Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Fuel testing”

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 1,477 records · Page 82
All sourcesNASA NTRSDOE OSTI
NASA NTRS2017

High-Resolution OH and CH2O Visualization in a Premixed Cavity-Anchored Ethylene-Air Flame in a M = 0.6 Flowfield

OH and CH2O were imaged in a premixed, cavity-anchored, ethylene-air turbulent flame using a high resolution planar laser-induced fluorescence (PLIF) system. The electrically-heated, continuous flow facility (UVa Supersonic Combustion Facility, Configuration E) consisted of a Mach 2 nozzle, an isolator with fuel injectors, a test section with a cavity flame holder and optical access, and an extender. Standard test conditions comprised total temperature 1200 K, total pressure 300 kPa, local equivalence ratio near 0.4, and local Mach number near 0.6. OH PLIF data was also collected for a case with reduced total temperature and another with reduced equivalence ratio. OH and CH2O were excited in separate experiments with light sheets at 283.55 nm and 352.48 nm, respectively. A light sheet of approximate thickness 25 ?m illuminated the stream-wise midplane. This plane was imaged for 120 mm downstream of the backward-facing step. The intensified camera system imaged OH with magnification 1.97, a square 6.67 mm field of view, and in-plane resolution of 39 ?m. The smallest observed OH structures observed were approximately 100 ?m wide. The CH2O PLIF image signal was much weaker; the smallest observed structures were approximately 200 ?m wide. Composite fluorescence images were computed for the observed area.

Geipel, Clayton M.↗
NASA NTRS2017

A Comparison of Combustion Dynamics for Multiple 7-Point Lean Direct Injection Combustor Configurations

The combustion dynamics of two 7-point lean direct injection (LDI) combustor configurations are compared. This 7-point LDI configuration has a circular cross section, with a center ("pilot") fuel-air mixer surrounded by six outer ("main") fuel-air mixers. Each fuel-air mixer consists of an axial air swirler followed by a converging-diverging venturi. A simplex fuel injector is inserted through the center of the air swirler, with the fuel injector tip located near the venturi throat. All 7 fuel-air mixers are identical except for the swirler blade angle. In the 'all-60' configuration, the swirler blade angle was 60 deg for all fuel-air mixers. In the '45-60' configuration, the swirler blade angle was 60 deg on the center and 45 deg on the outer fuel-air mixers. Testing was done in a 5-atm flame tube with inlet air temperatures from 630 to 830 F and equivalence ratios from 0.2 to 0.7. Combustion dynamics were measured using a cooled PCB pressure transducer flush-mounted in the wall of the combustor test section. Both configurations had large pressure fluctuations (greater than 2 psi peak-peak) near 730 Hz, the quarter-wave frequency. The all-60 configuration also had large pressure fluctuations near 1170 Hz; the 45-60 configuration did not. The 45-60 configuration had large pressure fluctuations near 480 Hz; the all-60 configuration did not.

combustion dynamics↗
DOE OSTI2021

NSUF BRR Cask Shipment Summary Report

On September 16, 2021 the Idaho National Laboratory shipped a Battelle Research Reactor (BRR) cask loaded with five Nuclear Science User Facilities (NSUF) experiments and one Accident Tolerant Fuels (ATF) experiment from the Advanced Test Reactor (ATR) to the Hot Fuel Examination Facility (HFEF) at the INL Materials and Fuels Complex (MFC) for post-irradiation examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Irradiations for Advanced Reactors

The presentation is for a public seminar explaining test reactor capabilities for testing advanced reactor fuels and materials

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Vacuum Assisted Filtered Salt Sampling Progress

Several different design iterations for a filtered salt sampling technique have been tested in non-radiological molten salts. A piston vacuum assembly was designed and reliably and repeatably used to take salt samples through multiple different porous quartz frit sizes (as small as 5 – 10 µm). This design has been modified slightly and sent into the HFEF hot cell for upcoming testing with used fuel electrorefiner salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1987

Antimisting kerosene JT3 engine fuel system integration study

An analytical study and laboratory tests were conducted to assist NASA in determining the safety and mission suitability of the modified fuel system and flight tests for the Full-Scale Transport Controlled Impact Demonstration (CID) program. This twelve-month study reviewed and analyzed both the use of antimisting kerosene (AMK) fuel and the incorporation of a fuel degrader on the operational and performance characteristics of the engines tested. Potential deficiencies and/or failures were identified and approaches to accommodate these deficiencies were recommended to NASA Ames -Dryden Flight Research Facility. The result of flow characterization tests on degraded AMK fuel samples indicated levels of degradation satisfactory for the planned missions of the B-720 aircraft. The operability and performance with the AMK in a ground test engine and in the aircraft engines during the test flights were comparable to those with unmodified Jet A. For the final CID test, the JT-3C-7 engines performed satisfactorily while operating on AMK right up to impact.

Fiorentino, A.↗
DOE OSTI2016

Test Plan for Cask Identification Detector

This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1977

Reduced truck fuel consumption through aerodynamic design

Full-scale fuel consumption and drag tests were performed on a conventional cab-over-engine tractor-trailer combination and a version of the same vehicle with significant forebody modifications. The modified configuration had greatly increased radii on all front corners and edges of the tractor and a smooth fairing of the modified tractor top and sides extending to the trailer. Concurrent highway testing of the two configurations showed that the modified design used 20% to 24% less fuel than the baseline configuration at 88.5 km/hr (55 mph) with near-calm wind conditions. Coastdown test results showed that the modified configuration reduced the drag coefficient by 0.43 from the baseline value of 1.17 at 88.5 km/hr (55 mph) in calm wind conditions.

Steers, L. L.↗
DOE OSTI2016

Use of Adaptive Injection Strategies to Increase the Full Load Limit of RCCI Operation

Dual-fuel combustion using port-injection of low reactivity fuel combined with direct injection (DI) of a higher reactivity fuel, otherwise known as reactivity controlled compression ignition (RCCI), has been shown as a method to achieve low-temperature combustion with moderate peak pressure rise rates, low engine-out soot and NO x emissions, and high indicated thermal efficiency. A key requirement for extending to high-load operation is moderating the reactivity of the premixed charge prior to the diesel injection. One way to accomplish this is to use a very low reactivity fuel such as natural gas. In this work, experimental testing was conducted on a 13 l multicylinder heavy-duty diesel engine modified to operate using RCCI combustion with port injection of natural gas and DI of diesel fuel. Engine testing was conducted at an engine speed of 1200 rpm over a wide variety of loads and injection conditions. The impact on dual-fuel engine performance and emissions with respect to varying the fuel injection parameters is quantified within this study. The injection strategies used in the work were found to affect the combustion process in similar ways to both conventional diesel combustion (CDC) and RCCI combustion for phasing control and emissions performance. As the load is increased, the port fuel injection (PFI) quantity was reduced to keep peak cylinder pressure (PCP) and maximum pressure rise rate (MPRR) under the imposed limits. Overall, the peak load using the new injection strategy was shown to reach 22 bar brake mean effective pressure (BMEP) with a peak brake thermal efficiency (BTE) of 47.6%.

Hanson, Reed↗
DOE OSTI2022

Materials and Fuels Complex FY-22 – FY-26 Five-Year Mission Strategy

The Idaho National Laboratory (INL), through its designated mission of advancing innovative nuclear energy solutions, is actively engaged in the research, development, demonstration and deployment of advanced nuclear technology. Key to the success of INL’s mission is the Materials and Fuels Complex (MFC), the only complex in the U.S. that hosts a world-class assemblage of facilities, capabilities and instruments for handling, testing, and characterizing nuclear fuel and radioactive materials. Driven by its mission/vision of “Engineering and Experiments that Drive the World’s Nuclear Energy Future,” MFC is at the center of INL’s – and indeed the Department of Energy’s – advanced nuclear technology development initiatives, providing essential capabilities such as research-scale high-assay low-enriched uranium (HALEU) fuel production, reactor demonstration facilities, post-irradiation examination, and transient irradiation testing. Furthermore, MFC provides an ideal environment for test beds that are utilized for research, development and demonstration (RD&D) activities on used fuel treatment, nuclear non-proliferation, forensics, and nuclear power sources used for space exploration missions conducted by the National Aeronautics and Space Administration (NASA).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Initial examination of fuel compacts and triso particles from the US AGR-2 irradiation test - 18443

Post-irradiation examination was completed on two as-irradiated compacts from the US Advanced Gas Reactor Fuel Development and Qualification Program's second irradiation test. These compacts were selected for examination due to indications that they may have contained particles that released cesium due to a failed or defective SiC layer. The coated particles were recovered from these compacts by electrolytic deconsolidation of the surrounding graphitic matrix in nitric acid. The leach-burn-leach (LBL) process was used to dissolve and analyze exposed metallic elements (actinides and fission products), and each particle was individually surveyed for relative cesium retention with the Irradiated Microsphere Gamma Analyzer (IMGA). Data from IMGA and LBL examinations provided information on fission product release during irradiation and whether any specific particles had below-average retention that could be related to coating layer defects or radiation-induced degradation. A few selected normal-retention particles and six with abnormally-low cesium inventory were analyzed using x-ray tomography to produce three-dimensional images of the internal coating structure. Four of the low-cesium particles had obviously damaged or degraded SiC, and x-ray imaging was able to guide subsequent grinding and polishing to expose the regions of interest for analysis by optical and electron microscopy. Additional particles from each compact were also sectioned and examined to study the overall radiation-induced microstructural changes in the kernel and coating layers. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Advancing Platooning with ADAS (Advanced Driver-Assistance Systems) Control Integration and Assessment

Application of Cooperative Adaptive Cruise Control (CACC) to heavy duty trucks known as truck platooning has shown fuel economy improvements on the test track under ideal driving conditions. However, limited test data is available to assess the truck platooning under real-world driving conditions. Under this Cummins-led project that was funded by the U.S. Department of Energy, truck platooning with CACC has been tested on a real-world interstate highway and the results of the project are reviewed in this report. At first, the real-world driving conditions were characterized using National Renewable Energy Laboratory (NREL) Fleet DNA database to define test factors, including route, terrain, and highway traffic. Afterward, both test track and on-highway testing guided by SAE J1321 procedures were conducted to assess truck platooning under controlled and real-world driving conditions. On-highway testing was done on a highway route in Indiana, consisting of low, medium, and high road grade segments. The highway test results of 2-truck and 3-truck platooning showed considerably reduced fuel savings compared to the controlled test track data, which mainly stems from the traffic or high-grade portions of the route. However, integration of Cummins powertrain and vehicle eco-driving features such as predictive cruise control and neutral coasting called ADEPT™ on the lead truck showed an improvement of fuel saving for the trucks in CACC operation. Furthermore, the importance of tire connectivity in efficient and safe operation of the trucks in platooning is characterized.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 4: Performance of Aluminum SNF in Dry Storage Experiment Test Plan

This document presents the activities to be performed in support of understanding the performance of aluminum spent nuclear fuel (ASNF) in dry storage. These activities will visually examine and acquire baseline samples of as-irradiated ATR fuel assembly components to determine fuel condition and the morphology of the aluminum oxyhydroxides present on the fuel external surfaces. Because the boehmite phase of oxyhydroxide is projected to be a significant contributor to radiolytic hydrogen production and general corrosion, being able to quantify its presence is of major importance to long term storage of aluminum spent nuclear fuel. Baseline samples will be taken from irradiated end adapters, removed from ATR SNF elements, stored in the ATR Canal. These will be transferred to Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) to perform non-destructive and destructive testing to determine the degree of hydration of the wet-stored aluminum components prior to drying and air storage. Examination of the dry stored SNF at Idaho Nuclear Technology and Engineering Center (INTEC) in CPP-603 Irradiated Fuel Storage Facility (IFSF) will allow the opportunity to identify gross degradation and as conditions permit, sample the surface oxide on fueled and un-fueled portions of the stored assemblies. The information resulting from analysis of the ATR Canal samples will be used to support validation of surrogate samples prepared under Task 6, sample surrogate preparation, and used in Task 1, oxyhydroxide layer behavior and chemistry, Task 2, oxide layer radiolytic gas generation, and Task 5, oxide layer response to drying. Surface oxide samples collected from SNF in IFSF dry storage will SNF dry storage prediction modeling from Task 3, sealed and vented system episodic breathing and gas generation prediction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1981

Catalytic combustion of residual fuels

A noble metal catalytic reactor was tested using two grades of petroleum derived residual fuels at specified inlet air temperatures, pressures, and reference velocities. Combustion efficiencies greater than 99.5 percent were obtained. Steady state operation of the catalytic reactor required inlet air temperatures of at least 800 K. At lower inlet air temperatures, upstream burning in the premixing zone occurred which was probably caused by fuel deposition and accumulation on the premixing zone walls. Increasing the inlet air temperature prevented this occurrence. Both residual fuels contained about 0.5 percent nitrogen by weight. NO sub x emissions ranged from 50 to 110 ppm by volume at 15 percent excess O2. Conversion of fuel-bound nitrogen to NO sub x ranged from 25 to 50 percent.

Bulzan, D. L.↗
DOE OSTI2016

Analysis of the TREAT LEU Conceptual Design

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO 2 -graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Interaction of Polymethyl Methacrylate with Boehmite-Filmed Aluminum Cladding Under Gamma Irradiation

This paper presents an overview of ongoing work to qualify the Advanced Test Reactor (ATR) driver fuel elements that have been affected by irradiation-degraded polymethyl methacrylate (PMMA) flux wands. Irradiation testing was performed on PMMA material in contact with aluminum clad material. The cladding was prefilmed with a boehmite oxide layer, an important feature of the ATR driver fuel. The effects on the boehmite layer due to gamma irradiation of the PMMA-aluminum clad system were investigated. PMMA embrittlement, followed by softening and degradation, occurred at high radiation levels. Adhesion between the cladding and irradiated PMMA was observed. Flow testing at prototypic ATR flow rates demonstrated the effective removal of the adhered material. Measurements of the boehmite layer thickness were performed, and Raman spectroscopy was utilized to detect the presence of boehmite in the irradiated PMMA material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗