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DOE OSTI2017

Testing Fast Reactor Fuels in a Thermal Reactor: A Comparison Report

The testing of fast reactor fuels in the thermal-spectrum of the Advanced Test Reactor using cadmium-filtering has been underway since 2003. The objective of this experimental approach has not been to create testing conditions that are fully prototypic of a fast reactor, but to create a thermal profile inside of test fuel rodlets that is nearly prototypic of corresponding fast reactor conditions. By doing so, this experimental approach should allow for the study of fast reactor fuel performance phenomena that are primarily dependent upon the conditions of temperature and/or temperature gradient inside the fuel to be possible using a thermal test reactor. Validation of this assertion has been undertaken by comparing fuel performance phenomena observed in fuels irradiated in cadmium-filtered positions in ATR to similar or identical fuels irradiated under similar conditions in genuine fast reactors. This report documents a preliminary comparison of limited fuel performance data available from the testing of metallic, oxide, and nitride fuels in cadmium-filtered positions in the ATR to the performance of identical or similar fuels irradiated in genuine fast reactors. Although a variety of fuel performance metrics are presented, particular attention is paid to those phenomena that are primarily, or significantly, dependent on conditions of temperature and temperature gradient with the fuel. Although the comparisons presented in this report are at this time limited in scope and preliminary in nature, they do support the assertion that the objectives of the cadmium-filtered testing approach in ATR are both feasible and appear sound. The available evidence supports the assertion that use of cadmium-filtering creates irradiation conditions in the ATR that are both relevant to, and suitable for, fast reactor fuel performance testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

HALEU Decontamination Investigations for EBR-II Recovered Uranium. HALEU Drip Casting Results in the Fuel Conditioning Facility Cathode Processor

Many of the new advanced reactor designs being proposed by the nuclear private sector require uranium fuel enriched between 5 and 20%, otherwise known as High Assay Low Enriched Uranium (HALEU), to support their designs. There are currently no commercial facilities based in the United States which have the capability to produce this material, and as a result the nuclear industry has requested support from the Department of Energy (DOE) to meet this need. One approach being considered by DOE to respond to this demand is to recover highly enriched uranium from used nuclear fuels and downblend it to enrichments less than 20 % U-235. DOE’s past research related to liquid metal fast breeder reactors resulted in the irradiation of more than 3 metric tons (MT) of nuclear fuel composed of highly enriched metallic uranium which resides at the Idaho National Laboratory (INL). This inventory of used nuclear fuel is currently being conditioned using an electrometallurgical treatment process (EMT) to facilitate its eventual disposition to a geologic repository. As part of this treatment process, the highly enriched component of this used fuel is separated, recovered, downblended to <20% U-235, and stored as a metallic uranium ingot. The majority of this fuel was irradiated in the Experimental Breeder Reactor II (EBR-II) over the course of its 30 years of operation. This sodium cooled fast reactor utilized a core composed of highly enriched, sodium-bonded, metallic fuel to sustain or “drive” the reactor. This EBR-II driver fuel is currently being treated for sodium neutralization in the Fuel Conditioning Facility (FCF) at the Materials and Fuels Complex (MFC) and the metallic uranium recovered as a component of this process has been proposed as a potential source of supply of high assay low enriched uranium. The irradiation of the metallic uranium fuel results in the creation of fission products and some minor actinides which are not compatible with many reuse scenario’s, and introduces significant challenges associated with fabrication of new fuel elements using this material. However, it is believed that the current EMT process can be slightly enhanced to address these concerns and result in the recovery of a product that can be considered as a source of supply for HALEU applications. Based on this, DOE requested the INL research methods to enhance the EMT process to produce a recovered uranium material that has a lower contaminant level and lower radiological dose rate and could be utilized in glovebox based fuel fabrication scenarios. The report that follows documents the progress toward meeting this goal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Phase Equilibria and Thermochemistry of Advanced Fuels: Modeling Burnup Behavior (Final Report)

Achieving the goal of developing advanced fuel concepts that meet the DOE objectives of being robust, demonstrating high performance, and are more tolerant of accident conditions than current fuel systems will require a thorough understanding of the thermophysical and thermochemical properties of the constituent materials. Non-oxide fuel systems are being explored under the Advanced Fuels Program that hold significant promise for improved performance and accident tolerance, including the uranium silicide-based system considered in the current work. Prospective cladding materials currently considered that contribute to improved accident tolerance include silicon carbide composites and ferritic alloys (Fe-Cr-Al base compositions). Thus, the effort developed thermochemical models and values, supported with targeted experiments, to evaluate the ferritic alloy and silicon carbide composite cladding systems in contrast to current zirconium alloy cladding. The developed detailed understanding will serve to aid in relatively early screening of candidate systems to avoid wasted effort, guide development of new fuel forms, and to provide a basis for predicting and modeling fuel performance. Major deliverables for the project included: Thermochemical assessment and models of phases in the U-Si and U-Si-N systems; thermochemical evaluation supported by experimental measurements of fuel-cladding interactions of silicide fuel with baseline zirconium, ceramic composite, and ferritic alloy cladding; thermochemical assessment and models of phases supported by experimental measurements for silicide fuel with key fission products provided in a dataset and reported in refereed publications. Within the project a significantly refined U-Si phase diagram was developed and reported that now includes homogeneity ranges for key phases, such at the U 3 Si 2 proposed fuel phase, and settles issues with regard to uncertainty in the stability of some phases. Computational efforts together with key experiments has determined phase formation in interactions between U 3 Si 2 and Zircaloy-4 cladding material, a ferritic FeCrAlY alloy of interest as an advanced cladding material, and silicon carbide, also of interest as a fiber-reinforced composite cladding. As expected, very significant reactions occur between U3Si2 and Zircaloy-4, with much less interaction at higher temperatures for the ferritic alloy, and finally interactions with SiC only in the region of contact. A potentially major issue is the stability of U 3 Si 2 fuel that has undergone significant burnup. The result is the loss of the uranium metal, liberating silicon, and the formation of concomitant fission product elements that either dissolve in the U 3 Si 2 phase or form independent, and possibly silicide phases. A combination of experimental determinations of phase formation of U 3 Si 2 reacted with representative fission products yttrium, gadolinium, cerium, zirconium, and molybdenum and first principles calculations has helped understand the fuel chemistry. The behavior of the U 3 Si 2 phase and the partitioning of silicon to possible fission product phases with burnup was thus determined, with significant dissolution in U 3 Si 2 of cerium, gadolinium, zirconium, and plutonium predicted along with independent phase formation of a U-Mo-Si ternary phase, yttrium silicide, and elemental selenium. It can be concluded that at significant burnup there will be a very minor amount of the U 3 Si 2 fuel phase that will decompose to a lower silicide or a uranium alloy as silicon preferentially forms a secondary phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Alternative Fabrication of Volatile Metal Fuel Slugs for Sodium-cooled Fast Reactor

In order to develop alternative fabrication method of fuel slugs for preventing the evaporation of volatile elements such as AM,U-10wt.%Zr-Mn fuel slugs for SFR (: Sodium-cooled Fast Reactor) have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication. Metal fuels such as U-Zr/U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's In order to develop innovative fabrication method of metal fuel slugs for preventing the evaporation of volatile elements such as Am, modified casting under inert atmosphere has been applied for metal fuel slugs for SFR. Alternative fabrication method of fuel slugs has been introduced to develop an improved fabrication process of metal fuel for preventing the evaporation of volatile elements. In this study, U- 10wt.%Zr-Mn fuel slugs for SFR have been fabricated by modified casting method and characterized to evaluate the feasibility of the alternative fabrication method. Alternative casting such as modified casting has been applied to develop fabrication method of fuel slugs for preventing the evaporation of volatile elements such as Am. U-10wt.%Zr-Mn containing a volatile surrogate Mn fuel slugs was soundly fabricated under inert atmosphere with dimensions of Φ5 - L 250 mm. Mass fraction of fuel loss was so low, up to 0.2%. Mn element was most recovered with prevention in evaporation of Mn. It was seen that the losses of volatile Am can be effectively controlled to below detectable levels using modest pressure. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Microanalysis of Irradiated Metallic Hydride Nuclear Fuel (U{sub 0.17}ZrH{sub 1.6})

Metallic hydride nuclear fuels (U{sub 0.17} ZrH{sub 1.6}) have been under consideration for use in Light Water Reactors (LWRs). The fuel matrix consists of isolated particles of uranium metal within a zirconium hydride matrix. Erbium was also added as a neutron poison enabling higher reactor performance. These fuels were encased in a Zircaloy-2 cladding with a Bi/Pb eutectic liquid metal filled-gap to improve thermal conductivity. The inner and outer cladding surfaces were oxidized to nominal 1 μm thickness to eliminate hydrogen attack. The fuels were termed UCB to note the location of fabrication as the University of California-Berkeley. Irradiations were performed in the MITR research reactor. The details are described elsewhere. We originally reported on post-irradiation analysis of the fuel elements of two of the rodlets using Scanning Electron Microscopy (SEM) and x-ray energy dispersive spectroscopy (EDS). In this further study, funded as a rapid turnaround experiment (RTE) through the Nuclear Science User Facilities (NSUF) program, we performed high-resolution Scanning Transmission Electron Microscopy (STEM) of thin lift-out specimens prepared from specific regions in the fuel matrix using an SEM-Focused Ion Beam (FIB) instrument, and again compared these results to the unirradiated material. Earlier investigations using SEM-EDS showed that ∼8% of the Zircaloy cladding had dissolved in the Pb-Bi eutectic during irradiation over 1000 hrs. However, the SEM-EDS failed to show any detectable evidence of damage of the fuel matrix. There was no obvious swelling or migration of elements; hence, we focused on these specific regions in greater detail. Much of the data on hydride fuels come from the Space Nuclear Auxiliary Power (SNAP) program from the 1960's. The principle difference between oxide and hydride fuel is the high thermal conductivity of the latter. This feature greatly decreases the temperature drop over the fuel during operation, thereby reducing the release of fission gases to the fraction due only to recoil. However, very unusual early swelling due to void formation around the uranium metal particles has been reported in hydride fuels over 700 deg. C. In this investigation, the fuel did not reach these extreme temperatures. Phases formed in this system have been previously experimentally investigated using thermal analysis techniques, and wavelength and energy dispersive spectroscopy (EDS/WDS) in scanning electron microscopes (SEM). Major changes were observed in the liquid metal and cladding components but at the scale of the SEM observations, it was not readily apparent that any major changes had occurred to the fuel matrix (i.e. U metal particles in the zirconium hydride matrix with interdispersed erbium metal). Rapid oxidation of the SEM mechanically polished cross-sections, may have obscured our ability to see major changes. By using the SEM-FIB slices, it was possible to remove oxide coatings and the smoother surfaces were found to be less susceptible to oxidation once removed from the vacuum chamber. The high spatial resolution of STEM allowed characterization of individual matrix phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluating the Impact of Tritium Permeation Membrane Performance and Direct Internal Recycling on Fusion Fuel Cycle Efficiency Using TMAP8

An efficient fuel cycle is vital to sustainable and cost-effective energy generation in fusion systems. Since tritium is not widely available, fusion systems must breed their own tritium for sustainable fusion deuterium-tritium reactions. An inefficient fuel cycle increases the tritium inventory needed for operations, which increases costs, constraints on tritium management systems, and safety concerns. A fuel cycle model is a powerful tool for understanding tritium inventories and flow rates across all systems in the fuel cycle. By simplifying the technical details into time-dependent tritium flow rates and inventories, the model can simulate the entire fuel cycle with high computational efficiency, even for technologies that are still under development. It can therefore quantify the impact of new tritium management technologies on fuel cycle efficiency. To evaluate the impact of key components on reducing tritium inventory, we are using and expanding an existing fuel cycle models based on latest advancements in fuel cycle research. The new model integrates Tritium Permeation Membrane (TPM) and Direct Internal Recycling (DIR) to enhance tritium transport from blanket breeders and plasma exhaust. These fuel cycle models are implemented in TMAP8 (Tritium Migration Analysis Program, version 8), a MOOSE-based open-source application designed to provide cutting-edge capabilities for tritium transport and fuel cycle modeling. The study aims to demonstrate the extensibility of existing fuel cycle modeling capability in TMAP8 and to offer a proof-of-principle design for future fusion plant systems. The presentation will cover the performance of fuel cycle modeling capabilities available in TMAP8, highlight advancements in fuel cycle research, and present a sensitivity analysis of these models. The results underline potential approaches and technology solutions to lower tritium inventory requirements, highlighting their role in shaping the future of fusion energy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

FY2016 Ceramic Fuels Development Annual Highlights

Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Encapsulation of Defective Fuel Rods for Transport and Dry Storage - 18190

Removing of the whole inventory of fissile and non-fissile material out of the spent fuel pool is performed when the power plant is about to start its last operation phases or may occur during the extended operating life cycle due to plant storage space limitation. Removal of defective spent fuel is then definitely required for emptying the pool and may in addition also be performed in the course of plant operation for reducing pool contamination. Depending on the general used fuel management solution chosen for post-storage pool operations, specific management of a defective fuel assembly (DFA) and/or a defective fuel rod (DFR) should be implemented accordingly to the regulation requirements to be followed. For meeting these DFA and/or DFR management needs, Framatome together with New AREVA offer wide portfolio of solutions with flexibility and adaptability advantages: - from simple to advanced technologies; - for several types of fuel assemblies and/or defective rods; - for suitable and compatible cask designs; - for compatible solution with both used fuel final management strategies (closed and open fuel cycles). For instance, a technology is jointly proposed and developed by New AREVA and Framatome for implementing dry storage of DFR. It consists in: - encapsulating each DFR by loading the fuel rod in a respective capsule, drying the fuel rod using a process which allows to physically verify with high accuracy the dryness inside the DFR, and seal welding the capsule - assembling the capsules in a capsule canister (a skeleton with external dimensions similar to a respective fuel assembly) and - transferring the capsule canister along with intact spent fuel assemblies in an associated dry transport and storage (dual purpose) cask. Such a project is currently being implemented in Belgium. After successfully obtaining the cask transport and storage license extensions for including the encapsulated DFR content, the first encapsulation campaign has been started in 2017. The purpose of the paper is to present the industrial solutions proposed by Framatome and New AREVA together, for management of DFR, notably the one related to dry storage of encapsulated DFR implemented in Belgium and the perspectives opened by this successful project. The paper presents: - the available solutions for management of DFA and/or DFR depending on applicable requirements - the drying method, welding of a capsule and associated cask as the state of art technology for dry storage of DFR and/or DFA; - lessons learned from project management and encapsulation campaign performed in Belgium; - perspectives for management of DFR and/or DFA. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparison of Subdivided Versus Full-Length Fuel Pin Burnup in the ATF-2 Safety Analysis

The Accident Tolerant Fuels (ATF) campaign, under the Fuel Cycle Research and Development (FCRD) Advanced Fuels program, was developed after the Fukushima Daiichi Nuclear Power Plant disaster in 2011. Under ATF, fuel and cladding concepts are to be developed and tested that may more readily withstand an accident similar to the one that occurred at Fukushima Daiichi. ATF is a combination of industry, universities, and national laboratories in an effort to develop advanced accident tolerant cladding and fuel concepts. The end goal of the ATF program is to insert a lead test rod (LTR) or lead test assembly (LTA) into an operating commercial power light water reactor by 2022. In order to meet this 2022 insertion goal, ATF is split into three separate phases: ATF-1, ATF-2, and ATF-3. The first phase of the ATF program, ATF-1, is the concept feasibility study and early irradiation testing. Feasibility, fabrication, and out-of-pile testing are happening at universities and laboratories all over the United States. Irradiation testing of these concepts is happening at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Early irradiation testing involves testing clad and/or fuel in drop-in capsules in the ATR. Currently, these capsules are in the relatively low-flux positions in the ATR, but some newer concepts and concepts that require higher burnup will be moved to the higher-flux positions later on during irradiation. Currently, Oak Ridge National Laboratory (ORNL), Westinghouse Electric Company, Areva, and General Electric (GE) have inserted capsules for ATF-1, although more concepts are planned. The second phase of ATF, ATF-2, is the focus of this paper. ATF-2 is the loop testing of fuel pins with new cladding and fuel concepts to simulate pressurized water reactor (PWR) and boiling water reactor (BWR) environments. ATF-2 will concurrently be happening at the ATR and at the Halden Reactor in Norway. The ATF-2 irradiation at the ATR will only simulate a PWR environment. ATF-2 will be inserted into the ATR pressurized water loop, Loop 2A, in the Center Flux Trap (CFT), where chemistry, temperature, and pressure can all be controlled separately from the primary coolant system and other loops in ATR. Loop 2A will be borated, pressurized to 2200 psi nominal, and have an average water temperature of 300 deg. C. A borated loop with fueled rodlets is extremely rare for Nuclear Energy experiments in the ATR, so ATF-2 will demonstrate new and unique capabilities for INL. ATF-3 is the testing of fresh and irradiated ATF concepts at the Transient Reactor Test (TREAT) Facility. The TREAT Reactor is planned for restart, re-establishing U.S. transient testing capabilities after nearly 20 years of non-operations. TREAT will test ATF concepts under transient conditions, such as a reactivity insertion accident. In addition to fresh fuel pins, TREAT will use from ATF2 and do transient testing on those irradiated pins. The information TREAT will provide to the vendors to support fuel licensing will be valuable, showing that these concepts do enhance the accident tolerance of the fuel in commercial reactors. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fuel property impacts on gaseous and PM emissions from a multi-mode single-cylinder engine

The U.S. Department of Energy’s Co-Optima initiative has focused on improving fuel economy and vehicle performance while reducing emissions through the simultaneous development of emerging sustainable fuels with beneficial properties and advanced combustion strategies. A major thrust has been the development of advanced compression ignition (ACI) combustion strategies of gasoline range fuels in combination with spark-ignited (SI) combustion in a single engine capable of multi-mode operation to achieve high power density with enhanced part load efficiency. The aim of this study was to further the understanding of how emissions from both ACI and SI strategies operating on the same fuels in the same engine are impacted by different fuel properties. This investigation focused on particulate matter (PM) and gaseous hydrocarbon emissions from 6 different fuels across 3 different combustion modes on the same single-cylinder engine designed for multi-mode operation: SI combustion, partial fuel stratification (PFS), and spark-assisted compression ignition (SACI). In each of these modes, 3 different CA50 phasings were studied such that all 6 fuels could be studied at the same phasings. Three of the six different fuels used were specially formulated in a previous investigation to study the impact of fuel distillation and aromatic content while maintaining the research octane number (RON) and octane sensitivity. Additionally, neat isooctane and two ethanol containing fuels (RD5-87 and Co-Optima E30) were studied. Different fuel and phasing impacts on emissions were observed across the three combustion modes. Fuel properties were found to impact soot PM and particle number more than the CA50 phasing, while the phasing had more impact on NOx emissions. The NOx emissions were reduced in the PFS mode for all fuels compared to SI combustion, but the SACI combustion mode did not reduce NOx emissions. Although PFS produced low soot PM emissions like SI, total PM mass emissions were significantly higher due to large organic carbon (OC) PM mass contribution. Both PFS and SACI had greater particle number emission than SI operation with small nuclei mode particles dominating in PFS compared to large agglomeration particles in SACI.

42 ENGINEERING↗
DOE OSTI2013

Fuel Assembly Shaker Test Simulation

This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.

Klymyshyn, Nicholas A.↗
DOE OSTI2020

Augmented Monitoring and Condition Assessment Program (AMCAP) Material Test Reactor (MTR) (Fuel Inspection Program Report)

The AMCAP MTR Fuel Inspection Program, a special inspection program comprised of four inspection campaigns, examined a total of ten (10) pre-selected aluminum-clad, aluminum-based fuel core spent nuclear fuel assemblies (ASNF) stored in the L Area Disassembly Basin (L Basin) at the Savannah River Site. A full description of the inspections and the results are reported. The fuel had been stored in bundled-tube storage (Vertical Tube Storage) for periods of 18 to 21 years. The prior service experience of the individual 10 assemblies varies, but all included irradiation followed by wet storage at international research reactor sites prior to shipment to the US and storage in L Basin. The 10 assemblies were expected to be among the “worst"’ in terms of prior corrosion damage of the entire inventory of the direct-bundle-stored ASNF in L Basin. The inventory of MTR ASNF in L Basin will continue to be stored in the bundled-tube configuration or in slug-storage buckets with inserts pending retrieval for ultimate disposition. The MTR fuel inspections focused on collecting information for characterization of the material condition of the ASNF considering various types of aluminum fuel corrosion degradation of its assembled materials in water storage. A custom-designed Fuel Inspection Table was used to stage the fuel for remote, enhanced visual examination (close-up video imaging & recording) with controlled lighting and positioning that enables reproducibility of imaging conditions. The inspections were conducted by fuel subject matter expert staff from Spent Fuel Project Engineering (SFPE) and the Savannah River National Laboratory (SRNL). Stills captured from the video records were used to compare the corrosion evolution from previous records, as available. This evaluation of the inspection results including the comparison to the previous inspection results demonstrate that the water quality and the storage configuration of ASNF in L Basin do not cause aggressive corrosion degradation of the fuel; mitigation of the prior corrosion degradation of the fuel also appears to have been achieved with the good water quality conditions of L Basin. Recommendations are made for future inspection of the fuel to trend corrosion degradation and demonstrate continued safe wet storage of the ASNF in L Basin. The next fuel examination is recommended to be performed in 5 years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Research and Test Reactor Fuels

PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Capturing the Impact of Fuel Octane Number on Modern Gasoline Vehicles with Octane Indices

The need for high efficiency automotive engines has led to more complex air handling and fuel injection systems, higher compression ratios, more advanced combustion and aftertreatment systems, and the use of fuels with higher octane ratings. Higher octane number fuels have a lower propensity to knock. This work studies the influence of changing fuel octane rating on two modern production gasoline vehicles, one with a naturally aspirated, port injected engine and the other with a turbocharged, direct injected engine, using fuels with four different octane number grades (with 85, 87, 91, and 93 anti-knock indices) and operated over a variety of driving cycles and temperature conditions. Unlike previous studies, this effort develops and demonstrates a methodology that isolates fuel effects on fuel consumption and provides a clear view of the octane impact on existing vehicles. While fuel octane rating can also impact factors such as the allowable compression ratio and gear shifting strategies, this study examines fuel consumption changes that are solely attributable to octane rating on production vehicles. The developed approach uses results from drive cycle as well as steady-state tests along with response surface modeling to predict the engine combustion phasing and changes in efficiency for each fuel relative to the baseline 87 anti-knock index (AKI) fuel. The methodology is then used to capture the role of fuel octane rating on powertrain efficiency over Environmental Protection Agency (EPA) certification drive cycles and demonstrate the variation of that impact as a function of powertrain output power. Based on the vehicles and fuels tested in this study, the metrics indicate that the effects of octane on efficiency are limited to a portion of the operating modes represented in the driving cycles used for vehicle certification. In addition, an octane potential benefit index (OPBI) suggests that the vehicle with the turbocharged, direct injected engine has a greater potential to benefit from higher octane number fuel compared to the naturally aspirated, port injected engine and proposes a way to quantify this impact for production vehicles.

Jehlik, Forrest↗
DOE OSTI2021

Experimental Fuel Consumption Results from a Heterogeneous Four-Truck Platoon

Platooning has the potential to reduce greenhouse gas missions of heavy-duty vehicles. Prior platooning studies have chiefly focused on the fuel economy characteristics and three-truck platoons, and most have investigated aerodynamically homogeneous platoons with trucks of the same trim. For real world application and accurate return on investment for potential adopters, non-uniform platoons and the impacts of grade and disturbances on a platoon’s fuel economy must also be characterized. This study investigates the fuel economy of a heterogeneous four-truck platoon on a closed test track. Tests were run for one hour at a speed of 45 mph. The trucks used for this study are two 2015 Peterbilt 579’s with a Cummins ISX15 and a Paccar MX-13, and two 2009 Freightliner M915A5’s, one armored and the other unarmored. Many analysis methodologies were leveraged to describe and compare the fuel data, including lap-wise and track-segment analysis. The methodology for dividing the data into laps is described in detail. The influence of other factors beyond the aerodynamics of platooning is discussed. CAN fuel rate analysis showed excellent agreement with previous experimental trends for two and three-truck platoons. In general, the indicated fuel economy benefits in this study were 5-11% for following vehicles and 0-4% for the lead vehicle in platoon relative to their baseline fuel consumption. On a cumulative basis, all platoons saved fuel, ranging from 6% to 8% versus the sum of the standalone trucks’ fuel consumption. The practical implications of the fuel economy results are discussed, as well as avenues for future research. Introduction and Motivation platooning is controlled coordination of two or more vehicles in a convoy, sometimes called CACC (Coordinated or Cooperative Adaptive Cruise Control). The distance between vehicles in a platoon can be controlled tightly at no additional fatigue to the driver. Platooning vehicles can also respond quickly to braking events of the leader, much more quickly than the typical human driver’s reaction time of 1-1.5 s. Therefore, vehicles in a platoon can follow each other much more closely than would usually be considered a safe following distance under human operation. It is implied that under very close following conditions, platooning technology must be extremely robust before it is safe for wide-scale implementation. Platooning is under investigation as a fuel-saving technology. Close-following significantly reduces aerodynamic drag for both leading and trailing vehicles. According to the NRC in 2010, aerodynamic drag represents roughly half of a Class-8 truck’s on-highway fuel usage, meaning a 20% reduction in drag roughly equals a 10% reduction in fuel usage, if all other sources of energy loss remain equal (i.e. accessory, rolling resistance, drivetrain, braking). It is by aerodynamics that platooning saves fuel. At risk of oversimplifying the aerodynamics, following (or trailing) vehicles experience reduced wind velocity due to shielding, and the leading vehicles experience an increased aft pressure, especially at distances closer than 75’ (23 m).

greenhouse gas emissions, truck platoon, heavy-dut↗
NASA NTRS2013

Alternate-Fueled Combustor-Sector Emissions

In order to meet rapidly growing demand for fuel, as well as address environmental concerns, the aviation industry has been testing alternate fuels for performance and technical usability in commercial and military aircraft. In order to make alternate fuels (and blends) a viable option for aviation, the fuel must be able to perform at a similar or higher level than traditional petroleum fuel. They also attempt to curb harmful emissions, and therefore a truly effective alternate fuel would emit at or under the level of currently used fuel. This report analyzes data from gaseous and particulate emissions of an aircraft combustor sector. The data were evaluated at various inlet conditions, including variation in pressure and temperature, fuel-to-air ratios, and percent composition of alternate fuel. Traditional JP-8+100 data were taken as a baseline, and blends of JP-8+100 with synthetic-paraffinic-kerosene (SPK) fuel (Fischer-Tropsch (FT)) were used for comparison. Gaseous and particulate emissions, as well as flame luminosity, were assessed for differences between FT composition of 0, 50, and 100 percent. The data show that SPK fuel (an FT-derived fuel) had slightly lower harmful gaseous emissions, and smoke number information corroborated the hypothesis that SPK-FT fuels are cleaner burning fuels.

Saxena, Nikita T.↗
NASA NTRS2014

Hot Hydrogen Testing of Tungsten-Uranium Dioxide (W-UO2) CERMET Fuel Materials for Nuclear Thermal Propulsion

CERMET fuel materials are being developed at the NASA Marshall Space Flight Center for a Nuclear Cryogenic Propulsion Stage. Recent work has resulted in the development and demonstration of a Compact Fuel Element Environmental Test (CFEET) System that is capable of subjecting depleted uranium fuel material samples to hot hydrogen. A critical obstacle to the development of an NCPS engine is the high-cost and safety concerns associated with developmental testing in nuclear environments. The purpose of this testing capability is to enable low-cost screening of candidate materials, fabrication processes, and further validation of concepts. The CERMET samples consist of depleted uranium dioxide (UO2) fuel particles in a tungsten metal matrix, which has been demonstrated on previous programs to provide improved performance and retention of fission products1. Numerous past programs have utilized hot hydrogen furnace testing to develop and evaluate fuel materials. The testing provides a reasonable simulation of temperature and thermal stress effects in a flowing hydrogen environment. Though no information is gained about radiation damage, the furnace testing is extremely valuable for development and verification of fuel element materials and processes. The current work includes testing of subscale W-UO2 slugs to evaluate fuel loss and stability. The materials are then fabricated into samples with seven cooling channels to test a more representative section of a fuel element. Several iterations of testing are being performed to evaluate fuel mass loss impacts from density, microstructure, fuel particle size and shape, chemistry, claddings, particle coatings, and stabilizers. The fuel materials and forms being evaluated on this effort have all been demonstrated to control fuel migration and loss. The objective is to verify performance improvements of the various materials and process options prior to expensive full scale fabrication and testing. Post test analysis will include weight percent fuel loss, microscopy, dimensional tolerance, and fuel stability.

Hickman, Robert↗