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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 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Creating free standing covalent organic framework membranes by nanocrystal suturing in sol gel solutions

The sol-gel synthesis represents a versatile platform to fabricate ceramic inorganic membranes. However, it is still a grand challenge to push the boundary of sol-gel chemistry towards high-quality organic membrane construction. Herein, a facile and controlled nanocrystal suturing strategy in sol-gel solutions is developed to afford highly crystalline and free-standing covalent organic framework membranes. The key chemistry design lies in deploying tiny threads (1 mol% dual-NH 2 -tail linear polymer) to efficiently suture the highly charged covalent organic framework nanocrystals stabilized and confined in sol-gel solutions, creating a continuous and intact membrane surface. A subsequent treatment heals the sutured covalent organic framework nanocrystals, yielding a free-standing membrane with high crystallinity and ordered pores. The structure evolution and role of the thread linker are elucidated via operando spectroscopy and microscopy. The as-afforded covalent organic framework membranes demonstrate attractive proton transport performance in high temperature and anhydrous fuel cell applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Energy-resolved and time-dependent unimolecular dissociation of hydroperoxyalkyl radicals (˙QOOH)

Hydroperoxyalkyl radicals (˙QOOH) are transient intermediates in the atmospheric oxidation of volatile organic compounds and combustion of hydrocarbon fuels in low temperature (<1000 K) environments. The carbon-centered ˙QOOH radicals are a critical juncture in the oxidation mechanism, but have generally eluded direct experimental observation of their structure, stability, and dissociation dynamics. Recently, this laboratory demonstrated that a prototypical ˙QOOH radical [˙CH 2 (CH 3 ) 2 COOH] can be synthesized by an alternative route, stabilized in a pulsed supersonic expansion, and characterized by its infrared (IR) spectroscopic signature and unimolecular dissociation rate to OH radical and cyclic ether products. The present study focuses on a partially deuterated ˙QOOD analog ˙CH 2 (CH 3 ) 2 COOD, generated in the laboratory by H-atom abstraction from partially deuterated tert-butyl hydroperoxide, (CH 3 ) 3 COOD. IR spectral features associated with jet-cooled and isolated ˙QOOD radicals are observed in the vicinity of the transition state (TS) barrier leading to OD radical and cyclic ether products. Furthermore, the overtone OD stretch (2ν OD ) of ˙QOOD is identified by IR action spectroscopy with UV laser-induced fluorescence detection of OD products. Direct time-domain measurement of the unimolecular dissociation rate for ˙QOOD (2ν OD ) extends prior rate measurements for ˙QOOH. Partial deuteration results in a small increase in the TS barrier predicted by high level electronic structure calculations due to changes in zero-point energies; the imaginary frequency is unchanged. Comparison of the unimolecular decay rates obtained experimentally with those predicted theoretically for both ˙QOOH and ˙QOOD confirm that unimolecular decay is enhanced by heavy-atom tunneling involving simultaneous O–O bond elongation and C–C–O angle contraction along the reaction pathway.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Influence of misfit dislocations on ionic conductivity at oxide interfaces

Mismatched complex oxide thin films and heterostructures have gained significant traction for use as electrolytes in intermediate temperature solid oxide fuel cells, wherein interfaces exhibit variation in ionic conductivity as compared to the bulk. Although misfit dislocations present at interfaces in these structures impact ionic conductivity, the fundamental mechanisms responsible for this effect are not well understood. To this end, a kinetic lattice Monte Carlo (KLMC) model was developed to trace oxygen vacancy diffusion at misfit dislocations in SrTiO 3 /BaZrO 3 heterostructures and elucidate the atomistic mechanisms governing ionic diffusion at oxide interfaces. The KLMC model utilized oxygen vacancy migration energy barriers computed using molecular statics. While some interfaces promote oxygen vacancy diffusion, others impede their transport. Fundamental factors such as interface layer chemistry, misfit dislocation structure, and starting and ending sites of migrating ions play a crucial role in oxygen diffusivity. Molecular dynamics (MD) simulations were further performed to support qualitative trends for oxygen vacancy diffusion. Overall, the agreement between KLMC and MD is quite good, though MD tends to predict slightly higher conductivities, perhaps a reflection of nuanced structural relaxations that are not captured by KLMC. The current framework comprising KLMC modeling integrated with molecular statics offers a powerful tool to perform mechanistic studies focused on ionic transport in thin film oxide electrolytes and facilitate their rational design.

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

The five line-of-sight neutron time-of-flight (nToF) suite on the National Ignition Facility (NIF)

Measurement of the neutron spectrum from inertial confinement fusion implosions is one of the primary diagnostics of implosion performance. Analysis of the spectrum gives access to quantities such as neutron yield, hot-spot velocity, apparent ion temperature, and compressed fuel ρr through measurement of the down-scatter ratio. On the National Ignition Facility, the neutron time-of-flight suite has been upgraded to include five independent, collimated lines of sight, each comprising a high dynamic range bibenzyl/diphenylacetylene-stilbene scintillator [R. Hatarik et al., Plasma Fusion Res. 9, 4404104 (2014)] and high-speed fused silica Cherenkov detectors [A. S. Moore et al., Rev. Sci. Instrum. 89, 10I120 (2018)].

Moore, A. S. (ORCID:0000000221258882)↗
DOE OSTI2021

High yields in direct-drive inertial confinement fusion using thin-ice DT liner targets

In this work, we report simulation results that suggest record neutron yields above 3 x 10 14 can be achieved in laser-driven fusion on OMEGA by imploding large-outer-diameter targets (~1010μm) containing ice layers as thin as 34μm. Historically, the performance of thin and fast capsules is thought to be highly degraded by hydrodynamic instabilities. We present pulse designs that push implosion velocities above 650 km/s while maintaining enough ablative stabilization and low enough in-flight aspect ratios to survive the acceleration phase. We show these targets also possess improved stability during deceleration stemming from augmented core temperatures and low fuel convergence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Observation of impurity accumulation and its compatibility with high plasma performance in W7-X

At the W7-X stellarator, the bolometer system has measured an intensive radiation zone in the inner plasma region (at a normalized radius $ρ$ ~ 0.3–0.4) in the hydrogen plasma generated by electron cyclotron resonance heating; it differs from the normal plasma radiation distribution with an edge-localized emission zone. Spectroscopic diagnostics have recorded high-Z elements such as iron. This phenomenon happens in the plasma phases after gas supply turn-off, which results in all impurity relevant diagnostic signals increasing for several seconds. Despite the enhancement of the core radiation, the plasma energy confinement is improved. A transport analysis shows that this impurity radiation behavior is associated with a low diffusion coefficient ($D$~ 0.02 m 2 s -1 ) and a reversal of the convection around the radial position of the emission peak, which, under normal conditions, separates the zones of outward convection in the central (|$V$| ~ 0.1 m s -1 ) and inward convection in the outer region (|$V$| ~ 0.3 m s -1 ). An impurity accumulation around this radial position has been identified. The transport coefficients obtained are comparable with the theoretical predictions of collisional impurity transport. In the plasma phases studied, both impurity and energy confinement are enhanced. The mechanism responsible for the improvement is believed to be a reduction of micro-instabilities associated with the observed steepening of the density profile, initiated by a low edge plasma density (<1.0 × 10 19 m -3 ) after switching off the gas fueling. The normalized temperature and density gradients fulfil the condition for the suppression of ITG turbulence.

suppression of ITG turbulence↗
DOE OSTI2020

Summary of the Postirradiation Examination of the First Samples from a MiniFuel Irradiation

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly screen different nuclear fuel concepts. Observations on fission gas release, irradiation conditions and microstructure of the irradiated fuel show good fuel performance at the low burnup achieved in this initial irradiation. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU as was achieved after only 68 effective full power days of irradiation. Results from silicon carbide thermometry measurements further benchmarked the MiniFuel irradiation vehicle and indicated fuel was irradiated at temperatures ranging from 410-460°C. Extensive microstructural characterization on the irradiated fuel was performed and no significant irradiation induced changes were observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

An Integrated High Pressure SOFC and Premixed Compression Ignition (PCI) Engine System

Components for a transformative, high pressure, intermediate temperature solid oxide fuel cell (SOFC) and premixed compression ignition (PCI) engine system were developed. The resulting system is a 1 MWe natural gas fueled hybrid SOFC and reciprocating internal combustion engine system capable of achieving over 70% electrical efficiency at a cost of $0.87/W. To achieve 70% electric efficiency, the SOFC system operates at 80% fuel utilization; however, when operating at this high level of fuel utilization, fuel to the engine (anode tailgas) is extremely dilute. Accordingly, a substantial portion of the effort focused on development and assessment of the combustion strategy capable of consuming this dilute fuel mixture. The engine testing effort showed that pre-chamber assisted compression ignition provides the best combination of controllability, efficiency and emissions. The engine experiments showed that high efficiency combustion can be achieved with near zero criteria pollutants. It is likely that the engine system could operate with minimal aftertreatment requirements, reducing complexity, cost, and precious metal usage. System level modeling was used to assess the overall efficiency, determine cost and efficiency tradeoffs, and demonstrate the potential to perform load following with the hybrid system. Finally, advanced manufacturing was used to demonstrate the potential to fabricate heat exchangers needed for the hybrid system from low chromium materials, which addresses the issue of chromium poisoning of the SOFC stack material.

03 NATURAL GAS↗
DOE OSTI2024

Uranium Mononitride (UN) Handbook

Uranium Mononitride (UN) is being reexamined as a high temperature advanced nuclear fuel in many reactor applications due to its high thermal conductivity, high melting point, and high fissionable uranium density, among other desirable properties. A thorough review of the experimental data for the material and irradiation properties has been collected and is presented in this handbook. As an outcome of this review, in many cases, it is apparent that additional experimental verification must be conducted to verify the available data and to verify the suggested empirical correlations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Prediction of Silicon Content in a Blast Furnace via Machine Learning: A Comprehensive Processing and Modeling Pipeline

Silicon content plays an important role in determining the operational efficiency of blast furnaces (BFs) and their downstream processes in integrated steelmaking; however, existing sampling methods and first-principles models are somewhat limited in their capability and flexibility. Current data-based prediction models primarily rely on a limited set of manually selected furnace parameters. Additionally, different BFs present a diverse set of operating parameters and state variables that are known to directly influence the hot metal’s silicon content, such as fuel injection, blast temperature, and raw material charge composition, among other process variables that have their own impacts. The expansiveness of the parameter set adds complexity to parameter selection and processing. This highlights the need for a comprehensive methodology to integrate and select from all relevant parameters for accurate silicon content prediction. Providing accurate silicon content predictions would enable operators to adjust furnace conditions dynamically, improving safety and reducing economic risk. To address these issues, a two-stage approach is proposed. First, a generalized data processing scheme is proposed to accommodate diverse furnace parameters. Second, a robust modeling pipeline is used to establish a machine learning (ML) model capable of predicting hot metal silicon content with reasonable accuracy. The method employed herein predicted the average Si content of the upcoming furnace cast with an accuracy of 91% among 200 target predictions for a specific furnace provisioned by the XGBoost model. This prediction is achieved using only the past shift’s operating conditions, which should be available in real time. This performance provides a strong baseline for the modeling approach with potential for further improvement through provision of real-time features.

Chemistry↗
DOE OSTI2022

Design of inertial fusion implosions reaching the burning plasma regime

One of the last remaining milestones in fusion research before reaching ignition is creating a burning plasma state, where alpha particles from deuterium-tritium (DT) fusion reactions redeposit their energy as the dominant source of heating in the plasma. The indirect-drive inertial confinement fusion approach at the National Ignition Facility (NIF) uses a laser-generated radiation cavity (hohlraum) to spherically implode DT fuel to high temperatures and densities in a central ”hot spot”. Here, we deliver more energy to the hot spot than ever before, while maintaining the extreme pressures required for inertial confinement, by increasing the size of the implosion compared to previous experiments. We develop more efficient hohlraums, to drive these larger implosions within NIF’s current laser energy and power capability and control symmetry by moving energy between laser beams and by changing the shape of the hohlraum. These designs resulted in record fusion powers of 1.5 petawatts, greater than the input power of the laser, and 170 kJ of fusion energy. Radiation hydrodynamics simulations show alpha particle heating as the dominant term in the hot spot energy balance, e.g. a burning plasma state. This work is expected to motivate future studies of burning plasmas and improve predictive capability by providing a benchmark for modeling used to understand the proximity to ignition.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

This is a conference presentation ppt file. The following abstract is for the entire conference paper - which will later be the subject of a PRS artifact The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Catalyst coating of a perovskite film and particles exsoluted from the perovskite film

A hybrid catalyst coating composed of a conformal thin film with exsoluted PrOx nano-particles. The conformal PNM thin film can be a perovskite composition of PrNi0.5Mn0.5O3 (PNM). The PrOx nano-particles dramatically enhance the oxygen reduction reaction kinetics via a high concentration of oxygen vacancies while the thin PNM film effectively suppresses strontium segregation from the cathode of an intermediate-temperature solid oxide fuel cell.

Liu, Meilin↗
DOE OSTI2023

Modeling and Simulation of Fuel Burnup in Pebble Bed Reactors

Modeling and simulation of fuel burnup plays important roles in reactor design, operation, safety, and security as well as nuclear material control and accounting (MC&A) [1]. This task is uniquely challenging for pebble bed reactors (PBR) because the pebbles are continuously added and recycled into the reactor core, and their paths through the core are random. To address this problem, we present two simulation models in this paper. Brookhaven National Laboratory (BNL) developed a simple lattice model of a PBR in Serpent software to generate used pebble isotopic concentrations. The benefit of using Serpent software in this specific application is that it helps streamline the data generation process without having to use too many independent software codes in combination to achieve a simple task. For example, transport, burnup and zero power decay can be implemented in a single pass. Three-dimensional core models were developed using Serpent to simulate the burnup process of 5 subject pebbles starting from fresh till they reach nearly target burnup, with each pebble placed in one of the five artificially designated radial channels to capture the changing neutron spectra along the core radius. Equilibrium isotopic concentrations were assumed in all other pebbles in the core. To provide a verification for the Serpent isotope transmutation and decay results, Oak Ridge National Laboratory (ORNL) performed simple SCALE/ORIGEN calculations using the average neutron spectra calculated by Serpent for each of the 5 pebbles. The 252-group neutron spectra from Serpent were then used by ORIGEN to produce the one-group library for depletion and decay calculations. The isotopic concentrations of a few nuclides of interest and neutron and gamma source terms produced from the ORIGEN calculations were compared with the ones from the Serpent calculations. The model simulated in this work was based on the Pebble Bed Modular Reactor (PBMR)-400 design because many data needed for the simulation such as core power profiles, fuel and reflector temperatures, and equilibrium core composition are publicly available. In this paper, we will compare the results between these two approaches and benchmark the results against a set of well-established simulation results for PBMR-400.

Dim, Odera↗
DOE OSTI2023

Modeling and Simulation of Fuel Burnup in Pebble Bed Reactors

Modeling and simulation of fuel burnup plays important roles in reactor design, operation, safety, and security as well as nuclear material control and accounting (MC&A) [1]. This task is uniquely challenging for pebble bed reactors (PBR) because the pebbles are continuously added and recycled into the reactor core, and their paths through the core are random. To address this problem, we present two simulation models in this paper. Brookhaven National Laboratory (BNL) developed a simple lattice model of a PBR in Serpent software to generate used pebble isotopic concentrations. The benefit of using Serpent software in this specific application is that it helps streamline the data generation process without having to use too many independent software codes in combination to achieve a simple task. For example, transport, burnup and zero power decay can be implemented in a single pass. Three-dimensional core models were developed using Serpent to simulate the burnup process of five subject pebbles starting from fresh till they reach nearly target burnup, with each pebble placed in one of the five artificially designated radial channels to capture the changing neutron spectra along the core radius. Equilibrium isotopic concentrations were assumed in all other pebbles in the core. To provide a verification for the Serpent isotope transmutation and decay results, Oak Ridge National Laboratory (ORNL) performed simple SCALE/ORIGEN calculations using the average neutron spectra calculated by Serpent for each of the five pebbles. The 252-group neutron spectra from Serpent were then used by ORIGEN to produce the one-group library for depletion and decay calculations. The isotopic concentrations of a few nuclides of interest and neutron and gamma source terms produced from the ORIGEN calculations were compared with the ones from the Serpent calculations. The model simulated in this work was based on the Pebble Bed Modular Reactor (PBMR)-400 design because many data needed for the simulation such as core power profiles, fuel and reflector temperatures, and equilibrium core composition are publicly available. In this paper, we will compare the results between these two approaches and benchmark the results against a set of well-established simulation results for PBMR-400.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Stability margin analysis of the Holos-quad microreactor design

A simplified point model of the Holos-Quad microreactor is introduced. The model is based on the point kinetics equations coupled to three heat balance equations, representing the mean temperatures of the fuel particles, the graphite moderator and the helium coolant. The differential equations are converted to the frequency domain, enabling the construction of the closed-loop reactor transfer function. Using this function the stability margins of the core design is analyzed for various power levels. It is shown that the gain margins approaches infinity, demonstrating the stability of the reactor for all power levels. The phase margin at nominal power is about 60 degrees, however it shows a non-monotonous dependence on power, with minimal value obtained for about 10% of nominal power. As power level is further increased, the phase margin also increases, demonstrating the reactor becomes more stable. This behavior may be of high importance in load-follow scenarios, where the power level of the reactor changes with time. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

DoD Symposium Presentation

TRISO fuel, or TRi-structural ISOtropic, is a ceramic based nuclear fuel capable of operating at high temperatures (up to 1600°C). The fuel, consisting of uranium oxycarbide (UCO) fuel kernels, is coated with three layers of carbon and ceramic (dense silicon carbide) materials to capture and contain radioactive fission products. BWXT, located in Lynchburg, VA, has perfected the manufacturing techniques to produce this fuel and is currently the only US company licensed to produce this irradiation-tested fuel. BWXT is currently producing fuel in support of a demonstration reactor scheduled for startup on the INL site in 2025. The first fuel delivery, estimated at 200 kilograms High Assay Low Enriched Uranium, is targeted for the end of CY2024.

07 ISOTOPE AND RADIATION SOURCES↗