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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Exploring the potential benefits of high-efficiency dual-fuel combustion on a heavy-duty multi-cylinder engine for SuperTruck I

In support of the Daimler SuperTruck I team’s 55% brake thermal efficiency (BTE) pathway goal, researchers at Oak Ridge National Laboratory performed an experimental investigation of the potential efficiency and emissions benefits of dual-fuel advanced combustion approaches on a modified heavy-duty 15-L Detroit™ DD15 engine. For this work, a natural gas port fuel injection system with an independent injection control for each cylinder was added to the DD15 engine. For the dual-fuel strategies investigated, 65%–90% of the total fuel energy was supplied through the added port fuel injection natural gas (NG) fueling system. The remaining fuel energy was supplied by one or more direct injections of diesel fuel using the production high pressure diesel fueling system. The production DD15 air handling system and combustion geometry were unmodified for this study. Efficiency and emissions with dual-fuel strategies including both low temperature combustion (LTC) and non-LTC approaches such as dual fuel direct-injection were investigated along with control authority over combustion phasing. Parametric studies of dual-fuel NG/diesel advanced combustion were conducted in order to experimentally investigate the potential of high-efficiency, dual-fuel combustion strategies to improve BTE in a multi-cylinder engine, understand the potential reductions in engine-out emissions, and characterize the range of combustion phasing controllability. Characterization of mode transitions from mixing-controlled diesel pilot ignition to kinetically controlled ignition is presented. Key findings from this study included a reproducible demonstration of BTE approaching 48% at up to a 13-bar brake mean effective pressure with significant reductions in engine-out NOx and soot emissions. Additional results from investigating load transients in dual-fuel mode and initial characterization of particle size distribution during dual-fuel operation are presented.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2021

Global Ethanol-Blended-Fuel Vehicle Compatibility Study

The objective of this study is to understand the impact of ethanol-blended fuel at various blending levels (10%, 15%, and 20% vol.) on "in-operation" vehicles built to differing emissions and manufacturing standards around the world. The study focuses on vehicles used in Canada, China, India, Indonesia, Japan, South Korea, and Mexico. Historical experience in the United States and Brazil informs the analysis. The primary study question is: Are vehicles in targeted countries physically and operationally compatible with ethanol blended fuel? For a fuel to be compatible with a vehicle, the fuel must perform its function as part of the integrated fuel-vehicle system, meaning: the car should start easily and drive normally, the fuel should not cause catastrophic fuel system leaks, the fuel should not cause corrosion or degradation of any engine or fuel system components (including emissions control components). The history of E10 use in the United States, beginning in 1978, was evaluated and shows no reliability or operability issues for cars dating back to pre-emissions-control times -- and likely included many cars manufactured in the 1960s. This strongly supports the contention that fuel chemistry and property differences between E0 and E10 are so small that any car made to international standards in the last 50 years will have a very high probability of being fully compatible with E10. This conclusion is supported by the experience in Brazil in the 1970s, where E10 was also introduced, and ethanol blending for conventional cars rapidly ramped up to even higher blend levels. A limited number of fuel system and component manufacturers supply the global market, including Bosch, Continental, Denso, Delphi, and Visteon. To reduce complexity, ethanol-compatible materials began to be integrated in fuel system designs globally. Fuel systems evolved over the following decades to incorporate ethanol-compatible materials with core subsystem families, such as in-tank fuel pumps used across several global vehicle original equipment manufacturers (OEMs). A similarly compelling case can be made that all cars at the Tier 1 (or equivalent) emissions-control technology level or higher are fully compatible with E15 blends, based on the data evaluated by the U.S. Environmental Protection Agency (EPA) and Ricardo in 2010. For cars at this technology level, the minor differences in fuel chemistry and properties between E10 and E15 are not significant. For E20, studies are not as extensive but are still highly significant. A long-term durability study conducted on mileage accumulation dynamometers presents convincing evidence that Tier 2 technology level cars have materials of construction and engine control authority for compatibility with E20, although this conclusion is not as strong as those drawn for E10 and E15, which are also partly based on real-world experience.

09 BIOMASS FUELS↗
DOE OSTI2013

Analysis of fuel options for the breakeven core configuration of the Advanced Recycling Reactor

A trade-off study is performed to determine the impacts of various fuel forms on the core design and core physics characteristics of the sodium-cooled Toshiba- Westinghouse Advanced Recycling Reactor (ARR). The fuel forms include oxide, nitride, and metallic forms of U and Th. The ARR core configuration is redesigned with driver and blanket regions in order to achieve breakeven fissile breeding performance with the various fuel types. State-of-the-art core physics tools are used for the analyses. In addition, a quasi-static reactivity balance approach is used for a preliminary comparison of the inherent safety performances of the various fuel options. Thorium-fueled cores exhibit lower breeding ratios and require larger blankets compared to the U-fueled cores, which is detrimental to core compactness and increases reprocessing and manufacturing requirements. The Th cores also exhibit higher reactivity swings through each cycle, which penalizes reactivity control and increases the number of control rods required. On the other hand, using Th leads to drastic reductions in void and coolant expansion coefficients of reactivity, with the potential for enhancing inherent core safety. Among the U-fueled ARR cores, metallic and nitride fuels result in higher breeding ratios due to their higher heavy metal densities. On the other hand, oxide fuels provide a softer spectrum, which increases the Doppler effect and reduces the positive sodium void worth. A lower fuel temperature is obtained with the metallic and nitride fuels due to their higher thermal conductivities and compatibility with sodium bonds. This is especially beneficial from an inherent safety point of view since it facilitates the reactor cool-down during loss of power removal transients. The advantages in terms of inherent safety of nitride and metallic fuels are maintained when using Th fuel. However, there is a lower relative increase in heavy metal density and in breeding ratio going from oxide to metallic or nitride Th fuels relative to the U counterpart fuels. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Innovative Implementation of Consolidated Dry Fuel Storage at the INL - 19655

An innovative nuclear-fuel storage bucket design provides increased fuel storage capacity at the Irradiated Fuel Storage Facility (IFSF) on the Idaho National Laboratory (INL) Site. The IFSF is a dry storage facility that is at approximately 93% of its fuel storage capacity. This innovative fuel storage bucket provides for higher density packaging, thereby allowing for the receipt of more fuel at the IFSF and subsequent completion of the wet-to-dry fuel transfer scope at the INL's Idaho Nuclear Technology and Engineering Center (INTEC). The subject fuel originates from the Advanced Test Reactor (ATR) at the INL Site. All spent nuclear fuel (SNF) at INTEC is required by the Idaho Settlement Agreement (ISA) to be out of wet storage by December 31, 2023. The ISA is a tri-party agreement between the U.S. Department of Energy, the U.S. Navy, and the State of Idaho. Currently, the INTEC CPP-666 pools contain all of the INL's SNF remaining in wet storage, and INTEC's Spent Fuel Project is on schedule to meet the ISA deadline. Interestingly, the INL Site's environmental cleanup contract awarded to Fluor Idaho in 2016 also initially required the Spent Fuel Project to continue SNF receipts from the ATR into wet storage at the CPP-666 pools. The continued receipts of ATR SNF into the CPP-666 basin were able to be redirected to the IFSF because of the new storage package design, enabling the Spent Fuel Project to meet the ISA requirement to have all SNF out of wet storage by 2023. Complicating the need to redirect ATR receipts from wet storage is the lack of interim storage capacity at the ATR site. This lack of ATR storage capacity threatened to terminate reactor operations if the SNF could not be received at INTEC. The contractual requirement for continued ATR SNF receipts to prevent termination of reactor operations is resolved by the implementation of the innovative consolidated dry storage configuration in the IFSF. Consolidated fuel storage at the IFSF provides increased fuel storage capacity to facilitate meeting the ISA and the Fluor Idaho contract requirements. It provides for the completion of the ATR wet-to-dry fuel transfer scope and supports continued operation of the ATR by allowing the receipt of additional ATR SNF directly into IFSF dry storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗
DOE OSTI2023

Sensitivity of UO 2 fuel performance to microstructural evolutions driven by dilute additives

Use of dilute additives to nuclear fuel is being considered to increase the security of commercial fuel management through traceability of fabricated fuel elements. Taggants, as additives are denoted when included for traceability purposes, may also improve fuel performance, as demonstrated in Cr-containing uranium dioxide as described in the literature, and they may also improve fuel safety. In fact, studies have shown that some additives affect fuel material properties such as grain size and density after sintering. Given the possible range of elements that could be used as additives, the impact of such fuel property variations on the fuel’s thermomechanical behavior becomes relevant. These effects can be evaluated through a sensitivity study of standard fuel models to analyze changes in these properties using a fuel performance code. In this work, the BISON code is being used to investigate these effects through a 2D axisymmetric model of smeared UO 2 fuel pellets and ZIRLO® cladding under realistic pressurized water reactor core irradiation conditions. Here, randomly sampled densities and grain sizes within specified ranges are used as input parameters in the simulations, and several fuel model-related outputs are evaluated. The thermomechanical response of the cladding is also addressed in this study. The simultaneous variation of both input parameters offers a more comprehensive path to identify key sensitivities. Outputs explored include temperature, fission gas release, creep, and radial stress. Results show that although most of these outputs are sensitive to grain size to a certain extent, density mainly affects fuel temperature and elastic strain. Furthermore, sensitivities can vary depending on the radial position within the fuel pellet.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

BISON: A Flexible Code for Advanced Simulation of the Performance of Multiple Nuclear Fuel Forms

BISON is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library. One of its major goals is to have a great amount of flexibility in how it is used, including in the types of fuel it can analyze, the geometry of the fuel being modeled, the modeling approach employed, and the dimensionality and size of the models. Fuel forms that can be modeled include standard light water reactor fuel, emerging light water reactor fuels, tri-structural isotropic fuel particles, and metallic fuels. BISON is a platform for research in nuclear fuel performance modeling while simultaneously serving as a tool for the analysis of nuclear fuel designs. Recent research in BISON includes techniques such as the extended finite element method for fuel cracking, exploration of high-burnup light water reactor fuel behavior, swelling behavior of metallic fuels, and central void formation in mixed-oxide fuel. BISON includes integrated documentation for each of its capabilities, follows rigorous software quality assurance procedures, and has a growing set of rigorous verification and validation tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Parametric Study of Used Nuclear Oxide Fuel Constituent Dissolution in Molten LiCl-KCl-UCl 3

Prior work identified dissolution of used nuclear oxide fuel constituents from a uranium oxide matrix into molten LiCl-KCl-UCl 3 at 500°C, prompting a subsequent series of three progressive studies (including an initial scoping study, an electrolytic dissolution study, and a chemical-seeded dissolution study) to further investigate associated parameters and mechanisms. Thermodynamic calculations were performed to identify possible reaction mechanisms and their propensities in used oxide fuel constituent dissolution. Used nuclear oxide fuels with varying preconditions from fast and thermal test reactors were separately immersed in the subject salt system to assess fuel constituent migration from the bulk fuel matrix to the salt phase in an initial scoping study. Dissolution of expected fuel constituents, including alkali, alkaline earth, lanthanide, and transuranium oxides, into the chloride salt phase varied widely, ranging from 12% to 99% in the initial study. Uranium isotope blending between the salt phase and bulk fuel matrix was also observed, which was attributed to reducing conditions in the fuel matrix. Electrolytic and chemical-seeded dissolution studies were subsequently performed to effect reducing conditions in the fuel. Other parameters, including temperature (at 500°C, 650°C, 725°C, and 800°C) and uranium trichloride concentrations (at 6, 9, and 19 wt% uranium), were investigated in the latter two studies, resulting in fuel constituent dissolution above 90%. Extents of dissolution were based on initial and final fuel constituent concentrations in the oxide fuels following operations in the salt and subsequent removal of the salt via distillation. Finally, in this series of progressive studies, oxide fuel preconditioning and in situ reducing conditions, along with elevated temperature and uranium trichloride concentrations, were the primary parameters promoting used nuclear oxide fuel constituent dissolution in accordance with identified reaction mechanisms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Direct Coal -to-Liquids (CTL) for Jet Fuel Using Biomass-Derived Solvents

Battelle has demonstrated a novel and potentially breakthrough technology for a direct coal-to-liquids (CTL) process for producing jet fuel using biomass-derived coal solvents (bio-solvents). The Battelle process offers a significant reduction in capital and operating costs and a substantial reduction in greenhouse gas (GHG) emissions, without requiring carbon capture and storage (CCS). The results of the project are the advancement of three steps of the hybrid coal/biomass-to-jet fuel process to the technology readiness level (TRL) of 5. The project objectives were achieved over two phases. In Phase 1, all three major process steps were explored and refined at bench-scale, including: (1) biomass conversion to high hydrogen-donor bio-solvent; (2) coal dissolution in biomass-derived bio-solvent, without requiring molecular H 2 , to produce a synthetic crude (syncrude); and (3) two-stage catalytic hydrotreating/hydrogenation of syncrude to jet fuel and other distillates. In Phase 2, all three subsystems of the CTL process were scaled up to a pre-pilot scale, and an economic analysis was carried out. A total of over 40 bio-solvents were identified and prepared. The most unique attribute of Battelle’s bio-solvents is their ability to provide much-needed hydrogen to liquefy coal and thus increase its hydrogen content so much that the resulting syncrude is liquid at room temperature. Based on the laboratory-scale testing with bituminous coals from Ohio and West Virginia, a total of 12 novel bio-solvent met the goal of greater than 80% coal solubility, with 8 bio-solvents being as good as or better than a well-known but expensive hydrogen-donor solvent, tetralin. The Battelle CTL process was then scaled up to 1 ton/day (1TPD) at a pre-pilot facility operated in Morgantown, WV. These tests were conducted, in part, to produce enough material for syncrude-upgrading testing. To convert the Battelle-CTL syncrude into a form suitable as a blending stock for jet turbine fuel, a two-step catalytic upgrading process was developed at laboratory scale and then demonstrated at pre-pilot scale facility in Pittsburg, PA. Several drums of distillate products were produced, which were then distilled into unblended (neat) synthetic jet fuel and diesel products for a detailed characterization. Based on a detailed characterization of the synthetic jet fuel, a 20% synthetic, 80% commercial jet fuel blend was prepared, which met all specifications. An analysis of the synthetic diesel product showed that it has the promise of being a drop-in fuel as super-low (less than 15 ppm)-sulfur diesel fuel. A detailed economic analysis showed that the Battelle liquefaction process is economical at between 1000 metric tons/day (MT/day) and 2000 MT/day. The unit capital cost for Battelle CTL process for making jet fuel is 50K USD/daily bbl compared to 151K USD/daily bbl for indirect CTL, based on 2011 dollars. The jet-fuel selling cost at the refinery, including a 12% capital cost factor (which included profit), for the Battelle CTL process is 61USD/bbl (1.45 USD/gallon). This is competitive with crude oil price of 48 USD/bbl. At the same time, the GHG emissions of 3.56 MT CO 2 /MT fuel were lower than the GHG emissions of 3.79 MT CO 2 /MTfuel for petroleum-based fuels and 7.77 MT CO 2 /MT fuel for indirect CTL. Thus, the use of bio-solvents completely eliminates the need for carbon capture in the case of Battelle CTL process. The superior economics and low GHG emissions for the Battelle CTL process has thus sparked worldwide interest and some potential commercialization opportunities are emerging.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2022

Engineering & Microstructure Scale PIE Report on EBR-II X441A Metallic Fuel Pins for the MORPH Experiment

The objectives of this project are to increase fundamental understanding of irradiation induced metallic U-Pu-Zr fuel behavior and to obtain data needed for the development of irradiation models for metallic fuels in MARMOT. The requested metallic fuel pins are from the X441A experiment irradiated in Experimental Breeder Reactor (EBR)-II and will be provided by the PIs for the duration of the experiment. The purpose of the pins irradiated in the X441A assembly was to vary Zr composition and fuel slug diameter in order to provide data for the metallic fuel performance code “LIFEMETAL” [ANL-IFR-125]. The fuels of interest to this project are metallic fuels with several Zr compositions (in wt.%): U-19Pu-6Zr, U-19Pu-10Zr, and U-19Pu-14Zr, which were irradiated to a peak burnup of approximately 11 at.%. The cladding was the same for all three fuel pins (austenitic stainless steel D9), which allows investigation of fuel-cladding interaction (FCI) phenomena. Varying Zr content in fuel pins enables investigation of the effect of Zr on fuel restructuring and fuel-cladding compatibility. Engineering and microstructure scale PIE activities will be focused on investigation of fundamental aspect of fuel performance such as species diffusion and migration, fission product behavior, and constituent redistribution. Obtained microstructural information will be used as the basis for the development of MARMOT models of U-Pu-Zr fuel performance at the mesoscale. Experimental data obtained through this proposal will be used to provide this fundamental understanding which will serve as the foundation of the development of radiation models for U-Pu-Zr in MARMOT. It will also provide a starting point for the design of new experiments to provide data for the validation of these MARMOT models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

A Revised Capsule Design for the Accelerated Testing of Advanced Reactor Fuels

Testing of fast reactor fuel is currently challenging because of the thermal neutron spectrum of the Advanced Test Reactor (ATR) and the fact that ATR coolant is substantially below typical temperatures found in liquid metal-cooled fast reactors. Despite these issues, since 2003 the Idaho National Laboratory (INL) has successfully tested a variety of metallic, oxide, and nitride fuels for fast reactor applications. There are, however, opportunities to improve the experiments so that the time required to reach a desired burnup can be reduced, and so the sensitivity of the experiments to fabrication tolerances can be minimized. A revised capsule design exploits the use of smaller diameter rodlets to increase power densities and correspondingly reduce the irradiation time required to reach high burnup. Thermal and nuclear analyses indicate that the time to reach 30at% burnup could be reduced from about 12 years to approximately 2-3 years if the fuel diameter is reduced by one-half and to around 1-2 years if the diameter is reduced by one-third. In addition, reducing fuel diameter improves the radial power distribution in the fuel without the use of cadmium shrouding. This observation opens up the possibility of testing advanced reactor fuel systems in Small I locations in the ATR where there is substantial unused capacity. Although not directly tied to a revised capsule design, post-irradiation furnace testing is proposed to provide time-to-failure data that will be helpful in the benchmarking of fuel performance models and in assessing the performance of new fuel designs relative to more standard fast reactor fuel systems. In this paradigm, ATR irradiation experiments will be viewed more as preparation for furnace testing than as experiments that exist in isolation. An added benefit to furnace testing is that uncertainties in temperatures will significantly be reduced relative to non-instrumented capsules in the ATR. The time and cost in developing this experimental capability is expected to be reduced considerably because of the availability of the blister anneal furnace used by the HPRR program. A key issue in the experiment approach is the potential for non-prototypical rodlet dimensions to influence the observed performance of a given fuel design. For example, a qualitative assessment indicates that the effects of fuel-cladding chemical interactions during furnace testing might be more pronounced in reduced diameter rodlets simply because thinner cladding would be more sensitive to a given depth of fission product penetration. In a traditional empirically-driven approach to fuel qualification, any change from prototypical behavior is undesirable because the resulting data cannot easily be assimilated into a statistical correlation. However, in the current approach, systematic variations in fuel diameter provide the opportunity to benchmark fuel performance models over a wider range of parameter space, which in turn is expected to improve understanding and result in better, more physically based models in the long term.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The application of MAMMOTH for a detailed tightly coupled fuel pin simulation with a station blackout

The accurate calculation of desired quantities to predict fuel behavior in a nuclear reactor requires the solution of inter linked equations representing different physics phenomena. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding outer surface. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. In this work the MAMMOTH application is employed in the demonstration of an approach that more consistent with first principles, in a simulation that couples the neutronics application Rattlesnake, the thermal fluids application RELAP-7 and the fuels performance application BISON. A single fuel pin was modeled based on the dimensions of a fuel rod from a Westinghouse 17*17 optimized PWR fuel assembly. The simulation consisted of a depletion period of 1343 days, representing three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted with a total pin power of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shutdown). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a tightly coupled approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain, which includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2-D axisymmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1-D flow equations. For this work, the mesh on the Rattlesnake side is either be 3-D (for low order PN transport or diffusion) or 2-D axisymmetric (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burnup are transferred accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burnup distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh to perform a fuels performance solution. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. For this paper, sets of results from the detailed calculation are provided for both depletion and the SBO event. We demonstrate that a detailed simulation akin to first principles can be achieved using MAMMOTH and the various MOOSE application on differing domains. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Study on Modeling of Nuclear Fuel End Drop during a Hypothetical Accident

It is well known that shell elements have been used to model the cladding of single nuclear fuel pins. These have been used, for example, to analyze the response of a fuel assembly by assuming that all fuel pins in a fuel assembly deform in the same pattern. It is increasingly significant to model the entire fuel assembly in order to capture more realistically the behavior of each individual nuclear fuel during a hypothetical accident in a manner that can account for interactions among single pins. However, modeling each individual fuel pin in an entire fuel assembly using shell elements will result in models with large amount of elements and degrees of freedom that take correspondingly larger amount of computational times to execute. Therefore, modeling the nuclear fuel assembly with beam elements may become an option. This paper presents the results of a set of analysis with the LS-DYNA finite element software that investigated the differences and similarities of modeling nuclear fuel assemblies using either shell elements or beam elements. This study provides additional insights on the appropriateness of using beam element models of the fuel pins in the simulation of a hypothetical accident. In this study, two single pin models, a single pin model using shell elements and a single beam model using beam elements, are constructed to analyze end drop scenarios with the LS-DYNA code. Each model is then analyzed for a 19 ft end drop scenario (linear behavior) and 30 ft end drop scenario (non-linear behavior). Comparisons are made between the results obtained with the single pin shell model and with the single pin beam model to study the differences and similarities resulting from the two modeling approaches. In this study, a B and W 15x15 PWR fuel is analyzed for both 30 ft and 19 ft end hypothetical accident drops using both single pin finite element model with shell elements and beam elements for the fuel cladding. It is found that for the 19 ft end drop, the difference in the Max/Min total axial strains is about 12% between the two models. For the 30 ft end drop, the maximum total axial strains calculated with both models is similar, and a difference of about 12% is observed for the minimum total axial strains calculated by the two models. Advantages of the single pin shell model are that it permits consideration of internal pressures and may capture local buckling. Therefore, use of beam elements to model the fuel cladding needs to take into account the possible significance of local buckling and internal pressurization for the problem being analyzed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2 ) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas, allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration. Complete dissolution of the Al 1100 and Al 6061 T6 alloys up to a final Al concentration of 2 M was obtained using a 7 M HNO 3 solution containing a 0.002 M Hg catalyst. However, following the dissolutions, solids were observed in the solution. The solids were amorphous, but likely originated from the Si present in the alloys. No crystalline materials, such as Al(NO 3 ) 3 were observed. During the course of the dissolution experiments, it was determined that delaying the addition of Hg once the HNO 3 solution reached the boiling point can reduce the total offgas and H 2 generation rates. The delay in starting the Hg addition is not necessary for HFIR fuel dissolution, but could be useful in other research reactor dissolution campaigns. The potential to generate flammable concentrations of H 2 in the offgas during a HFIR fuel dissolution was evaluated using the experimental data. The predicted H 2 concentration in the dissolver offgas stream was compared with 60% of the calculated H 2 LFL at 200 °C using several prototypical experiments. The calculations showed that a full HFIR core can be dissolved using nominally 0.002 M Hg to catalyze the dissolution. The margin between the predicted H 2 concentration and the calculated LFL was greater when the solution was allowed to boil for 45 min prior to initiating the Hg addition. When the Hg was increased to 0.004 M, the predicted H 2 concentration exceeded the calculated LFL early in the dissolution. The dissolution experiments also demonstrated that additional Hg (beyond the initial 0.002 M) could be added as the Al concentration increases. The ability to add more Hg during a HFIR fuel dissolution could be beneficial if slow dissolution rates are observed at high Al concentrations. Experimental data were used to demonstrate that the predicted H 2 concentration in a dissolver was below 60% of the calculated LFL at 200 °C when 0.004 M Hg was used to catalyze the dissolution if the Al concentration is conservatively greater than 0.5 M. Data also show that the Hg concentration during a HFIR fuel dissolution can be increased from 0.002 to 0.008 M at an Al concentration of 1.3 M.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Particle Image Velocimetry Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel assemblies using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. The fuel pins are typically wrapped with a helical wire spacer. The primary function of the wire spacers is to maintain a gap between neighboring fuel pins. They also mitigate vortex-induced vibration and enhance sub-channel mixing to increase convective heat transfer. Many experiments have been performed to study thermal-hydraulic characteristics, primarily pressure drops, of the wire-wrapped fuel bundles. Novendstern developed a semi-empirical model to predict pressure losses in a hexagonal array of wire-wrapped pins in the turbulent flow regime. Rehme proposed pressure drop correlations for a wide range of geometrical parameters, such as pitch-to-diameter ratio P/D from 1.125 to 1.417, lead-to-wire diameter ratio H / dm from 6 to 45, and number of pins n from 7 to 61. Cheng and Todreas introduced their hydrodynamic models for subchannel friction factors and mixing parameters in a 37-pin fuel bundle, which covered the laminar, transition and turbulent flow regimes. Computational fluid dynamics (CFD) offers an opportunity to validate the above correlations and investigate the flow characteristics in the subchannels. Several numerical studies of the wire-wrapped fuel bundles have been performed and most of them are Reynolds-Averaged Navier-Stokes (RANS) based simulations with various turbulence models. Ahmad and Kim performed RANS simulations of 7- and 19-pin bundles using the k-omega SST turbulent model, while Gajapathy et al. performed RANS simulations on 7-pin bundles using the k-epsilons model. Recently, Fischer et al. carried out large-eddy simulation (LES) calculations on wire-wrapped fuel bundles to obtain higher-resolution numerical solutions of subchannel behavior. The geometrical complexity of the hexagonal wire-wrapped fuel bundle has limited experimental activities to pressure drop measurements. A database of pressure drops and flow-field measurements in a 61-pin wire-wrapped hexagonal fuel bundle is not available to benchmark the existing correlations and validate the CFD calculations. Texas A and M University is conducting isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these tests is to perform high spatial and temporal resolution measurements of the flow and pressure measurements at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV) and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and flow-field measurements that will be suitable for validating CFD computer codes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Criticality safety analysis of spent fuel storage or handling disposals using burnup credit - loading curve calculation - 14445

MILLENNIUM and REEL SAS Companies have shared their own experience to develop a neutron and criticality calculations chain in order to perform isotopic composition predictions of spent fuel after irradiation in a reactor and criticality analysis of fuel storages taking into account burnup credit. The whole calculation sequence is divided into three steps, all calling SCALE 6.1.2 modules. The first two steps are dedicated to the evaluation of isotopic compositions of burnt fuel. As transport and depletion calculations in three dimensions appears to be way too long for the whole core (about several months), dealing with each fuel assembly one by one is the alternative solution MILLENNIUM and REEL SAS commonly chose to study. Therefore, the first step aims at determining the axial burnup profile of a given fuel assembly using a 3-D model with infinite reflections on lateral axes. The T6-depl sequence of TRITON module coupling KENO-VI and ORIGEN-S was used in this step. Sensitivity studies confirmed that addition of the 8 closest neighbors of the studied fuel assembly in the model has an insignificant impact on the axial burnup profile. However, fuel assemblies in the peripheral zone of the core must be treated separately with a model including the nearest fuel assemblies and the effect of peripheral water. Then, the second phase evaluates the isotopic compositions of each fuel assemblies at the center of the active height with a 2-D model under conservative core conditions. The T-depl sequence of the TRITON module coupling NEWT to ORIGEN-S is called in this step. Each fuel assembly must be studied separately to consider loading pattern changes occurring at each cycle. An infinite model with reflections on lateral axes is used and the closest neighbors of the fuel assembly are limited to 2 pins - which is around the size of the neutrons mean free path in the core. This is an innovative physical trick aiming at compensating for a limited computing power. Finally, isotopic compositions calculated are combined with the axial burnup profile of the first step to extrapolate the concentrations along the whole active length. Data can then be added in any criticality model, such as a spent fuel pit or any disposal handling fuel assemblies. Criticality analyses of burnt fissile material using burnup credit can be managed within a reasonable amount of time, on condition that sufficient proofs of code qualification for the studied range of parameters are provided, especially on fission products isotopic compositions calculation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High-speed phase contrast imaging of spray breakup of jet fuels under combusting conditions

X-ray phase contrast imaging is conducted at 90,517 Hz on a fuel spray to characterize the breakup of the bulk fuel into ligaments and individual droplets in a single-sector, swirl-stabilized combustor under reacting conditions. Phase contrast imaging is used to qualitatively assess spray breakup for two fuels: Jet-A, which represents a normal viscosity fuel, and a blend of JP-5 (64%) and farnesane (36% by volume), which represents a high-viscosity fuel, at three fuel-flow rates and two air inlet temperatures. Time-series images of the fuels show that atomization occurs much more rapidly at higher fuel pressure with thinner ligaments being rapidly shredded. The high-viscosity fuel is qualitatively observed to have markedly longer, thicker ligaments and larger droplet diameters after the primary breakup than the normal viscosity fuel. Using an image processing and droplet quantification strategy, the mean droplet velocity magnitudes were found to decrease with decreasing fuel pressure and fuel-flow rate. The increase in preheat temperature from 323 to 370 K results in an increase in the mean droplet velocity magnitude and a decrease in numerical average and Sauter mean diameter, indicating improved atomization. The high-viscosity fuel was found to have larger average droplet diameters at each condition than the normal viscosity fuel. The increased preheat is found to enhance the combustion efficiency, which is hypothesized to couple with the improved atomization due to the higher combustion temperatures. This indicates the importance of examining spray atomization and breakup under realistic combusting conditions for relevant fuels as the heat release directly couples with the atomization process.

42 ENGINEERING↗