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NASA NTRSDate not supplied

Performance Analysis of Aerocapture Systems for Uranus Orbiters

A Uranus orbiter and probe mission is the highest priority science mission of the current decade. Aerocapture can be employed to support these missions by enabling shorter interplanetary trajectories and requiring less fuel for orbit insertion. This paper investigates the trajectory design and performance analysis of Uranus aerocapture using an MSL-derived aeroshell design. The trajectory tradespace for Uranus aerocapture is investigated to understand the relationship between interplanetary arrival speed and aeroshell aerodynamics to controllability. A 3 degree-of-freedom simulation framework is developed to assess the performance of bank angle fully numerical predictor-corrector aerocapture guidance. A series of Monte Carlo sensitivity studies are conducted to assess the effects that arrival navigation, arrival speeds, and atmosphere knowledge have on the aerocapture robustness and performance. The results suggest that bank angle modulation is a feasible option for Uranus orbit insertion where aerocapture can reduce transit times by 40% and save 1950 kg in propellant mass.

Rohan G. Deshmukh↗

NASA NTRSDate not supplied

Performance Analysis of Aerocapture Systems for Uranus Orbiters

Rohan Deshmukh↗

DOE OSTI2019

Status PIE Report on Legacy EBR-II and FFTF Metallic Fuel Experiments

A long term research focus of the US Department of Energy (DOE) Advanced Fuels Campaign (AFC) is to investigate technologies that can allow for increased actinide utilization in nuclear fuel for fast neutron spectrum reactors. In order to reach these goals, many experimental campaigns have been tested in an approximated fast reactor neutron spectra at the Idaho National Laboratory Advanced Test Reactor (ATR) in order to study the performance of metallic, nitride, ceramic and advance fuel forms. In addition and complementary to ATR testing, systematic characterization of experiments irradiated in true fast reactors are performed in order to assess ATR in qualifying fast reactor fuel. For this purpose, complementary postirradiation examinations (PIE) are performed on several metallic fuel pins previously irradiated in the EBR II. Advanced PIE techniques developed since the initial PIE of EBR II samples in the 1980’s have been applied in new examinations of irradiated samples during past EBR-II operations. The EBR-II metallic fuel pins analyzed in this report are X501 G591, an early minor actinide-bearing metallic fuel alloy; X430-T653, a U-19Pu-10Zr metallic fuel alloy with a larger diameter compared to EBR-II driver fuel, and more prototypic of future commercial fast reactors; X521-G594, a unique experiment to test the feasibility to burn weapon grade Pu without reprocessing; X496 CL47&CL39 from a long-life metallic fuel experiment with the goal to reach very high burnup; and X486-J555 a U-10Zr Mk-IIIA qualification fuel test that will serve as first of a kind “modelling driven PIE.” These legacy pins were evaluated against the historical fuel performance of previously irradiated fuel from literature. Fuel performance was investigated through a suite of PIE techniques including neutron radiography, gamma spectrometry, dimensional inspection, fission gas release measurements, chemical burnup analysis, optical microscopy, and advance microscopy. So far, the PIE campaign revealed fuel performance and characteristics that are generally in line with literature or EBR-II experience. Some exceptions are present, and they are directly discussed in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1989

Space Shuttle Main Engine structural analysis and data reduction/evaluation. Volume 3B: High pressure fuel turbo-pump preburner pump bearing assembly analysis

The analysis performed on the High Pressure Oxidizer Turbopump (HPOTP) preburner pump bearing assembly located on the Space Shuttle Main Engine (SSME) is summarized. An ANSYS finite element model for the inlet assembly was built and executed. Thermal and static analyses were performed.

Power, Gloria B.↗

DOE OSTI2016

TREAT neutronics analysis and design support, Part I: Multi-SERTTA

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.73 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Statistical Analysis in Support of Fuel Temperature Control in the Advanced Gas Reactor Experiments

There are four completed tests among a series of Advanced Gas Reactor (AGR) irradiation experiments that are being conducted within the Advanced Reactor Technologies (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaigns are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and hypothetical accident conditions, and support development and validation of fuel performance and fission product transport models and codes. Since this tri-structural isotropic fuel is qualified to be used in high temperature gas-cooled reactors (HTGRs), it is essential that the testing conditions of fuel irradiations are designed to mimic those in the conceptual gas reactors. As a result, temperatures of the AGR test fuel have to be maintained within a specified range that is typical of the HTGRs. In each AGR capsule, the test fuel temperature is controlled independently using a variable fraction neon and helium gas mixture that fills the gap volume for each capsule. The temperature as measured by a thermocouple (TC) terminated in the graphite holder (which houses fuel compacts) is used as a feedback for the gas flow controllers. Direct temperature measurement for fuel compacts is not possible because contact between a TC head and fuel particle could lead to undesirable failures. Therefore, an ABAQUS-based finite element thermal model was created to predict the daily average fuel temperatures during the entire irradiation period. It is well-known that extended exposure to neutron irradiation and high temperatures leads to changes in geometries of fuel compacts and graphite holder, which results in changes of the gas gap, leading to a change in the relationship between TC and fuel temperatures. In addition, TCs are expected to drift and sometimes fail under irradiation conditions. These facts make maintaining test fuel temperature within a specified range challenging. This paper describes how the regression analysis methods are used in conjunction with calculated fuel temperatures from early irradiation periods (when the irradiation effect is relatively small) to effectively control fuel temperature for the remaining irradiation (when the irradiation effect became profound). This is particularly important in the event when all TCs in a capsule fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Development of a layered two-dimensional approach for modelling fuel rod behaviour

Historically, light water reactor (LWR) fuel performance codes have been developed using layered one-dimensional (1D) or two-dimensional (2D) axisymmetric approaches. In a layered 1D analysis the fuel rod is divided into a number of axial layers, each of which is represented using a 1D axisymmetric model with generalized plane strain assumptions. These simulations are typically solved using a finite difference technique. A limited set of data such as fission gas behaviour, has an effect on all layers. The 2D axisymmetric approach improves the axial fidelity of the simulation by employing a finite element representation of the fuel rod geometry, and more readily allows for phenomena such as axial heat transfer to be represented. Depending upon the fuel rod analysis being investigated the assumptions made in layered 1D or 2D axisymmetric approaches can result in important physics not being adequately captured. Therefore, the Bison fuel performance code has been developed over the last 10 years to provide flexibility in modelling of fuel behaviour. The code is capable of investigating fuel behaviour in 1D, layered 1D, 2D, 2D-axisymmetric, and 3D geometric representations depending upon the fidelity required for a particular analysis. Recently, a new layered 2D approach has been developed and implemented. This approach utilizes a similar technique to layered 1D except the thermo-mechanical analysis in each layer is solved on layered disks that can have azimuthally varying boundary conditions. These azimuthally varying conditions are important for a variety of fuel rod applications including discrete fuel fracture and rod behaviour during Loss of Coolant Accidents (LOCA). This paper describes the development of the layered 2D approach and the types of analyses for which it is applicable. Results are presented that illustrate the improvement in model predictions provided by this approach over layered 1D and 2D-axisymmetric approaches for the LOCA rod behaviour.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Whole core analysis of the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

The SD-TMSR (2,250 MW th ) is a Single-fluid Double-zone Thorium-based Molten Salt Reactor. The active core of the SD-TMSR is divided into the inner zone (486 fuel tubes) and the outer zone (522 fuel tubes) to improve the Th-U3 breeding performance. This work adopted the SERPENT-2 Monte Carlo code to analyze the whole core model of the SD-TMSR. Built-in SERPENT-2 capabilities simulated online reprocessing and refueling and calculated the multiplication factor and Breeding Ratio (BR). We found that the molten salt Temperature Coefficient of Reactivity ( α T ) was negative for initial and equilibrium states. This study investigated the variation of the multiplication factor, BR, and build-up of important nuclides in the core as a function of burnup. Under online reprocessing and refueling, we studied the variation of the reactivity during 60 years of reactor operation. Additionally, the neutron spectrum shift during the reactor operation was calculated. Finally, these simulations determined the appropriate 232 Th and 233 U feed rates for maintaining criticality and enabled analysis of the overall SD-TMSR fuel cycle performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

A Standardized Analysis Process Using Digital Image Correlation to Calculate In Situ Cladding Strain from Modified Burst Tests for Fuel Performance Code Validation

Historical data collection on nuclear fuel cladding materials has focused on generating a statistically significant amount of data to assess the material and its failure behavior. Furthermore, data generated to support material model and failure criteria development were previously posttest evaluations, so a large number of tests was required to gain new understanding. A way to expedite this process is to develop techniques capable of generating large, high-fidelity data sets from a single test with lower uncertainty or quantified uncertainty. One such example of this approach is Oak Ridge National Laboratory’s use of modified burst tests (MBTs) to analyze the mechanical behavior and failure conditions of cladding during a simulated reactivity-initiated accident (RIA). Each test incorporates digital image correlation (DIC) analysis techniques that are used to assess the accumulated strain in situ, as well as eventual cladding failure. This work has been fruitful in defining strain-to-failure conditions for materials like silicon carbide (SiC) fiber–reinforced/SiC matrix composite tubes (SiC/SiC), iron-chromium-aluminum (FeCrAl) alloy tubes, and chromium-coated Zircaloy-4 tubes. However, there are numerous DIC software available, including open-source and proprietary software. The different DIC software use various algorithms to process images and calculate displacement values. Using these different software and algorithms can lead to varying results, and perhaps larger-than-expected uncertainties. In the present study, previously published MBT data encompassing a variety of test conditions were reanalyzed with two different DIC software to assess the variance in the calculated strain results. The data consisted of SiC/SiC, FeCrAl, and chromium-coated Zircaloy-4 tubes. Plots of the calculated strains during the transient revealed good agreement between the two DIC software. The average root-mean-square errors between the two software was 0.20% strain, which is slightly larger than a previously reported error value for these tests. In conclusion, this variance in results is low enough that this analysis method can be used for code validation.

Reactivity-initiated accident↗

DOE OSTI2016

Analyses of SPERT-CDC test 859 by FEMAXI-7 and RANNS codes

In the current Japanese regulation concerning fuel safety of water-cooled reactors, the criterion of fuel failure due to pellet-cladding mechanical interaction (PCMI) in a burnup range of 25-40 GWd/t is determined substantially based on the result of SPERT-CDC test 859 (SPERT859). However, there is room for discussion if the result of SPERT859 test rod with a burnup of 31.8 GWd/t represents the failure behavior of the fuel rod for commercial light water reactor because it was reported in some literatures that the cladding of SPERT859 test rod corroded excessively during pre-irradiation. On the other hand, the oxide thickness on the cladding of SPERT859 test rod was not reported in the data of its post-irradiation examinations (PIEs) after pre-irradiation, and the probability of the excessive corrosion on the cladding of the test rod and the effect of such excessive corrosion on the fuel enthalpy at failure of the test rod has not yet been clarified. In this study, the oxide thickness of the cladding formed on the cladding outer surface of SPERT859 test rod and its fuel enthalpy at failure due to PCMI under this corrosion condition were analyzed by using fuel performance codes FEMAXI-7 and RANNS. From the analysis results of FEMAXI-7, it was found that the excessive corrosion of the cladding of the test rod should be assumed to reproduce the fuel outer diameter measured in PIEs after pre-irradiation. The fuel enthalpy at failure of the test rod was calculated by using RANNS, and the calculated fuel enthalpy at failure was relatively close to the value of SPERT859 test rod. These results suggested that the fuel enthalpy at failure of SPERT859 was affected by the excessive corrosion on the cladding of the test rod and was likely lower than that of the typical fuel for light water reactors. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1991

Evaluation of on-board hydrogen storage methods f or high-speed aircraft

Hydrogen is the fuel of choice for hypersonic vehicles. Its main disadvantage is its low liquid and solid density. This increases the vehicle volume and hence the drag losses during atmospheric flight. In addition, the dry mass of the vehicle is larger due to larger vehicle structure and fuel tankage. Therefore it is very desirable to find a fuel system with smaller fuel storage requirements without deteriorating the vehicle performance substantially. To evaluate various candidate fuel systems, they were first screened thermodynamically with respect to their energy content and cooling capacities. To evaluate the vehicle performance with different fuel systems, a simple computer model is developed to compute the vehicle parameters such as the vehicle volume, dry mass, effective specific impulse, and payload capacity. The results indicate that if the payload capacity (or the gross lift-off mass) is the most important criterion, only slush hydrogen and liquid hydrogen - liquid methane gel shows better performance than the liquid hydrogen vehicle. If all the advantages of a smaller vehicle are considered and a more accurate mass analysis can be performed, other systems using endothermic fuels such as cyclohexane, and some boranes may prove to be worthy of further consideration.

Akyurtlu, Ates↗

DOE OSTI2020

BISON High Burnup Structure Modeling Capabilities Validated with a Selection of the Halden IFA-650 Rods

The U.S. Department of Energy (DOE)’s Nuclear Energy Advanced Modeling and Simulation (NEAMS)program aims to develop predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity and high-resolution fuel performance tool at the engineering scale. This document continues analysis and refinement of capabilities added to BISON early this calendar year in regards to the incorporation of capabilities applicable to extended burnups in response to industry interest. Details are provided on high burnup thermal conductivity models, a refitting of the high-burnup structure (HBS) porosity formation model to include additional data, the coupling of the HBS volume fraction model to thermal conductivity and fine fragmentation models, and validation activities. The IFA-650.4 and IFA-650.9 loss of coolant accident (LOCA) analyses are revisited with the latest developments in this report. A new validation case, IFA-650.14 has also been added to the BISON test suite.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Core and Fuel Cycle Performance of a Molten Salt Fast Reactor

This paper discusses the reactor core design and modeling approaches for a specific molten salt fast reactor (MSFR) design that is started up with high-assay low-enriched uranium (HALEU) within a chloride fuel salt and focuses on the technical assumptions and design choices. These include the fuel salt properties, first core and make-up heavy metal requirements, fuel composition evolution, equilibrium fission product concentration, etc. This MSFR concept was developed as part of a wider study that sought to compare the fuel cycle performance tradeoffs between deploying a fleet of sodium-cooled fast reactors (SFRs) and deploying a fleet of MSFRs. MSFRs that use flowing liquid fuel salt and on-line salt processing are of particular interest because these technology characteristics are significantly different from those of SFRs and can lead to major differences in fuel cycle impacts. The goal of this study was not to draw general conclusions based on a direct comparison of two very specific point designs for an SFR and MSFR, but rather to use this example MSFR design and associated parametric studies to gain insight into the key technology characteristics that impact fuel cycle performance as it relates to resource utilization (i.e., uranium consumption). The MSFR reactor physics, fission product removal, and fuel salt mass flows were modeled using Argonne’s recently-developed molten salt reactor (MSR) analysis tools while the time-dependent fuel cycle performance of this specific reactor was compared to that of a representative SFR design using the fuel cycle system dynamics code DYMOND.

chloride salt↗

DOE OSTI2017

Fuel element design and analysis for potential LEU conversion of the Advanced Test Reactor

This study provides a summary of the status of the design process for a low-enrichment replacement for the high-enrichment fuel that has historically been used in the Advanced Test Reactor (ATR) at Idaho National Laboratory. Under current long-term Department of Energy policy, the ATR will be converted to operate using low-enrichment fuel by 2030. To this end, an engineering evaluation and design process has been underway for more than ten years. Initially performed as a series of scoping studies for different design options, the last five years have seen a transition to a formal Safety-in-Design approach to the design and safety evaluation process. This paper focuses on such efforts in progress since 2012 to identify pre-conceptual and conceptual design candidates for ATR conversion. The Enhanced LEU Fuel (ELF) design is selected as the most promising conceptual design. Analyses and trade-off studies performed that led to that selection are described. Analysis of certain aspects of core performance for an ELF-fueled core relative to the current high-enrichment fuel show that only small and manageable changes to core operations will be necessary to support conversion of ATR to low-enrichment fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

TREAT Neutronics Analysis and Design Support, Part I: Multi-SERTTA

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.74 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Decay Heat Quantification in the High Temperature Test Reactor

The proper calculation of decay heat is necessary for both postulated accidents as well as normal spent fuel repository safety analysis. Since the 1970s, the nuclear industry has primarily relied on the use of the ANSI/ANS-5.1 standard or one of it’s predecessors as the tool for calculating such heat. However, different reactor systems can have substantially different materials, neutron spectrums, operating characteristics, etc.…that lead to vastly different isotopic compositions in the fuel over time. The ANSI/ANS-5.1 standard is specific to Light Water Reactors, and thus may not be well suited for other systems such as high temperature gas reactors. Recent developments in the SCALE 6 code developed at Oak Ridge National Laboratory have allowed for a more accurate treatment of the complexity associated with high temperature reactor TRISO fuel. Also, the standalone isotopic depletion and generation code ORIGEN-S, a module of SCALE, has been validated against experimental data for LWRs for properly characterizing decay heat. And most recently, Japan Atomic Energy Agency has used the ORIGEN code to generate HTGR cross section libraries to be used to quantify the decay heat generated in the High Temperature Engineering Test Reactor at the Oarai Research and Development Center. Given the Idaho National Laboratory’s continued research into the viability of HTGRs as a next generation power reactor, an independent decay heat quantification analysis has been performed for a variety of fuel enrichments, TRISO particle packing fractions, and fuel temperatures using a similar approach. Furthermore, these data are then used as input into the RAVEN code developed at INL. A Reduced Order Model mathematical model in RAVEN is ‘trained’ to create a response mechanism to the various input parameters. Preliminary results demonstrate the potential for using this methodology to accurately predict decay heat generation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

A Phenomenological Thermal Spray Wall Interaction Modeling Framework Applied to a High Temperature Ignition Assistant Device

Airborne compression ignition engines must operate with reliable ignition systems to achieve proper ignition at every cycle, particularly at high altitudes. Glow-plug-based ignition-assistant (IA) devices can provide the necessary energy to preheat the fuel and ensure ignitability of the fuel-air mixture. Ignitability of liquid sprays can be facilitated via direct impingement onto the hot IA surface, however this comes with adverse effects on the IA durability. Therefore, optimizing an IA’s design requires detailed understanding of the physics of fuel spray impingement of superheated surfaces. While spray impingement on relatively low wall temperatures has been extensively studied and appropriate numerical models have been proposed through the years, fundamental understanding of high-speed liquid spray impingement on superheated walls is still elusive. This work aims to formulate a phenomenological thermal spray-wall interaction framework for modeling the film-boiling-induced heat transfer, atomization, and dispersion of fuel spray droplets impinging onto a superheated IA device. A qualitative comparison of the new phenomenological model is performed against optical experiments from the literature of an F-24 fuel spray injected onto an IA device located 12 mm away from the injector tip. The temperature of the IA was set at 1400 K. The fuel injection pressure was 400 bar, while the ambient gas pressure and temperature were 30 bar and 800 K, respectively. The performance of the phenomenological model is evaluated in comparison with two other state-of-art models from the literature. A qualitative analysis of the different spray and fuel-air mixture characteristics is performed to outline the differences in the predictions offered by the new phenomenological model and the two state-of-art spray-wall interaction models.

droplet dispersion↗

DOE OSTI2020

Numerical Investigation of Fuel Property Effects on Mixed-Mode Combustion in a Spark-Ignition Engine

In this research, lean mixed-mode combustion is numerically investigated using computational fluid dynamics (CFD) in a spark-ignition engine. A new E30 fuel surrogate is developed using a neural network model with matched octane numbers. A skeletal mechanism is also developed by automated mechanism reduction and by incorporating a NO x submechanism. A hybrid approach that couples the G-equation model and the well-stirred reactor model is employed for turbulent combustion modeling. The developed CFD model is shown to well predict pressure and apparent heat release rate (AHRR) traces compared with experiment. Two types of combustion cycles (deflagration-only and mixed-mode cycles) are observed. The mixed-mode cycles feature early flame propagation and subsequent end-gas auto-ignition, leading to two distinctive AHRR peaks. The validated CFD model is then employed to investigate the effects of NO x chemistry. The NO x chemistry is found to promote auto-ignition through the residual gas, while the deflagration phase remains largely unaffected. Sensitivity analysis is finally performed to understand effects of fuel properties, including heat of vaporization (HoV) and laminar flame speed (S L ). An increased HoV tends to suppress auto-ignition through charge cooling, while the impact of HoV on flame propagation is insignificant. In contrast, an increased S L is found to significantly promote both flame propagation and end-gas auto-ignition. The promoting effect of S L on auto-ignition is not a direct chemical effect; it is rather caused by an advancement of the combustion phasing, which increases compression heating of the end-gas.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗