Search NASA⌕ Search

SEARCH · Search NASA

Results for “Fuel Performance”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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

At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

AGR-5/6/7 Irradiation Experiment Fission Product Mass Balance

This report presents the fission product mass balance for the AGR-5/6/7 TRISO fuel irradiation experiment. The fission product inventories deposited on capsule components outside of the fuel (e.g., stainless-steel shells, graphite holders, Grafoil disks, and associated hardware) were quantified as part of the post-irradiation examination (PIE) to assess the performance of this fuel. Comparisons were made between these inventories and depletion calculations, non-destructive measurements of fuel fission product inventories in the intact fuel compacts, prior AGR experiments, and results from among each of the five distinct AGR-5/6/7 capsules. The data served as estimates of the condensable fission products released from the fuel during irradiation. Excluding Capsule 1 (which experienced accidental damage during irradiation) and Capsule 3 (which was tested at very high irradiation temperatures of >1300°C), the results indicate that the AGR-5/6/7 fuel performed comparably to fuel from earlier AGR experiments. The mass balance results were also used to estimate the number of particles with in-pile SiC failures. Subject to the assumptions made in these estimates, and excluding Capsules 1 and 3, the in-pile SiC failure rates for AGR-5/6/7 are comparable to those observed in AGR-2.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Irradiation of UC1+x kernels using the MiniFuel vehicle: Microstructure, phase analysis, and initial post-irradiation examination

Uranium carbide is a candidate fuel form for a wide range of advanced reactors, including larger Generation IV reactors as well as small modular reactors and microreactors. However, its commercial deployment timeline faces challenges via traditional qualification approaches. To address this issue, an accelerated fission rate irradiation test was performed to collect basic fuel performance data to inform fuel performance models and potential future integral tests. Hyperstoichiometric UC (UC1+x) kernels were irradiated in the High Flux Isotope Reactor using the MiniFuel irradiation vehicle. The test matrix spanned two temperature regimes (700 °C and 800 °C) and burnup levels (1.8 % FIMA and 2.8 % FIMA). Between 21 and 63 kernels were tested at each unique temperature and burnup condition. As-fabricated microstructural analysis revealed a multiphase composition with UC, UC2, UC2−y, and U-C-O bearing phases for the irradiated kernels. Following irradiation, fission gas release, swelling, and microstructures were analyzed. Measured fission gas release was below 5 % for all irradiation conditions, reaching a maximum at the highest temperature and burnup condition. A binary swelling response was observed; the lower burnup and temperature conditions resulted in negligible swelling, but the higher burnup and temperature conditions produced significant anisotropic swelling and densification in a subset of kernels. The basic microstructural exams of kernels following irradiation were not capable of showing a correlation between kernels that exhibited excessive swelling and those that did not. Characterization of a subset of samples using the Advanced Photon Source and more detailed microstructural examination of unirradiated kernels revealed that a subset of kernels contained very high UC2 phase fractions. The anomalous swelling response is hypothesized to have been driven by this chemical variation. The results of this irradiation highlight the potential of accelerated fission rate irradiation testing to explore such behaviors and inform the development of fuel specifications.

Adorno Lopes, Denise [ORNL] (ORCID:000900023705987↗
DOE OSTI2025

Rheology and engine performance of very low sulfur fuel oil blended with 10% fast pyrolysis and hydrothermal liquefaction oils in a 2-stroke crosshead engine

The performance and emissions for a downscaled single-cylinder 2-stroke crosshead engine were determined for a very low sulfur fuel oil (VLSFO) when blended with 10 wt.% fast pyrolysis (FP) or hydrothermal liquefaction (HTL) bio-intermediates. The FP and HTL oils were derived from biomass and were observed to contain lower molecular weight (MW) hydrocarbons than neat VLSFO (which was evaluated as a baseline comparison). The addition of either biofuel reduced the overall viscosity of the VLSFO. Aging tests at 50, 90, and 120°C showed that the dynamic viscosity of VLSFO increased with exposure time up to two weeks. Similar trends were observed for the FP and HTL blends, but a pronounced spike in viscosity occurred for these fuels during the early period of exposure. None of the viscosity increases exceeded the operational limits of fuel system pumps. Engine performance studies were conducted under low, medium and high load operational settings. The relative performance of the test fuels was highly dependent on operating condition. In general, the engine results for the three test fuels were similar, but modest improvements in brake thermal efficiency and brake specific fuel consumption were observed, which may be attributed to the heightened reactivity of low molecular weight fraction of the FP and HTL oils.

09 BIOMASS FUELS↗
DOE OSTI2024

Understanding Direct-Ammonia Protonic Ceramic Fuel Cells: High-Performance in the Absence of Precious Metal Catalysts

Ammonia has received considerable attention as a promising carbon-free hydrogen carrier. At temperatures above 400 °C, NH 3 is thermodynamically unstable with respect to decomposition into nitrogen and hydrogen and is, thus, suitable for direct use in solid oxide fuel cells (SOFCs) without external reforming. However, poor catalytic activity for ammonia decomposition at the moderate temperatures of protonic ceramic fuel cell (PCFC) operation has resulted in low fuel cell power output relative to operation on hydrogen and likely contributes to reported cell degradation. Here we prepared cells based on a thermodynamically robust electrolyte, a high activity cathode, and an anode with a distinctive structure to overcome challenges of poor activity and stability. Furthermore, the cells delivered peak power densities of 0.59 and 0.44 W cm –2 under H 2 and NH 3 , respectively, at 500 °C, excellent stability over a period of 200 h, and no detectable NO x in the anode exhaust gas.

25 ENERGY STORAGE↗
DOE OSTI2024

Report on DIC Cyclic Dryout Testing of ATF Cladding Materials

Boiling water reactors can experience anticipated operational occurrences, where an expected deviation from normal operating conditions is witnessed for short time scales. One example of such transients arises from cladding dryout, where stable boiling becomes unstable, resulting in cladding heat up before re-wetting occurs. To understand the impact of such transients on fuel cladding lifetime and performance, cyclic dryout conditions were imposed upon fresh fuel cladding in the absence of irradiation. Pressurized cladding tube segments were exposed to rapid temperature increases to intermediate temperatures below a cladding rupture threshold and deformation was monitored in situ via 3-D digital image correlation. The impact of cycle frequency on cladding deformation was assessed. A clear understanding of elastic and plastic strain responses were developed for Zr alloy deformation under temperature transients, and these results can be used to support material model development and validate fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A Multiphysics Evaluation of Annular Uranium-Zirconium Metallic Fuels [Poster]

This study examines the performance of U-10Zr annular metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included annular metallic U-Zr fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Loss of Flow Conditions in a Modern Pool-type SFR and In-Pile Experiment Design

The metallic fuel safety performance under unprotected design-basis transients is a key consideration for the deployment of advanced sodium fast reactors (SFRs). Reliable data are needed to validate advanced safety codes, reduce uncertainty in cladding failure thresholds, and strengthen confidence in licensing approaches. To address this need, this report develops blueprints for a conceptual sodium loss-of-flow (LOF) experiment in the Mk-IIIR loop at the Transient Reactor Test Facility (TREAT), providing the technical foundation for future integral testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

3D TRISO particle-explicit compact meshing

The TRI-structural ISOtropic (TRISO) layered fuel particle is a robust nuclear fuel form offering enhanced safety and performance for advanced reactor concepts, including high-temperature gas-cooled reactors and other Generation IV designs. These poppy-seed-sized particles are embedded in a graphite matrix to form fuel elements that must withstand elevated temperatures and high burn-up levels. The heterogeneous nature of these fuel elements — comprising thousands of randomly distributed TRISO particles — produces complex stress fields and thermal gradients that one- and two-dimensional models cannot accurately capture. While three-dimensional modeling has improved predictions of dimensional changes, internal pressure buildup, and fission product transport under irradiation, current approaches rely on homogenized material properties that are known to have considerable divergence from experimental observations. This work presents a methodology for optimized random packing of TRISO fuel compacts and full three-dimensional mesh generation within the BISON fuel performance code, with each particle coating layer individually discretized. The resulting mesh was demonstrated through heat conduction simulations under representative in-reactor operating conditions, showing strong agreement with expected behavior. This capability enables detailed analysis of particle-to-particle interactions, matrix cracking mechanisms, and the statistical distribution of coating layer failures — all of which directly govern fuel performance and safety margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

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

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

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Influence of Pt-Metal Alloy Catalysts with Various Ionomers on Oxygen Reduction Reaction in Fuel Cell Application

Pt-M/C (M = Co, Ni, Mn, etc.) alloy catalysts exhibit superior oxygen reduction reaction (ORR) activity compared to pure Pt/C, leading to a high energy efficiency in hydrogen fuel cells. However, many Pt-M/C alloy catalysts were synthesized and evaluated at the lab scale in model test-bed systems like rotating disc electrodes, which don't always correlate to performance within a fuel cell system; there is a clear need to evaluate catalysts in electrodes that can be prepared at industrially relevant scales to evaluate how factors like ink formulation can greatly affect device-level of fuel cell performance. Herein, three commercial Pt-M/C alloy catalysts (two Pt-Co/C and one Pt-Ni/C) were comprehensively characterized by various techniques. The results show that the average particle sizes of the three catalysts are close to 5 nm; the atomic ratio of Pt/M is around 4; and the M was successfully embedded into Pt lattice, resulting in the positive shift of Pt 4f in XPS spectra and XRD patterns. These catalytic materials were incorporated into 9 different cathode catalyst layers (CCLs) with three kinds of ionomers (Nafion D2020, high oxygen permeability ionomer (HOPI), and Aquivion D79-25BS), and their performance in proton exchange membrane fuel cells (PEMFCs) were investigated. The results demonstrate that the Pt-Co/C catalysts possess a higher mass activity (MA) than Pt-Ni/C; the cathodes with Nafion ionomer provide the highest MA while electrodes with Aquivion ionomer showed the lowest activity, attributed to poor H+ conductivity resulting from suboptimal ionomer incorporation. Finally, these alloys were shown to exceed DOE targets for MA and H2/Air performance reported in the recent publications at beginning of life and after 90k cycle catalyst AST protocol. This study provides valuable performance benchmarks for these materials guiding future Pt-M/C catalyst design and material integration for heavy duty PEMFC applications.

08 HYDROGEN↗
DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Heterostructured nano-catalysts with efficient metal-oxide interfaces unlock high-performance direct methanol protonic ceramic fuel cells

Direct methanol protonic ceramic fuel cells (PCFCs) are attractive due to their low cost, convenient storage, and high volumetric energy density, as well as their suitability for transportation. However, the poor coking tolerance of conventional nickel-based anodes leads to their susceptibility to severe carbon deposition and significant deactivation after long-term exposure to hydrocarbons. Herein, we report a nano-catalyst of Ce 0.6 Ni 0.2 Cu 0.2 O 2 with a heterogeneous structure that is spontaneously reduced into a Ce 0.6 Ni 0.2-x Cu 0.2-x O 2-δ (CeNCO) oxide framework interfaced with a nano NiCu alloy (denoted as NC/CeNCO) under operating conditions, as confirmed by analyses of X-ray diffraction, X-ray photoelectron spectroscopy, scanning electron microscopy, and transmission electron microscopy. A Ni-BaCe 0.7 Y 0.06 Yb 0.06 Zr 0.06 Hf 0.06 Gd 0.06 O 3-δ anode-supported PCFC employing the NC/CeNCO metal-oxide catalyst achieved a peak power density of 1.11 W cm −2 and operational stability of about 100 h at 700 °C when fueled by 35 % CH 3 OH-15 % H 2 O-50 % N 2 . In conclusion, the enhanced performance and coking resistance are attributed to the efficient interfaces of Ni, Cu, and ceria-based oxide in NC/CeNCO for CH 3 OH reforming, as confirmed by analyses of electrochemical performance and Raman spectroscopy with density functional theory calculations, revealing that these interfaces can enhance CH 3 OH activation and promote efficient OH-mediated carbon removal via COH intermediates.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2026

Mechanical and durability properties of ultra-high-performance concrete of spent nuclear fuel dry storage systems: a review

Dry storage systems are used for interim storage of spent nuclear fuel (SNF). However, with the growing need to extend the operational periods of these systems, there are concerns about the degradation of their concrete overpacks, which could compromise the system's structural integrity and safety during hazardous events. Traditional concrete mixtures used in SNF dry storage systems have remained largely unchanged since their inception and often use conventional ingredients. These materials are susceptible to degradation mechanisms such as chemical attacks, alkali-silica reactions (ASR), and freeze–thaw cycles, which can lead to a loss of strength and durability over time. To address these challenges, this paper reviews the application of ultra-high-performance concrete (UHPC) as a promising alternative for spent nuclear fuel dry storage system overpacks. UHPC offers superior mechanical properties, exceptional durability, and reduced susceptibility to degradation mechanisms compared to conventional concrete. This paper focuses on the role of supplementary cementitious materials (SCMs) such as silica fume, fly ash, and metakaolin in enhancing UHPC performance for SNF storage applications. These SCMs have been shown to significantly improve the material’s microstructure, strength, and resistance to environmental stressors typically encountered in SNF storage environments. Moreover, incorporating SCMs supports sustainable construction by reducing cement consumption and associated carbon emissions. The review brings together existing research and experimental data, providing insights for engineers and researchers on developing UHPC mixtures that meet the rigorous demands of spent nuclear fuel dry storage systems, extending their service life and minimizing inspection intervals.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Developing an Energy-Conscious Traffic Signal Control System for Optimized Fuel Consumption in Connected Vehicle Environments

The project titled “Developing an Energy-Conscious Traffic Signal Control System for Optimized Fuel Consumption in Connected Vehicle Environments” addresses energy-related challenges associated with adaptive traffic control systems by integrating connected vehicles (CV) and connected infrastructure (CI). The system developed in this project, a CV-based adaptive traffic control system, aims to improve fuel consumption in mixed traffic environments by capitalizing on emerging CV and CI communication technologies, as well as leveraging recent advances in Artificial Intelligence (AI), optimization, and edge computing. The system was tested at the MLK Smart Corridor, an urban testbed managed by the University of Tennessee at Chattanooga (UTC) and the City of Chattanooga. The system was validated through extensive simulations, both Software-in-the-Loop (SILS) and Hardware-in-the-Loop (HILS), and was further implemented and tested in real-world conditions at several intersections along the corridor. The Fuel Consumption Performance Index (FC-PI) and the Ecological Performance Index (Eco-PI) were developed as the key components for evaluating the system’s impact on fuel consumption and emissions. These metrics provided a comprehensive means of understanding the impact of traffic signal control optimization in mixed traffic environments. The report presents an in-depth analysis of the Eco-PI, FC-PI, adaptive traffic control system integration, and the testing and field implementation of the system. The results demonstrate significant reductions in fuel consumption and emissions, showcasing the system’s capability to contribute to more sustainable urban traffic management. The report also documents the challenges encountered and recommendations for scaling and further improving the system.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗