and spectral modeling. The physical and chemical processes involved in the evolution from U atoms to U oxide molecules to nanoparticles and agglomerates (i.e., debris) are discussed in the context of optical spectroscopic studies. The article concludes by highlighting opportunities for future research efforts based on existing knowledge published in the literature.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Uranium corrosion characterization by handheld laser-induced breakdown spectroscopy

In this study, a commercial-off-the-shelf handheld laser-induced breakdown spectroscopy system was modified, calibrated and tested for discrimination between uranium oxide and uranium hydride when assessing uranium corrosion products. Operating parameters, system modifications, and onboard data analysis were developed to accomplish the analytical task. This work represents the first laser-induced breakdown spectroscopy analysis of bulk uranium hydride, a pyrophoric material, within an inert glovebox. Finally, the work shown here illustrates a rapid identification technique for uranium corrosion types from pure UO 2 or U 3 O 8 to pure UH 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗