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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Sensitive Resources Assessment and Forest Analysis for the Proposed Versatile Test Reactor, Oak Ridge, Tennessee

The US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) is a leading institution in advanced materials, supercomputing, neutrons, and nuclear science. As a research laboratory managed by UT-Battelle, LLC for DOE, ORNL has national priorities in energy, security, and scientific discovery that necessitate facility improvements and expansions. DOE is also committed to environmental stewardship. The laboratory is located on the ~32,000-acre (~13,000-ha) Oak Ridge Reservation (ORR), much of which is categorized as a National Environmental Research Park (NERP) and a state Wildlife Management Area. DOE works with the Tennessee Wildlife Resources Agency (TWRA), Tennessee Department of Environment and Conservation (TDEC), US Fish and Wildlife Service (USFWS), US Department of Agriculture, and other agencies to serve as an effective steward of the ORR. Accordingly, project managers must conform to environmental regulations, agreements, and policies at the federal, state, and institutional levels. Per 40 CFR (Code of Federal Regulations) 1508.14, potential effects on research and science education also represent potential effects of federal actions on the NERP, and impacts on, e.g., deer harvest, must be considered on the Oak Ridge Wildlife Management Area when other aspects of the human environment are affected. The United States currently has no fast neutron testing capability to support advanced nuclear research and development. The proposed Versatile Test Reactor (VTR) will take advantage of current investments by the US government and private industry in nuclear reactors to expedite the design and construction process, using proven technology to create a world-class scientific infrastructure. The VTR will take advantage of fast neutrons provided by this proven technology, along with a capability to rapidly insert, conduct, and remove state-of-the-art experiments. An advantage of the VTR is that it can support future innovations in experimental capabilities without modifying the facility. The VTR will support progress in a variety of science and technology areas, including testing and qualification of advanced reactor fuels; testing and qualification of innovative structural materials; testing of innovative components and instruments; validation of advanced modeling and simulation tools; and versatility for future technical missions. Through proven technology, the VTR can take advantage of existing reactor designs and operating experience to reduce the risk, cost, and time for design and construction. The top available resources of DOE laboratories, industry, and universities will be used to expedite reactor design and construction toward developing the scientific infrastructure that affords a strong testing capability that can be sustained over many years. This report summarizes current knowledge of natural and cultural resources primarily within the VTR construction area. At the time of this report, the proposed VTR site design includes a construction area of ~150.4 acres (~69.9 ha), which contains an ~51.3-acre (~20.8-ha) operations area, located within forested natural areas of the ORR. The primary goal of the work presented here was to evaluate potential effects on sensitive resources that might result from development and construction activities associated with VTR. In addition to on-the-ground surveys during spring and summer 2020 by the ORNL Natural Resources Management Program and Aquatic Ecology Group staff, this report makes use of historical (pre-1995) and contemporary (1995 to present) data from additional confirmed sources (e.g., TDEC). Likewise, forest conditions were compiled from a 2011 forest inventory and supplemented with limited ground observations in 2020. The individuals who obtained and compiled the data presented here are familiar with and routinely assess sensitive resources on the ORR. Anyone who references this report must consider that the timing of surveys did not permit a complete delineation of the resources that will be affected. If the VTR project proceeds, additional surveys will be required to account for the seasonal patterns of various threatened and endangered species. Data deficiencies and potential resources that likely went undetected are indicated where possible. Accordingly, this report should facilitate more environmentally sound decisions during planning and 2 development of the VTR site, provide a foundation for further assessment of sensitive and cultural resources, and help project managers better address regulatory guidance and DOE policies on sustainable development in compliance with, for example, the US Endangered Species Act (ESA), Migratory Bird Treaty Act (MBTA), Tennessee Rare Plant Protection and Conservation Act of 1985, Tennessee Nongame and Endangered or Threatened Wildlife Species Conservation Act of 1974, several federal and state regulations regarding aquatic resource protection, and site-specific policies as outlined in various ORR management plans developed by ORNL and TWRA for DOE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1983

Catalytic oxidation of very fuel rich mixtures

The objective of this research is to demonstrate the feasibility of using a catalytic reactor as a tool to study soot formation from the fuel rich side of the soot limit (null set = 3 - 6). The experimental approach to be taken in the first phase of the research is to document that a hydrocarbon fuel can be burnt at very rich equivalence ratios without forming soot. A simple mono-component fuel, iso-octane, will be used as the test fuel. To insure that combustion is uniform across the catalyst bed, measurements will be made of the fuel-air equivalence ratio profile across the inlet and the temperature and product distribution across the outlet. Phase Two will be to use this environment as a testing ground for determining the effect the structure of a hydrocarbon fuel has on its tendency to form soot. Various amounts of organic compounds such as benzene will be added to the iso-octane and the reaction products studied. Other compounds to be tested will xylene, toluene, and naphthalene.

Brabbs, T. A.↗
DOE OSTI2018

Potential Research and Development Opportunities for Light Water Reactor Spent Nuclear Fuel at INL

This report documents the commercial spent nuclear fuel currently in storage at Idaho National Laboratory with a specific focus on the potential research and development (R&D) opportunities for this material. Commercial spent nuclear fuel is stored in dry casks on the INL site at the Idaho Nuclear Technology and Engineering Center (INTEC, formerly the Idaho Chemical Processing Plant or ICPP). This spent fuel was part of the DOE-industry cooperative dry cask demonstration project that began in 1983 and from historical fuel testing and treatment activities that had been performed at the Test Area North (TAN)-607 facility. Most of the fuel is “low burnup,” but given that it was removed from the respective reactors between 1970 and 1985, it was representative for its time. The majority of the fuel is of the 15×15 Westinghouse PWR variety, though additional fuel types (e.g., TN-BRP) and configurations are also stored. Potential R&D opportunities identified are (1) the REA-2023 cask, (2) the TN-BRP cask, and (3) the Castor V/21 cask. Fuel from the TN-24P may be an alternative to the intact assemblies in the Castor V/21. Specifically, the REA-2023 cask contains 9 rods that have been exposed to air since 2005. These rods have a detailed pre-loading characterization and thus any changes in corrosion or geometry can most likely be attributed to the storage environment. In addition, similar rods to the nine in REA-2023 reside in the Castor V/21 cask whose environment has been maintained with inert nitrogen. The Castor V/21 cask also contains intact PWR assemblies that could be used to perform limited confirmation of the pyroprocessing approach. An alternative to the use of the Castor assemblies for pyroprocessing is the consolidated fuel that is stored in the TN-24P, which would not require removal from the assembly grid since it has already been packaged in accessible two-piece shell strongback containers. The TN-BRP cask contains a number of Gd-doped, low burnup assemblies that could be of use to validate code packages used to justify burnup-up credit for boiling water reactors (BWRs). The INL dry casks are stored on the CPP-2707 pad at INTEC, adjacent to the CPP-603 Irradiated Fuel Storage Facility, which incorporates a dry hot cell that has the potential to receive a full scale commercial cask of the types identified. Several evaluations have been made regarding this task, including full scale mockup testing of the physical clearances required to remove and handle fuel assemblies and components. A proposed schedule of 14 months and cost of $5.3 million was the outcome of analysis performed by Fluor Idaho and is included as Appendix A & B. In addition to the spent fuel at INL, there is also a full-scale mockup assembly that was fabricated for the Protypical Consolidation Demonstration Project to validate fuel rod consolidation approaches. The assembly was fabricated by Westinghouse as a 15 x 15 PWR, using Zircaloy-4 tubing and steel end fittings in accordance with product specifications. The only departure from specification is the use of copper metal in lieu of uranium oxide pellets to achieve a weight that is within 10% of the standard.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1979

Fuel quality combustion analysis

A high pressure research combustor operating over a wide range of burner inlet conditions was used to determine the effects of fuel molecular structure on soot formation. Six test fuels with equal hydrogen content (12.8%) were blended to stress different molecular components and final boiling points. The fuels containing high concentrations (20%) of polycyclic aromatics and partially saturated polycyclic structures such as tetralin, produced more soot than would be expected from a hydrogen content correlation for typical petroleum based fuels. Fuels containing naphthenes such as decalin agreed with the hydrogen content correlation. The contribution of polycyclic aromatics to soot formation was equivalent to a reduction in fuel hydrogen content of about one percent. The fuel sensitivity to soot formation due to the polycyclic aromatic contribution decreased as burner inlet pressure and fuel/air ratio increased.

Naegeli, D. W.↗
NASA NTRS1980

Effect of fuel molecular structure on soot formation in gas turbine engines

A high-pressure research combustor operating over a wide range of burner inlet conditions was used to determine the effects of fuel molecular structure on soot formation. Six test fuels with equal hydrogen content (12.8 percent) were blended to stress different molecular components and final boiling points. The fuels containing high concentrations (20 percent) of poly-cyclic aromatics and partially saturated polycyclic structures such as tetralin, produced more soot than would be expected from a hydrogen content correlation for typical petroleum based fuels. However, fuels containing naphthenes, such as decalin, agreed with the hydrogen content correlation. The contribution of polycylic aromatics to soot formation was equivalent to a reduction in fuel hydrogen content of about 1%. The fuel sensitivity to soot formation due to the polycyclic aromatic contribution decreased as burner inlet pressure and fuel/air ratio increased.

Naegeli, D. W.↗
DOE OSTI2026

Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel

Transient testing of nuclear fuel involves the evaluation of fuel performance under off-normal and accident conditions and is essential for proving design performance. Instrumentation included in such experiments commonly includes thermocouples attached to the outer surface of the cladding to provide temperature measurements throughout the transient. However, the presence of thermocouples on the cladding surface can alter the local heat transfer characteristics with the surrounding coolant. These localized effects can influence the temperature of the nearby cladding surface and introduce uncertainties in interpreting the thermocouple data. Understanding the impact of thermocouples attached to the outer surface of the cladding is crucial for accurate data interpretation as well as its effect on the thermomechanical behavior of the cladding. This paper presents a novel methodology for simulating the impact of outer cladding thermocouples during transient testing of nuclear fuels. The simulation framework leverages the thermal-hydraulic capabilities of RELAP5-3D coupled to the BISON fuel performance code through the RELAPCouplingApp interface. The methodology is compared against Accident Tolerant Fuel Reactivity Initiated Accident-1-E experiment performed at Idaho National Laboratory. The results reveal approximately 100°C difference between thermocouple-altered temperature and virgin cladding surface. The model overpredicts surface rewet time due to conservative correlations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microstructural Characterization of As-Cast MP-1 Experiment U-10Mo Alloy

The Mini-Plate-1 (MP-1) test is the first in a series of fuel testing campaigns with the purpose of achieving regulatory qualification for the U-10Mo monolithic plate-type fuel system. The objective of MP-1 is to assess fuel performance behavior of fuel plates fabricated by a commercial fuel fabricator (BWX Technologies, Inc.) and make sure that the fuel maintains mechanical integrity and geometric stability and behaves in a stable and predictable manner. As a part of MP1 fabrication campaign several castings were prepared. The castings were vacuum induction melted using HEU pieces mixed with a master alloy made up of DU and molybdenum pieces. The Master Alloy was poured by a commercial vendor. There was little mixing during melting. A three-plate mold was utilized to form three U-10Mo plates per melt pour. The work described here shows that the resulting microstructure varies greatly from one cast plate to another, and from the top to the bottom of each cast plate. The extent of variation in microstructural features such as grain size, second phase particle (SPP) distribution and morphology, microsegregation of constituents within the dendritic microstructure, porosity, and casting abnormalities are presented and discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accelerated Fuel Qualification of Fast Modular Reactor Fuel in a Thermal Reactor: Modeling and Simulation Paired with Irradiation Testing

The accelerated fuel qualification (AFQ) methodology is applied by simulating accelerated fuel tests of the General Atomics Electromagnetic Systems’ fuel system for its 44-MW(electric) gas-cooled, fast-spectrum fast modular reactor (FMR). This fuel is comprised of UO 2 pellets in SiGA® cladding, a silicon carbide ceramic matrix composite. Fast reactors, like the FMR, offer many benefits, including high fuel utilization and flexibility, but may require a lengthy material design process if tests are performed using fast neutron irradiation alone. A thermal neutron irradiation can instead be used to rapidly test how well key components of the current material models extend to high burnup. Thermal neutrons produce a different radial power distribution within the pin than fast neutrons. However, the temperature and burnup values for the two neutron types are comparable, and the differences between the simulated fuel responses are relatively small, demonstrating the weak sensitivity of the physics-based fuel model calculations on the neutron type and the irradiation rate. Furthermore, the deformation of the SiGA cladding saturates after about 1 displacement per atom for both neutron spectra. In an accelerated fuel test, the irradiation time required to reach the target fuel burnup can be reduced by a factor of 3 by using a small rodlet with a 45% smaller pellet diameter while maintaining the same linear power. Therefore, the time for data collection up to high burnup can be significantly reduced while maintaining the same temperature profile, which largely determines the material response. Tests of fuel rodlets of standard and compact size will be carried out in the Idaho National Laboratory’s Advanced Test Reactor (ATR), including full size and compact rodlets with varying gap sizes. By applying physics-based mechanistic modeling and simulation in accordance with the AFQ methodology, this type of compact rodlet testing in a thermal test reactor captures the necessary phenomena to test fuel material models up to high burnup and to simulate the expected impact of fast neutron radiation on the fuel in FMR operations. Finally, this approach to testing fast reactor fuels in existing thermal test reactors, paired with advanced physics-based mechanistic modeling and simulation, is expected to be applicable to a range of advanced fuels and will decrease the overall fuel qualification timeframe from decades to years.

Advanced test reactor (ATR)↗
NASA NTRS1983

Antimisting fuel breakup and flammability

The breakup behavior and flammability of antimisting turbine fuels subjected to aerodynamic shear are investigated. Fuels tested were Jet A containing 0.3% FM-9 polymer at various levels of degradation ranging from virgin AMK to neat Jet A. The misting behavior of the fuels was quantified by droplet size distribution measurements. A technique based on high resolution laser photography and digital image processing of photographic records for rapid determination of droplet size distribution was developed. The flammability of flowing droplet-air mixtures was quantified by direct measurements of temperature rise in a flame established in the wake of a continuous ignition source. The temperature rise measurements were correlated with droplet size measurements. The flame anchoring phenomenon associated with the breakup of a liquid fuel in the wake of bluff body was shown to be important in the context of a survivable crash scenario. A pass/fail criterion for flammability testing of antimisting fuels, based on this flame-anchoring phenomenon, was proposed. The role of various ignition sources and their intensity in ignition and post-ignition behavior of antimisting fuels was also investigated.

Parikh, P.↗
NASA NTRS1991

Thermal analysis of the FSP-1 fuel pin irradiation test

Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow.

Lyon, William F., III↗
DOE OSTI2019

The Fission Accelerated Steady State Test (FAST) -- A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

In an effort to improve the development and testing of novel nuclear fuels, a detailed evaluation on reducing the size of fuel specimens for irradiation has been performed. Theoretically speaking, reducing the fuels diameter and incorporating an inner capsule while maintaining the original outside diameter of the experiment significantly reduces the experiment’s sensitivity to fabrication tolerances and drastically reduces the time necessary to reach a desired fuel burnup; both of which improve the cost and schedule associated with the fuel testing. Simulations of this reduced diameter, double encapsulated concept have been performed using Abaqus, MCNP, and ORIGEN. The results of the Abaqus simulations are beyond the scope of this summary and will not be presented herein. For results of these Abaqus simulations, see Beausoleil [2]. The models were tested using various parameters and design deviations which were then compared to the original Advanced Fuels Campaign (AFC) designs. The results of these simulations confirmed that the introduction of a second capsule coupled to a reduction in the fuel pin diameter significantly reduces the experiment’s sensitivity to fabrication tolerances and reduces the time to reach a desired burnup by nearly an order of magnitude. The original AFC design requires approximately 11.7 years to reach 30 Atom % Heavy Metal (at% HM) burnup. A burnup of 30 at% HM can be reached in approximately 2.5 to 3.5 years (irradiation position dependent) with a fuel diameter ½ (~0.097 in. diameter) that of the original AFC design. An added benefit to the fuel pin diameter reduction is that more specimens can be irradiated in any given cycle. Essentially, significantly more specimens can be tested and incur more burnup over a given operating cycle relative to the AFC design. This allows the development of a large test matrix of samples with rapid throughput in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

MODELING A SPENT NUCLEAR FUEL CASK SEISMIC TEST

The US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a canister dry storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental US to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. The earthquake conditions will represent seismic hazards in the 2,000-to-20,000-year return period range. The test will use instrumented pressurized water reactor fuel assemblies with surrogate mass inside the fuel rods instead of radioactive fuel pellets. Dummy assemblies with similar mass and dimensions to actual fuel assemblies will occupy the rest of the fuel assembly locations in a SNF canister that holds 32 fuel assemblies. The canister will be located inside a mockup vertical concrete cask. Instrumentation will record the motion of the major components of the complex dynamic system, and strain gauges will be used to record the cladding strain at select locations. Preparations for the test require modeling predictions to identify the range of response of the system and to help select specific earthquake cases to be simulated on the shake table from a large set of potential cases. This paper describes the pretest nonlinear finite element modeling efforts that have been completed to date, including cask system level modeling and fuel assembly modeling in LS-DYNA. The cask system level models are critical for anticipating sliding or tipping of an unanchored cask during the test. The fuel assembly model is needed to estimate the range of cladding strain response and fuel assembly structural response to be expected during the test and cladding strain measurements will be key metrics for model validation and the development of modeling best practices after the test is concluded.

Klymyshyn, Nicholas A.↗
NASA NTRS1982

Low NO subx heavy fuel combustor concept program. Phase 1A: Coal gas addendum

The performance and emissions from a rich-lean combustor fired on simulated coal gas fuels were investigated using a 12.7-cm diameter axially-staged burner originally designed for operation with high heating value liquid fuels. A simple, tubular fuel injector was substituted for the liquid fuel nozzle; no other combustor modifications were made. Four test fuels were studied including three chemically bound nitrogen-free gas mixtures with higher heating values of 88, 227, and 308 kj/mol (103, 258 and 349 Btu/scf), and a 227 kj/mol (258 Btu/scf) heating value doped with ammonia to produce a fuel nitrogen content of 0.5% (wt). Stable, ultra-low nitrogen oxide, smoke-free combustion was attained for the nitrogen-free fuels. Results with the doped fuel indicated that less than 5% conversion of NH3 to nitrogen oxide levels below Environmental Protection Agency limits could be achieved. In some instances, excessive CO levels were encountered. It is shown that use of a burner design employing a less fuel-rich primary zone than that found optimum for liquid fuels would yield more acceptable CO emissions.

Rosfjord, T.↗
DOE OSTI2024

Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion

The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1980

Initial characterization of an Experimental Referee Broadened-Specification (ERBS) aviation turbine fuel

Characterization data and a hydrocarbon compositional analysis are presented for a research test fuel designated as an experimental referee broadened-specification aviation turbine fuel. This research fuel, which is a special blend of kerosene and hydrotreated catalytic gas oil, is a hypothetical representation of a future fuel should it become necessary to broaden current kerojet specifications. It is used as a reference fuel in research investigations into the effects of fuel property variations on the performance and durability of jet aircraft components, including combustors and fuel systems.

Prok, G. M.↗
DOE OSTI2022

The impact of fuel and injection strategy on combustion characteristics, emissions and efficiency in gasoline compression ignition operation

Gasoline compression ignition in diesel engines has been proposed in order to meet increasingly stringent emission regulations without sacrificing efficiency. In this study, a six-cylinder heavy-duty diesel engine was operated in a mixing controlled gasoline compression ignition mode to investigate the influence of fuels and injection strategies on the combustion characteristics, emissions, and thermal efficiencies. Fuels, including ethanol (E), isobutanol (IB), and diisobutylene (DIB), were blended with a gasoline fuel to form E10, E30, IB30, and DIB30 based on volumetric fraction. These four blends along with gasoline formed the five test fuels. With these fuels, three injections strategies were investigated, including late pilot injection, early pilot injection, and port fuel injection/direct injection. The impact of moderate exhaust gas recirculation on nitrogen oxides and soot emissions was examined to determine the most promising fuel/injection strategy for emissions reduction. In addition, first and second law analyses were performed to provide insights into the efficiency, loss, and exergy destruction of the various gasoline fuel blends at low and medium load conditions. Overall, the emission output, thermal efficiency, and combustion performances of the five fuels were found to be similar and their differences are modest under most test conditions. E30 with the port/direct fuel injection strategy obtained the peak brake thermal efficiency (46.9%) and gross indicated thermal efficiency (52.2%) at 14 bar. At 7 bar, the peak brake thermal efficiency (44.2%) and gross indicated thermal efficiency (52.5%) also belonged to E30 with the late pilot injection strategy. The second law analysis revealed that exergy destruction was mostly influenced by the fuel type and not by the injection strategy for the 14 bar operating conditions. However, the opposite trend was observed for the 7 bar case. In addition, port fuel injection in combination with oxygenated fuel blends was found to be helpful to suppress soot emissions for operation with exhaust gas recirculation.

co-optima fuels↗
NASA NTRS1987

Carbon deposition characteristics of LO2/HC propellants

The generation and deposition of carbon have been studied using subscale hardware with LO2/RP-1, LO2/propane, and LO2/methane at low mixture ratio conditions. The deposition of carbon on the turbine simulator tubes was evaluated at mixture ratios of 0.20 to 0.60, and at chamber pressures from 720 to 1650 psia. The carbon-deposition rate is a strong function of mixture ratio and a weak function of chamber pressure. There is a mixture ratio that will minimize deposition for LO2/RP-1; a threshold mixture ratio for LO2/propane; and no deposition for LO2/methane at any mixture ratio tested. The turbine drive operating limits were defined for each fuel tested.

Hernandez, Rosemary↗