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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Transition of Spent Nuclear Fuel to Dry Storage: Modeling activities concerning aluminum spent nuclear fuel cladding integrity

In 2017, a report was prepared by the Spent Nuclear Fuel Working Group in 2017 and titled “Technical Considerations and Challenges for Extended (>50 Years) Dry Storage” [1-2]. In this report several key scientific issues were identified pertaining to the problem of aluminum spent nuclear fuel (ASNF) and its safe removal from wet storage with subsequent drying and transfer to dry storage at the Department of Energy (DOE) Idaho National Laboratory (INL). This report outlines two modeling analyses performed for aluminum and aluminum oxides. The first analysis examined the solubility of aluminum oxides as a function of temperature and pH. This report presents existing literature produced as far back as the 1950’s and including significant analysis and reporting performed at Alcoa in the 1980’s. The primary technical gap was the lack of solubility data beyond that reported at standard temperature (25°C). The analysis examined these solubility curves as a function of temperature as well as phase type. The pH value of minimum solubility shifted with temperature but was insensitive to the oxide phase type. The solubility increases with temperature as would be expected, where an order of magnitude increase was observed from 25°C and 90°C. This data provides greater information to understand the various conditions experiences by ASNF including in dry storage in cases where water filming may occur. A second thermodynamic analysis was performed to examine the aluminum phase behavior as a function of temperature. This was performed to understand limitations and possible problems with fuel drying at elevated temperatures. Based on what has been established in previous reports, water remains chemically entrained within boehmite until temperatures exceeding 400°C. Analysis showed that significant phase changes occurred with highly alloyed AA6061 and AA5052 while the more pure aluminum (AA1100) showed nearly ideal behavior (i.e. limited phase changes). In the case of the more alloyed AA5052 and AA6061, the phase changes could take place at relatively low temperatures (100°C to 250°C), and thus may have already experienced some of these phase changes in service. As results are thermodynamic in nature, assessment of kinetics will be examined to determine if the phase changes are a concern for the relatively short fuel drying step.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Thermal Modeling of a Loaded Magnastor Storage System at Catawba Nuclear Station

This report represents fulfillment of the M2 Milestone M2FT-16PN080203031, “High Heat Load Thermal Analyses,” under Work Package FT-16PN08020303. The report documents the analysis of the NAC International, Inc. (NAC) Modular, Advanced Generation, Nuclear All-purpose STORage (MAGNASTOR) module at Duke’s Catawba Nuclear Station. As part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, a high-burnup fuel storage demonstration is planned for a storage module in the North Anna Nuclear Station’s Independent Spent Fuel Storage Installation (ISFSI). The storage module selected for this demonstration is an AREVA TN-32B High Burnup cask. The main goals of this proposed test are to provide confirmatory data for model validation and potential improvement, support license renewals and new licenses for ISFSIs, and support transportation licensing for high burnup spent nuclear fuel (EPRI 2014). The focus of the Demonstration test is the performance of the high-burnup fuel. To support the High Burnup Spent Fuel Data Project, Pacific Northwest National Laboratory (PNNL) staff performed a detailed thermal analysis (Fort et al. 2016) of the TN-32B cask. It was found that when conservatisms were removed from the thermal models and especially from the calculated decay heats, the predicted peak cladding temperatures (PCTs) were significantly below the 400°C (752°F) regulatory guidance limit, as specified in Interim Staff Guidance 11, Revision 3, (U.S. NRC 2003).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ROADRUNNER MiniFuel Experiment: Irradiation Target Design and Sample Characterization

High-density uranium nitride (UN) is a fuel candidate for several advanced nuclear reactor designs currently under development. Because there are limited UN performance data relative to fuel fabrication impurity and density variation, an irradiation campaign has been developed as part of a collaborative effort among the University of Texas at San Antonio (UTSA), Westinghouse Electric Company, Oak Ridge National Laboratory (ORNL), and Los Alamos National Laboratory (LANL) under the Nuclear Science User Facilities program. This project, entitled ROADRUNNER, or Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments, aimsto support UN fuel qualification for advanced reactors by investigating the impact of density and impurity variations on UN performance as a function of irradiation temperature and burnup. The MiniFuel experiment vehicle developed by ORNL, which leverages the High Flux Isotope Reactor, was selected to perform this accelerated separate-effects irradiation testing. The experiment test matrix consists of six MiniFuel targets containing miniature UN fuel disks, and targets three distinct burnup levels (37.5, 60, and 75 MWd/kg U) and three distinct temperatures (600, 900, and 1200°C). Neutronics and thermal analyses were performed to determine the experimental parameters needed to meet the desired irradiation conditions and to predict the experiment components temperatures. UN pellets were fabricated at LANL with tightly controlled parameters to produce specimens with three distinct densities and three levels of carbon content. The pellets were then thinned down by UTSA to the experiment-required thickness. The pre-characterization of the specimens includes density measurements, carbon and oxygen contents, microstructure analysis, and x-ray computed tomography. The selected specimens will be assembled into the MiniFuel experiment, and the first ROADRUNNER MiniFuel targets are intended for HFIR insertion during the Fall of 2024. After irradiation, the targets will be shipped to ORNL’s hot cell facility for disassembly. The post-irradiation examination on the fuel specimens includes fission gas release measurements, visual inspection, fuel swelling measurements, gamma spectroscopy, and microstructure analysis. The data collected post-irradiation will be used to develop fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2005

A Materials Compatibility and Thermal Stability Analysis of Common Hydrocarbon Fuels

A materials compatibility and thermal stability investigation was conducted using five common liquid hydrocarbon fuels and two structural materials. The tests were performed at the NASA Glenn Research Center Heated Tube Facility under environmental conditions similar to those encountered in regeneratively cooled rocket engines. Scanning-electron microscopic analysis in conjunction with energy dispersive spectroscopy (EDS) was utilized to characterize the condition of the tube inner wall surface and any carbon deposition or corrosion that was formed during selected runs. Results show that the carbon deposition process in stainless steel tubes was relatively insensitive to fuel type or test condition. The deposition rates were comparable for all fuels and none of the stainless steel test pieces showed any signs of corrosion. For tests conducted with copper tubing, the sulfur content of the fuel had a significant impact on both the condition of the tube wall and carbon deposition rates. Carbon deposition rates for the lowest sulfur fuels (2 ppm) were slightly higher than those recorded in the stainless steel tubes with no corrosion observed on the inner wall surface. For slightly higher sulfur content (25 ppm) fuels, nodules that intruded into the flow area were observed to form on the inner wall surface. These nodules induced moderate tube pressure drop increases. The highest sulfur content fuels (400 ppm) produced extensive wall pitting and dendritic copper sulfide growth that was continuous along the entire tube wall surface. The result of this tube degradation was the inability to maintain flow rate due to rapidly increasing test section pressure drops. Accompanying this corrosion were carbon deposition rates an order of magnitude greater than those observed in comparable stainless steel tests. The results of this investigation indicate that trace impurities in fuels (i.e. sulfur) can significantly impact the carbon deposition process and produce unacceptable corrosion levels in copper based structural materials.

Meyer, M. L.↗
NASA NTRSDate not supplied

Advanced Turboprop Transport Aircraft Modeling for the Electrified Powertrain Flight Demonstration Project

The objective of this study is to establish models for state of the art, representative regional non-electrified turboprop configurations in the 19-50 passenger capability range and assess performance impacts from advanced aircraft technologies. Two of these models will serve as reference baseline configurations for comparison to electrified aircraft propulsion (EAP) demonstrator concepts for the Electrified Powertrain Flight Demonstration (EPFD) project. The other two non-electrified configurations will primarily assess the impacts of advanced technologies on fuel consumption and CO₂ emissions. This paper details an aircraft synthesis and performance analysis approach using the General Aviation Synthesis Program (GASP) which is used to establish baseline performance capabilities, calibrate airplane and engine models using published manufacturer data and conduct technology sensitivity studies. Advanced technologies are selected from the categories of propulsion, aerodynamics, and flight systems. Performance assessments use Monte Carlo method coupled with GASP simulations to obtain performance distributions of the expected fuel burn, propulsive, aerodynamic, and weight benefits on the representative turboprop configurations from the application of advanced technologies. The final paper will present the impact of the advanced technologies on fuel burn and CO₂ emissions.

aircraft performance↗
DOE OSTI2026

Triso Analysis Tool For Predictive Source Terms

Source term modeling for TRi-structural ISOtropic (TRISO) fuel has been performed for previous reactor designs, but few are available in the open literature. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that can be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. To meet this need, the TRISO Analysis Tool for Predictive Source terms (TRISO-ATOPS) was developed. This model calculates the release of the key safety-important fission products by diffusion through the kernel, silicon carbide and graphite based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system for removing fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. From this initial condition, the model calculates the fission product release for any transient temperature profile, and the fission product releases can then be used to assess radiological dose to the workers and the public using conventional radiological dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy Advanced Gas Reactor TRISO fuel development program. The example cases in this work demonstrate the flexibility of the model

Stoyer, Benjamin [Idaho National Laboratory (INL),↗
DOE OSTI2016

Modelling of LOCA Tests with the BISON Fuel Performance Code

BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

BISON code↗
DOE OSTI2019

Effects of system design and Co-product treatment strategies on the life cycle performance of biofuels from microalgae

This study presents a life cycle greenhouse gas and energy assessment for two algal biofuel production pathways: biodiesel produced through lipid extraction (LE) and renewable diesel produced through hydrothermal liquefaction (HTL). The two production pathways generate different co-products, which are handled through allocation in life cycle assessment-based analyses. The method and assumptions used for co-product allocation affect the performance of the analyzed fuels, and are thus examined through scenario analysis; five co-product allocation strategies are tested for the LE pathway and six are tested for the HTL pathway. After allocation, the carbon intensity of renewable diesel varies from 36 to 54 gCO 2 e/MJ, and the primary energy consumption of renewable diesel varies from 0.7 to 1.2 MJ/MJ; while the carbon intensity of biodiesel ranges, remarkably, from -59 to 125 gCO 2 e/MJ, and the primary energy consumption of biodiesel ranges from 0.1 to 1.7 MJ/MJ. The optimal algal oil production pathway is determined by comparing open-loop and closed-loop systems, considering not only the estimated net environmental impacts, but also the confidence or uncertainty of those outcomes.

09 BIOMASS FUELS↗
DOE OSTI2024

The Use of High-Density UN Fuel in Heat-Pipe Microreactors

Heat-pipe microreactors (HPMRs) are very small-scale nuclear reactors that employ heat pipes (HPs) for heat removal. HPMRs can be easily integrated with other forms of renewable energies, can be used for emergency responses to disaster relief zones, can be deployed in remote locations not connected to the grid, and can be removed from sites and replaced by new ones. HPMRs can also be used for space missions as HPs do not rely on gravity for heat transfer. Conventional fuel materials, such as uranium oxide (UO 2 ) and uranium oxycarbide (UCO), are currently considered in most existing HPMR designs, but ceramic uranium nitride (UN) fuel that has high uranium density, high thermal conductivity, and high melting point may become a better fuel candidate. Through neutronics calculations, this paper assesses the impact of using UN fuel in HPMRs with two different neutron spectra (fast and thermal) and two different fuel forms [traditional solid fuel pellets and TRi-structural-ISOtropic (TRISO) fuel compacts]. It was concluded that retrofitting HPMRs with UN fuel has the potential to reduce the initial 235 U enrichment requirement by ~3 wt% (to keep the same cycle length) or increase the cycle length (by keeping the same initial 235 U enrichment), which enables more compact and transportable HPMR core designs. However, using UN fuel decreases the control element worth [by up to 20% for the Special Purpose Reactor (SPR) and 5% for HP-MR] and is up to 80% more costly. Increasing 15 N enrichment can further decrease the initial 235 U enrichment requirement and increase the control element worth but is more costly. In conclusion, compared to fast-spectrum HPMRs fueled with solid pellet fuels, retrofitting UN fuel is more suitable for thermal-spectrum HPMRs fueled with TRISO fuel compacts, where the neutron spectrum hardening caused by using UN is less significant.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Fabrication, Welding, and Inspection Techniques in Preparation for In-Reactor Testing of Accident Tolerant Fuels

After the Fukushima events in 2011, DOE-NE in collaboration with nuclear industry shifted R&D emphasis to accident performance of LWR fuels under extended loss of active cooling and steam exposure. DOE-NE has created a roadmap for the “Development of Light Water Reactor Fuels with Enhanced Accident Tolerance.” The mission of the Accident Tolerant Fuel (ATF) Roadmap is to develop the next generation of LWR fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. The ultimate goal of the ATF roadmap is to support the insertion of lead fuel rods (LFRs) or lead fuel assemblies (LFAs) of an Accident Tolerant Fuel into a commercial LWR within 10 years (i.e., by the end of FY-2022). As a step toward this goal, an irradiation test series has been developed to assess the performance of proposed ATF concepts under normal LWR operating conditions. Data generated by this test program will be used to establish the feasibility of certain aspects of proposed ATF concepts, as well as provide information to support screening among concepts; as such, it is an integral part of Phase I: Feasibility Assessment and Down-Selection outlined in the ATF Roadmap. This irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL), and it has been designated as the ATF-1 test series. Current fission reactors use zirconium-based fuel cladding because of its extremely low macroscopic thermal neutron absorption cross-section, good high temperature strength, and decent corrosion resistance. However, advanced, innovative materials may provide these same benefits while increasing reactor safety margin, core power density, and fuel utilization. These advanced fuel cladding systems will allow revolutionary cladding performance and enhanced fuel mechanical designs, however, challenges exist in design, analysis and fabrication of innovative, never before tested, fuel cladding systems for in-reactor testing. This paper highlights the challenges associated with design, fabrication and welding, and inspection of innovative materials and actions taken to address those challenges in preparation for the Phase I ATR irradiation testing. The lessons learned from Phase I of this experiment can be used to guide researchers for design and analysis of future in-reactor testing of advanced fuel cladding systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Industry Application ECCS / LOCA Integrated Cladding/Emergency Core Cooling System Performance: Demonstration of LOTUS-Baseline Coupled Analysis of the South Texas Plant Model

Under the auspices of the DOE LWRS Program RISMC Industry Application ECCS/LOCA, INL has engaged staff from both South Texas Project (STP) and the Texas A&M University (TAMU) to produce a generic pressurized water reactor (PWR) model including reactor core, clad/fuel design and systems thermal hydraulics based on the South Texas Project (STP) nuclear power plant, a 4-Loop Westinghouse PWR. A RISMC toolkit, named LOCA Toolkit for the U.S. (LOTUS), has been developed for use in this generic PWR plant model to assess safety margins for the proposed NRC 10 CFR 50.46c rule, Emergency Core Cooling System (ECCS) performance during LOCA. This demonstration includes coupled analysis of core design, fuel design, thermalhydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results. Within this context, a multi-physics best estimate plus uncertainty (MPBEPU) methodology framework is proposed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Volume 1: Introduction to the HFIR Futures - Enhanced Capabilities Series

Since it began operating at full power in 1966, the High Flux Isotope Reactor (HFIR) has contributed unparalleled neutron science capabilities to research on neutron scattering, isotope production, materials and fuels irradiation, and neutron activation analysis. HFIR is a high-performance, multi-mission research reactor operated on behalf of the US Department of Energy (DOE) at the Oak Ridge National Laboratory (ORNL). In 2020, a DOE Basic Energy Sciences Advisory Committee Subcommittee (BESAC) published a report recommending that DOE make significant investments in HFIR to enable continued operations beyond the year 2100 while also enabling new research and isotope production capabilities. ORNL organized an initiative to investigate how to specifically address the report’s key recommendations, including enabling long-term operation, brainstorming future scientific research needs, and outlining the infrastructure required to realize these future research capabilities. A multivolume series of reports has been developed to document the nonscattering enhancements. This volume, the introductory report, provides an overview of HFIR and the initiative.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1989

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 6: Primary nozzle diffuser analysis

The primary nozzle diffuser routes fuel from the main fuel valve on the Space Shuttle Main Engine (SSME) to the nozzle coolant inlet mainfold, main combustion chamber coolant inlet mainfold, chamber coolant valve, and the augmented spark igniters. The diffuser also includes the fuel system purge check valve connection. A static stress analysis was performed on the diffuser because no detailed analysis was done on this part in the past. Structural concerns were in the area of the welds because approximately 10 percent are in areas inaccessible by X-ray testing devices. Flow dynamics and thermodynamics were not included in the analysis load case. Constant internal pressure at maximum SSME power was used instead. A three-dimensional, finite element method was generated using ANSYS version 4.3A on the Lockheed VAX 11/785 computer to perform the stress computations. IDEAS Supertab on a Sun 3/60 computer was used to create the finite element model. Rocketdyne drawing number RS009156 was used for the model interpretation. The flight diffuser is denoted as -101. A description of the model, boundary conditions/load case, material properties, structural analysis/results, and a summary are included for documentation.

Foley, Michael J.↗
NASA NTRS1989

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 7: High pressure fuel turbo-pump third stage impeller analysis

This volume summarizes the analysis used to assess the structural life of the Space Shuttle Main Engine (SSME) High Pressure Fuel Turbo-Pump (HPFTP) Third Stage Impeller. This analysis was performed in three phases, all using the DIAL finite element code. The first phase was a static stress analysis to determine the mean (non-varying) stress and static margin of safety for the part. The loads involved were steady state pressure and centrifugal force due to spinning. The second phase of the analysis was a modal survey to determine the vibrational modes and natural frequencies of the impeller. The third phase was a dynamic response analysis to determine the alternating component of the stress due to time varying pressure impulses at the outlet (diffuser) side of the impeller. The results of the three phases of the analysis show that the Third Stage Impeller operates very near the upper limits of its capability at full power level (FPL) loading. The static loading alone creates stresses in some areas of the shroud which exceed the yield point of the material. Additional cyclic loading due to the dynamic force could lead to a significant reduction in the life of this part. The cyclic stresses determined in the dynamic response phase of this study are based on an assumption regarding the magnitude of the forcing function.

Pool, Kirby V.↗
DOE OSTI2016

Neutronics and computational fluid dynamics analyses for a natural circulation loop of molten fuel salt

A conceptual reactor system of molten fuel salt loop designed for Mo-99 production was analyzed by coupling TRITON neutronics and Fluent computational fluid dynamics (CFD) simulations. The loop system operated in 1 MW thermal power features natural circulation and potential extraction of non-soluble fission products (FP) in fuel salt. The behavior of natural circulation was studied and the power and temperature distributions in equilibrium were obtained. A long-term fuel depletion analysis of the system considering fuel circulation and reprocessing was then performed using an approximate stepwise sequence of simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1998

Results of a 24-inch Hybrid Motor Performance Uncertainty Analysis

The subscale (11 and 24-inch) hybrid motors at the Marshall Space Flight Center (MSFC) have been used as versatile and cost effective testbeds for developing new technology. Comparisons between motor configuration, ignition systems, feed systems, fuel formulations, and nozzle materials have been carried out without detailed consideration as to haw "good" the motor performance data were. For the 250,000 lb/thrust motor developed by the Hybrid Propulsion Demonstration Program consortium, this shortcoming is particularly risky because motor performance will likely be used as put of a set of downselect criteria to choose between competing ignition and feed systems under development. This analysis directly addresses that shortcoming by applying uncertainty analysis techniques to the experimental determination of the characteristic velocity, theoretical characteristic velocity, and characteristic velocity efficiency for a 24-inch motor firing. With the adoption of fuel-lined headends, flow restriction, and aft mixing chambers, state of the an 24-inch hybrid motors have become very efficient However, impossibly high combustion efficiencies (some computed as high as 108%) have been measured in some tests with 11-inch motors. This analysis has given new insight into explaining how these efficiencies were measured to be so high, and into which experimental measurements contribute the most to the overall uncertainty.

Sims, Joseph D.↗
DOE OSTI2024

A Physical Metric for Inertial Confinement Fusion Capsules

The performance of fusion capsules on the National Ignition Facility (NIF) is strongly affected by the physical properties of the hot deuterium–tritium (DT) fuel, such as the mass, areal density, and pressure of the hot spot at the stagnation time. All of these critical quantities depend on one measured quantity, which is the ratio of the specific peak implosion energy to the specific internal energy of the hot spot. This unique physical quantity not only can measure the incremental progress of the inertial confinement fusion capsules towards ignition but also measures the conversion of the peak implosion kinetic energy of the pusher shell into the internal energy of the hot fuel in a capsule. Analysis of existing NIF shots to date are performed. The ratio metric is compared quantitatively with the ignition criterion. Results provide new perspectives on the NIF experiments by which the performance of the burning plasma can be determined and controlled through the fine tune of the implosion parameters, which improves future designs and predictions of the ignition capsules.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRSDate not supplied

Advanced Turboprop Transport Aircraft Modeling for the Electrified Powertrain Flight Demonstration Project

The objective of this study is to establish models for state of the art, representative regional non-electrified turboprop configurations in the 19 to 50 passenger capability range and assess performance impacts from advanced aircraft technologies. Two of these models will serve as reference baseline configurations for comparison to electrified aircraft propulsion (EAP) demonstrator concepts for the Electrified Powertrain Flight Demonstration (EPFD) project. The other two non-electrified configurations will primarily assess the impacts of advanced technologies on fuel consumption and CO₂ emissions. This paper details an aircraft synthesis and performance analysis approach using the General Aviation Synthesis Program (GASP) which is used to establish baseline performance capabilities, calibrate airplane and engine models using published manufacturer data and conduct technology sensitivity studies. Advanced technologies are selected from the categories of propulsion, aerodynamics, and flight systems. A GASP-Monte Carlo simulation framework for uncertainty propagation is used to obtain performance distributions of the expected fuel burn, weight reduction and CO₂ emissions reductions from advanced aircraft technologies. Significant fuel burn savings from application of aerodynamics and propulsion technologies was observed.

Dahlia D.V. Pham↗