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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Additive Manufacturing for Powering the Blue Economy Applications: A Tidal Turbine Blade Case Study: Preprint

As the marine renewable energy industry continues to expand, innovation in the manufacturing space must grow accordingly to reduce costs and ensure the economic feasibility of new technologies. Additive manufacturing, more commonly known as 3D printing, provides an alternative for rapid prototyping of marine hydrokinetic technologies, particularly supporting Powering the Blue Economy initiatives of the U.S. Department of Energy Water Power Technologies Office. This study explores the application of additive manufacturing in the development of marine hydrokinetic structures, focusing on material and printing method selection, design, and analysis of a 3D-printed spar for an axial-flow tidal turbine blade. Corrosion-resistant metals were deemed ideal due to the loads and harsh marine environment the blade would experience. Laser metal deposition methods were determined to be the most effective and scalable for the considered scale. The designed spar adapts its geometry to the blade - a feature uniquely suited to additive manufacturing - and is intended to serve as the blade's primary structural component. A finite element model was used to study stresses and deformations under loading conditions. The spar was manufactured using 316L stainless steel through direct energy deposition, and defects were assessed and recorded. Future efforts will include mechanical testing of the spar. This research establishes a benchmark process for using additive manufacturing in developing marine hydrokinetic structures, paving the way for future optimization and techno-economic analysis.

additive manufacturing↗
DOE OSTI2021

Effects of Electron Irradiation on Candidate Materials for Target Window in Accelerator-Driven Mo-99 Production

Room-temperature (RT) axial tensile tests were conducted in 2018 and 2021 with flat, sub-sized, dog-boned samples of non-irradiated and irradiated Inconel 718 and irradiated beryllium (Be). The sample gauge length was 7.62 mm. Tensile properties of interest are the engineering yield stress (YS), ultimate tensile stress (UTS), uniform elongation (UE) and total elongation (TE). The electron dose of the 2021 samples was ≈2 times the dose of samples tested in 2018. Relative dose rates of 0, 1 (2018), and 2 (2021) are assigned to these samples. The 2021 electron-irradiated Be samples were brittle with a low failure stress of 338±64 MPa based on two samples. The failures occurred in the gauge section. The 2018 results for electron-irradiated Be indicated a slight decrease in ductility and strength for the one electron-irradiated sample that failed in the gauge section.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A System-Level Model for Estimating Residual Strain and Life of Nuclear Reactor Coolant System Components Under Connected-System-Thermal–Mechanical Boundary Conditions

Background: Environmental-assisted fatigue (EAF) is a major issue for the long-term survival of nuclear power plant fleets in the U.S. and worldwide. Multi-material welded regions (e.g., nozzles) and other high-stress regions of reactor coolant system (RCS) components are prone to EAF-related damage. Objective: The discussed work describes a system-level finite element (FE) model of RCS components of a pressurized water reactor (PWR). This is with the goal of predicting the stress hotspots, strain residuals, strain amplitudes and the resulting fatigue lives. Methods: Herein, the FE model was developed considering system-level loading conditions (under connected system thermal-mechanical boundary conditions). Thermal-mechanical stress analysis was performed considering thermal stratification and a design-basis reactor loading cycle. Based on the FE model results, the strain residuals, strain amplitudes and resulting fatigue lives of RCS components were predicted. Results: The results show that some of the RCS components can have significantly different strain amplitudes, residual strain, and fatigue lives, despite having similar geometry and material. Higher residual strain can lead to accelerated cyclic hardening of material and the associated effect of EAF. The simulated component-level strain profile (under realistic multi-axial-multi-physics loading cycle) can guide the selection of appropriate test inputs for conducting laboratory-scale EAF tests, which is a focus of future works. Conclusions: Despite similar geometry and material the RCS component can have significantly different strain profiles and resulting fatigue lives. The discussed approach can help to identify and prioritize the RCS components for conducting expensive nondestructive evaluation (NDE) inspections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Progress Report on Computational Analyses of Water-Based NSTF

CFD analysis has been focused on important component-level phenomena using STARCCM+ to supplement the system analysis of integral system behavior. A notable area of interest was the cavity region. This area is of particular interest for CFD analysis due to the multi-dimensional flow and complex heat transfer (thermal radiation heat transfer and natural convection), which are not simulated directly by RELAP5. CFD simulations allow for the estimation of the boundary heat flux distribution along the riser tubes, which is needed in the RELAP5 simulations. The CFD results can also provide additional data to help establish what level of modeling detail is necessary in RELAP5. It was found that the flow profiles in the cavity region are simpler for the water-based concept than for the air-cooled concept. The local heat flux noticeably increases axially, and is higher in the fins than in the riser tubes. These results were utilized in RELAP5 simulations as boundary conditions, to provide better temperature predictions in the system level analyses. It was also determined that temperatures were higher in the fins than the riser tubes, but within design limits for thermal stresses. Higher temperature predictions were identified in the edge fins, in part due to additional thermal radiation from the side cavity walls.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Mechanical Testing of FeCrAl Tubing

The axial tensile properties of C26M tube were successful measured at Los Alamos National Laboratory. Test were conducted at a constant strain rate of 10 -3 s -1 and three different temperatures (22, 300, and 600°C). As typically expected, an increase in the testing temperature lead to a decrease in both the yield stress (YS), from 710 to 273 MPa, and ultimate tensile strength (UTS), from 730 to 290 MPa. The ultimate tensile strength was observed almost immediately post-yielding for all test conditions, and thus the uniform elongation was limited to 1.2-1.6%. The total elongation of the 600°C specimens was notably higher (~47%) than both the 22 and 300°C specimens, which exhibited a similar response (~10%).

36 MATERIALS SCIENCE↗
DOE OSTI2019

A Salt Repository Concept for CSNF in 21-PWR Size Canisters

The most straightforward concept for disposal of large, heavy packages containing commercial spent nuclear fuel (CSNF) in a repository in bedded salt, would be to emplace them directly on the floor in emplacement tunnels. In-tunnel axially aligned horizontal emplacement would minimize excavated volume and avoid drilling of large-diameter emplacement boreholes. A similar concept was proposed in Germany for direct disposal of POLLUX ® canisters. The repository would be constructed at a depth of 500 to 1,000 m for isolation from the surface, and for sufficient overburden stress to ensure creep reconsolidation of repository openings. It could entail modular panels of emplacement tunnels arranged on headings oriented in cardinal directions from a central core, to accommodate the estimated 140,000 MTU total U.S. CSNF inventory. To do so, the overall area of the repository layout would be approximately 20 km 2 . Many layouts are possible, but the approach should be modular, excavation should be deferred for as long as possible to avoid maintenance, and the emplacement areas should share support facilities and shafts. Vertical shafts would be used in accordance with mining practice in sedimentary basins. Large diameter shafts would be needed for ventilation exhaust and waste transport, with smaller shafts for waste salt removal, men & materials, and ventilation intake. The spacing between disposal tunnels as estimated from thermal modeling, seeks to limit the maximum average areal thermal load in the panels to 11 W/m 2 to control long-term heat buildup in the host rock. Peak salt temperature would occur within a few years and would be dominated by each waste package locally, simplifying thermal management. There would be some flexibility to decrease the package spacing or increase the emplacement thermal power limit Backfilling emplaced waste packages immediately with mine-run crushed salt would provide shielding and expedite reconsolidation. This arrangement would isolate adjacent waste packages from one another by the intervening backfill, especially after it reconsolidates and its properties approach those of intact salt. After the repository is fully loaded and the performance confirmation program is complete, activities to permanently close the repository would be initiated. During closure operations all openings in the host salt would be backfilled, then shafts would be sealed, and boreholes plugged. Plans for the Waste Isolation Pilot Plant (WIPP) show how sealing and plugging could be done. A monitoring program could continue for 50 years or longer after repository closure. With an emplacement thermal power limit of 10 kW per waste package, nearly all the CSNF that is projected to be produced by the current fleet of reactors in the U.S. could be emplaced over a period of approximately 50 years starting in calendar 2048. No barriers to implementation in a reasonable timeframe have been identified from this generic analysis. Engineering challenges include: 1) shaft construction; 2) a very-large capacity shaft hoist; 3) overpack design; 4) a transport-emplacement-vehicle (TEV) for transporting waste packages once they are underground and emplacing them remotely; and 5) remotely operated equipment for emplacing backfill. The method of shaft construction would depend on site-specific conditions, and could involve freezing the subsurface. A shaft hoist with payload capacity of 175 MT seems technically and economically feasible based on development work in Germany, and it would be the largest hoist of its kind. The function of disposal overpacks would be to provide reliable containment during repository operations, which could be accomplished using a corrosion allowance material such as a low-carbon steel. Development effort would be needed to determine overpack thickness (e.g., 7 to 20 cm) that can resist corrosion and loading from salt creep, to rovide containment throughout the repository operational period. The transport-emplacement vehicle (TEV) would be similar to previous concepts, particularly one option proposed for a Yucca Mountain repository. It would move over a rough salt floor on independently driven and steered wheels, and carry heavy shielding in addition to a waste package. By analogy to the safety case for the WIPP in New Mexico, human intrusion is likely to be the dominant mode of radionuclide release from the repository. Treatment of human intrusion for a CSNF repository in salt could depend on promulgation of site-specific changes in the regulations. Radionuclide release and migration would be quite limited for undisturbed conditions. There may be opportunities for improved understanding of salt performance with waste heating, based on future in situ testing in an underground salt research laboratory. This report also discusses a developing area of salt rock mechanics that involves low-stress, low strain-rate creep that might cause large, heavy waste packages to slowly sink. Site-specific sampling, testing, and modeling would be used to determine if the mechanism is important enough to merit consideration in design, or inclusion in performance assessment. Part of engineering design and postclosure safety assessment for a CSNF repository in salt would be to implement a methodology to show that the probability of a criticality event in the repository, when waste packages eventually breach and are flooded, is less than the probability screening threshold for performance assessment. In the methodology, a criticality analysis would be performed for waste packages in the repository, incorporating measures that could be introduced as needed to limit reactivity, for example using fuel selection and loading rules, and crediting the absorption of thermal neutrons by natural chlorine in the environment. A similar analysis has been underway for CSNF stored in dual-purpose canisters. Ideally the strategy would be developed prior to actually loading SNF assemblies into canisters used in waste packages for disposal.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Effect of Clad Fast Neutron Flux Distribution on Quarter-Core Fuel Performance Calculations with BISON

As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Virtual Environment for Reactor Applications (VERA) is being developed to provide high-fidelity multiphysics simulations of nuclear reactor cores. Over the past year, a one-way coupling capability has been developed, taking power and temperature data from MPACT/CTF calculations and running standalone BISON fuel rod cases. This has allowed CASL contributors to gain insight into the fuel performance characteristics while a more tightly coupled methodology between MPACT, CTF, and BISON has been under development. Many of the initial analyses have focused on the Watts Bar Unit 1 core using a fast flux factor—the default approach—which defines the clad fast flux as a linear function of the local linear heat rate. While this approach is more valid when considering single rod cases in isolation, in larger problems such as full-sized reactors, the local linear heat rate does not directly correlate to the clad fast neutron flux. For example, variations in assembly enrichment can lead to larger neutron fluxes in neighboring assemblies that may be at lower enrichment and power. Similarly, considering axial effects, spacer grids suppress power locally, but the fast neutron flux is significantly less affected, and the approach can lead to an underestimation of the fast neutron flux at that time. A poor estimation of the clad fast flux values can affect the clad creep rate, fuel-clad gap, and clad stresses. Additionally, the fast flux factor has been considered constant throughout the life of the rod. As the cycle progresses, the spectrum will harden as burnable absorbers and 235U deplete, so the fast flux would be expected to increase as the spectrum hardens. The effect on quantities of interest are highlighted for the first cycle of the Watts Bar Unit 1 core, comparing the results from an explicit clad fast flux representation from MPACT to the constant fast flux factor approach used in previous analyses. Future work will consider multicycle effects to assess the impact at higher burnups.

Stimpson, Shane G.↗
DOE OSTI2025

Strength, deformation, and the fcc–hcp phase transition in condensed Kr and Xe to the 100 GPa pressure range

The rare gas solids exhibit systematic differences in crystal structure, phase transition conditions, bond strength, and other physical properties. The physical properties of heavy rare gas solids krypton and xenon are modified by the martensitic phase transition from face-centered cubic to hexagonal close packed structure over a broad pressure range. Crystal structure, strength, and plastic deformation of krypton and xenon have been investigated at 300 K using compression in the diamond-anvil cell with synchrotron angle-dispersive x-ray diffraction and complementary ruby fluorescence spectroscopy for Xe. Stacking faults indicative of the fcc–hcp phase transition are observed at pressures at and above 1.23 ± 0.05 and 1.9 ± 0.6 GPa in Kr and Xe, respectively. The transition remains incomplete in both solids to pressures greater than 100 GPa. Strength determined from stress measurements in Pt and ruby standards at pressures up to 111 GPa and complemented by observations of strain and texture measurements obtained by x-ray diffraction in the radial geometry to 100 GPa indicates similar or higher strength than Ar at all conditions, with significant stiffening at 15–20 GPa. Radial diffraction data reveal the persistence of broad highly textured fcc diffraction lines to 101 GPa in Xe, suggesting that the axial measurements may underestimate the metastable persistence of the fcc phase due to biased sampling of hcp crystallites resulting from preferred crystallite orientation. Kr and Xe are compared with He, Ne, and Ar for a systematic understanding of physical properties and phase equilibria of rare gas solids.

Compressive stress↗
DOE OSTI2020

Neutron Diffraction Analysis of Residual Strain in High-Pressure Die Cast A383 Engine Blocks

The residual strains in three high-pressure die cast A383 engine blocks were measured using neutron diffraction. Residual strains and stresses can be generated during high-pressure die casting process due to high thermal gradients and may cause fatigue failure or dimensional distortion. Neutron diffraction can measure the residual strain distributions deep inside castings with complex geometries, because neutrons can penetrate substantially more material than x-rays. In this paper, the residual strains (in the axial and hoop directions) along the whole cylinder bridge of high-pressure die cast A383 engine blocks in as-cast and heat-treated conditions were measured by neutron diffraction methods. The effect of heat treatment on the distribution and magnitude of residual strains was also investigated. The results indicate that the residual strain in the cylinder bridge is tensile in both the axial and hoop components for both the as-cast and the heat-treated engine blocks. The residual strain in the hoop direction is bigger in magnitude than the residual strain in the axial direction. Heat treatment significantly relaxes the residual strain in the axial direction but only slightly relaxes the residual strain in the hoop direction.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Design of a Biomass Scale Cubical Triaxial Tester

The bioenergy industry is dependent on predictable and reliable feedstock handling equipment. High profile failures in the industry demonstrate the need for better modeling tools for feedstock handling, specifically tools which can model the bulk flow of biomass feedstock materials in handling equipment such as hoppers or screw conveyors. Measurement of bulk material flow characteristics is a critical component to inform modeling tools. Following examples in the soils and pharmaceuticals industries, a Cubical Triaxial Tester (CTT) was selected as a critical component to study relationships between applied stress on a bulk sample and the resulting strain on the bulk material. In cooperation with The Pennsylvania State University, Forest Concepts designed a CTT specifically to measure bulk biomass flowability characteristics. Several mechanical design constraints were identified to ensure proper scaling and repeatability. Major design constraints included sample chamber size versus particle length, volumetric strain sensing accuracy, independent axial pressure control, and sample chamber handling. This report explains how major design constraints were determined.

09 BIOMASS FUELS↗
DOE OSTI2025

Assessment of irradiation embrittlement effect on fatigue life of a pressurized-water reactor pressure vessel using the fracture toughness master curve approach

The reactor pressure vessel (RPV) is a critical structural component in pressurized-water reactors, and it is designed to withstand extreme conditions, such as high pressures, elevated temperatures, and prolonged radiation exposure. Ensuring RPV integrity is essential for the safe and reliable long-term operation of nuclear power plants, especially as aging mechanisms such as fatigue and irradiation embrittlement pose increased risks. Fatigue, caused by cyclic thermal and mechanical loading, can lead to crack initiation in localized high-stress regions. Simultaneously, neutron irradiation, particularly in the beltline region, progressively reduces fracture toughness, increasing susceptibility to brittle fracture. These combined effects of fatigue and irradiation embrittlement potentially impact the RPV structural integrity, necessitating fitness-for-service assessments. This study applies the fracture toughness master curve approach to evaluate the impact of irradiation embrittlement on RPV fatigue life. A coupled thermo-mechanical stress analysis identifies critical stress locations under normal service transients, pinpointing regions most vulnerable to fatigue crack initiation and growth. Stress intensity factors for postulated flaws at these locations are calculated, enabling an assessment of fatigue life under irradiated and unirradiated conditions. The results indicate that neutron irradiation embrittlement accelerates the conditions in which a critical crack can form and lead to failure, particularly at lower temperatures. The failure occurs where reduced fracture toughness limits the material’s resistance to crack growth. Axial cracks at nozzle corners are the most life limiting without irradiation, while circumferential cracks demonstrate longer fatigue lives. The findings highlight the importance of incorporating irradiation effects into fatigue life predictions to ensure the long-term structural integrity of RPVs.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermal Modeling and Limitations for Power Electronics Embedded in Medium-Voltage Cables

As next-generation energy technologies gain traction and power demand increases, the existing electrical infrastructure faces significant stress, prompting innovative solutions to enhance the grid's capacity and lifespan. This work explores the possibility of embedding medium-voltage (MV) power electronics directly inline with the cable, and the resulting thermal challenges. Since the majority of power distribution cables installed in the U.S. are passively cooled, the work focuses primarily on passive cooling, with an emphasis on the limitations of axial heat spreading within the cable. To date, literature on axial spreading of high incident heat loads on cables and cable environments is limited, typically reporting cases with <10 W of incident heat load. This work will explore the considerations, limits, and tradeoffs of cable-embedded heat loads significantly larger than the cable losses. Both external and internal effects are modeled analytically in nondimensional terms via a Biot number analysis, allowing fundamental limits and tradeoffs to be derived. The work culminates in the design and experimental validation of a cable-embedded thermal system capable of passively dissipating 300 W of heat from a coaxial SiC mosfet switch module over a length of 20 cm, thus validating the possibility of MV cable-embedded power electronics from a thermal standpoint.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2025

Spectral proper orthogonal decomposition of active wake mixing dynamics in a stable atmospheric boundary layer

Recent advancements in the use of active wake mixing (AWM) to reduce wake effects on downstream turbines open new avenues for increasing power generation in wind farms. However, a better understanding of the fluid dynamics underlying AWM is still needed to make wake mixing a reliable strategy for wind farm flow control. In this work, a spectral proper orthogonal decomposition (SPOD) is used to analyze the dynamics of coherent flow structures that are induced in the wake through blade pitch actuation. The data are generated using the ExaWind software suite to perform large eddy simulations of an NREL 2.8 MW turbine operating in a stable atmospheric boundary layer. SPOD tracks the modal behavior of flow structures from their generation in the turbine induction field through their growth in the near-wake region and to their subsequent evolution and energy transfers in the far wake. SPOD is shown to be a useful tool in the context of AWM because it translates the wavenumber and frequency inputs to the turbine controller to structures in the wake. A decomposition of the radial shear stress flux in the wake is also developed using SPOD to measure the contribution of coherent flow structures to mean flow turbulent entrainment and wake recovery. The effectiveness of AWM is connected to its ability to excite inherent structures in the wake of the turbine that arise using baseline controls. The effects of AWM on blade loading are also analyzed by connecting the axial force along the blade to the SPOD analysis of the turbine induction field. Lastly, the performance of different AWM strategies is demonstrated in a two-turbine array.

17 WIND ENERGY↗
DOE OSTI2020

Mapping of Texture and Phase Fractions in Heterogeneous Stress States during Multiaxial Loading of Biomedical Superelastic NiTi

Thermoelastic deformation mechanisms in polycrystalline biomedical-grade superelastic NiTi are spatially mapped using in situ neutron diffraction during multiaxial loading and heating. The trigonal R-phase is formed from the cubic phase during cooling to room temperature and subsequently deforms in compression, tension, and torsion. The resulting R-phase variant microstructure from the variant reorientation and detwinning processes are equivalent for the corresponding strain in tension and compression, and the variant microstructure is reversible by isothermal loading. The R-phase variant microstructure is consistent between uniaxial and torsional loading when the principal stress directions of the stress state are considered (for the crystallographic directions observed here). The variant microstructure evolution is tracked and the similarity in general behavior between uniaxial and torsional loading, in spite of the implicit heterogeneous stress state associated with torsional loading, pointed to the ability of the reversible thermoelastic transformation in NiTi to accommodate stress and strain mismatch with deformation. This ability of the R-phase, despite its limited variants, to accommodate stress and strain and satisfy strain incompatibility in addition to the existing internal stresses has significance for reducing irrecoverable deformation mechanisms during loading and cycling through the phase transformation.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Experimental investigation of cross flow mixing in a randomly packed bed and streamwise vortex characteristics using particle image velocimetry and proper orthogonal decomposition analysis

The study of flow and heat transfer through porous media or randomly packed beds is important as these configurations are widely used in many engineering applications, for example, heat energy storage, chemical catalytic reactors, and nuclear reactors. The flow mixing characteristics in a cross-flow plane of a facility with randomly packed spheres at an aspect ratio of 6.3 were experimentally investigated. The velocity fields at several regions of the cross-flow plane located in the vicinity of the wall and in the pores between spheres were obtained by applying the matched-index-of-refraction and time-resolved particle image velocimetry (TR-PIV) techniques for Reynolds numbers ranging from 700 to 1700. The TR-PIV results revealed various flow patterns in the transverse plane of the packed spheres, including swirling flow structures aligned with the axial flow direction, a strong bypass flow near the enclosure wall, and a circulation region created when the bypass flow ejected into a large spatial gap. When the Reynolds number was increased, the peaks of root-mean-square fluctuating velocities, $u'_{rms}$ and $v'_{rms}$, were found to increase approximately at the same ratio as the increase in Reynolds number, and the magnitude of the Reynolds stress increased considerably. In addition, the characteristics of flow mixing in different flow regions were investigated via the two-point cross-correlation of fluctuating velocities. Using Taylor’s hypothesis, the vorticity iso-surfaces were constructed. Thus, constructed iso-surfaces showed that shear layers generated from the bypass flow gaps were stretched, broken into smaller flow structures, and then evolved as vortex pairs when entering the neighboring gaps. The results obtained by applying proper orthogonal decomposition (POD) analysis to the velocity fields showed that the statistically dominant flow structures had approximately the same size and shape as those depicted by Taylor’s hypothesis. Vortex characteristics, such as populations, spatial distributions, and strengths, for various spatial regions and Reynolds numbers were obtained by a combination of POD analysis and vortex identification.

42 ENGINEERING↗
DOE OSTI2025

On the bulk compaction of brittle granular materials, part I: SeS analysis of axial compression to 4000 MPa

The bulk compaction of granular materials has been studied for decades to interpret and manage responses for soils and powder-based component fabrication, and geophysical, celestial, and ballistic impact. Their bulk or macroscopic compaction response is limited by what occurs at the granular or microstructural scale. Motivation existed to more closely examine that association specific to granular brittle materials (e.g., ceramics and glasses). That examination is offered in a series of three companion papers where Part I describes a new supplemental analysis adopted to bulk compaction response involving relatively high compaction stresses (4000 MPa). Bulk compactions of vitreous silicates and crystalline quartzes were interpreted in three ways, including that of a new analysis that considers the product of void ratio (e) and stress (S) as a function of S, hereafter referred to as “SeS analysis”. In conclusion, the SeS analysis was found to be an informative supplement to conventional bulk compaction analyses because it provides more consistent higher sensitivity for the identification of bulk density rate increase with increasing compaction (softening); a rate increase that arises from the cumulative effect of the onsets and progression of compaction-induced yielding, fracture or comminution, densification, phase change, or combinations thereof occurring at the granular or microstructural scale.

compaction↗
DOE OSTI2024

Uniaxial compressive creep tests by spark plasma sintering of 70% theoretical density α -uranium and U-10Zr

Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6±1.6 and 5.7±1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6±1.1kJ/mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗