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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Impact of a Diesel High Pressure Common Rail Fuel System and Onboard Vehicle Storage on B20 Biodiesel Blend Stability

Adoption of high-pressure common-rail (HPCR) fuel systems, which subject diesel fuels to higher temperatures and pressures, has brought into question the efficacy of ASTM International specifications for biodiesel and biodiesel blend oxidation stability, as well as the lack of any stability parameter for diesel fuel. A controlled experiment was developed to investigate the impact of a light-duty diesel HPCR fuel system on the stability of 20% biodiesel (B20) blends under conditions of intermittent use and long-term storage in a relatively hot and dry climate. B20 samples with Rancimat induction periods (IPs) near the current 6.0-hour minimum specification (6.5 hr) and roughly double the ASTM specification (13.5 hr) were prepared from a conventional diesel and a highly unsaturated biodiesel. Four 2011 model year Volkswagen Passats equipped with HPCR fuel injection systems were utilized: one on B0, two on B20-6.5 hr, and one on B20-13.5 hr. Each vehicle was operated over a one-hour drive cycle in a hot running loss test cell to initially stress the fuel. The cars were then kept at Volkswagen's Arizona Proving Ground for two (35 degrees C average daily maximum) to six months (26 degrees C average daily maximum). The fuel was then stressed again by running a portion of the one-hour dynamometer drive cycle (limited by the amount of fuel in the tank). Fuel rail and fuel tank samples were analyzed for IP, acid number, peroxide content, polymer content, and ester profile. The HPCR fuel pumps were removed, dismantled, and inspected for deposits or abnormal wear. Analysis of fuels collected during initial dynamometer tests showed no impact of exposure to HPCR conditions. Long-term storage with intermittent use showed that IP remained above 3 hours, acid number below 0.3 mg KOH/g, peroxides low, no change in ester profile, and no production of polymers. Final dynamometer tests produced only small changes in fuel properties. Inspection of the HPCR fuel pumps revealed no deposits or abnormal wear for any fuel. The results provide some confidence that the ASTM D7467 stability requirement of 6 hr. minimum IP for B6 to B20 blends provides adequate protection for modern engine fuel systems.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2019

Characteristics of TRU Fueled SFR Core during the Transition with Partially Loaded LEU Assemblies

The fast reactor (FR) fuels typically require higher fissile content than the thermal reactor fuels because the cross sections of the fissile isotopes are smaller relative to the fertile isotopes and higher neutron leakage from the reactor core. The medium-sized high-burnup sodium-cooled fast reactors (SFRs) in the framework of uranium-plutonium/transuranic (U/Pu or U/TRU) fuel-cycle can be optimized to operate with less than 17 % of Pu/TRU content in the driver fuel while having higher TRU conversion ratios (CR~1.0) [1]. However, possible shortage or inaccessibility to TRU based fuel could be accommodated by the use of uranium-based fuels on demand such as low enriched uranium (LEU) having significantly higher 235U content than used in the commercial light water reactors (LWRs), which is defined as high-assay LEU. The reactor systems having low TRU conversion ratio are commonly categorized as burner reactors, which include LWRs and even can be achieved by FRs with designed to utilize higher TRU fractions (e.g., increased neutron leakage and lower fuel volume fractions). Such reactor systems need to be initially loaded with high excess reactivity, which decreases toward the end of cycle, in order to achieve target burnup conditions. Therefore, in order to offset the excess reactivity, neutron absorbing materials need to be inserted into the core to compensate (e.g., control rods). By reducing the neutron leakage (e.g., increasing the fuel volume fraction or thicker blankets), the TRU conversion ratio in the FR can be increased. If the TRU conversion ratio becomes 1.0 (technically slightly greater than 1.0 to account for process losses during recycle), the reactor is in the condition known as “breakeven”, in which fuel contains the same amount of fissile material after recycling that was used to fuel the reactor. In such a condition, combined with the fact that fission products do not have strong poisoning effects in FRs, the reactivity swing becomes smaller than that of burner type reactors. For breeders, the reactivity can increase toward the end of the cycle because more fissile material is produced than consumed. In this regard, the breakeven FR core is implicitly optimized in a sense by eliminating the large insertion of control elements to offset the reactivity during the normal operation. However, the breakeven condition here is assumed to be achieved only by the TRU based fuel. Possible lack of TRU based fuel needs to be compensated by available LEU resources containing equivalent amount of fissile (i.e., 235U). The use of LEU will have some impact on important safety parameters that need to be evaluated. U-235 has a higher effective delayed neutron fraction than Pu-239. Also, the number of neutrons produced by fissions per neutron absorption (i.e. ?) of 235U in the fast energy range (>1e4 eV) is lower than that of 239Pu, which may lead to the poor fissile conversion when LEU is used as a high-conversion FR fuel. The full-paper presents the study of the impact of partially loaded LEU fuels within TRU fueled breakeven SFR on the transition of nuclide evolutions and core characteristics particularly focusing on the burnup reactivity swing. The full paper analyzes how the large reactivity swing occurs during the LEU loading and addresses possible options to reduce it to the level equivalent to that of the TRU fueled SFR core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and {sup 238}U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with {sup 238}U alone should be reprocessed in the usual way.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Overview of the U.S. DOE advanced fuels campaign

Advanced Fuel Campaign (AFC) has responsibility to develop advanced nuclear fuel technologies for the Fuel Cycle Research and Development (FCRD) program. AFC focus areas include 4 topics: advanced LWR fuels, actinide transmutation, research infrastructure development for the study of new nuclear fuels and advanced modeling and simulation. This paper reviews the latest technical advances in the 4 topics. Fuel and cladding materials under investigation include the following general classes of technologies: iron-based cladding, ceramic cladding, coated Zirconium alloy cladding, molybdenum-based cladding, high-density fuel, fully ceramic matrix fuel, and metallic alloy-based nuclear fuel. Concerning actinide transmutation fuel alloy additions to immobilize the lanthanide fission products, sodium-free annular fuel concepts, and cladding coatings and liners for mitigation of fuel-cladding chemical interaction have been investigated. Direct casting of metallic fuel into a zirconium liner was demonstrated. A Talos Transmission Electron Microscope was acquired for the study of advanced nuclear fuels and materials. The conceptual and preliminary designs for a static capsule, a flowing water loop, and a flowing sodium loop to support future transient testing needs have been completed. AFC interfaces with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop multi-scale, multi-physics fuel performance codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

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

Assessing High Burnup U-19Pu-10Zr Fuel Performance against Historical and Modeled Behavior

Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Assess How Changes in Fuel Cycle Operation Impact Safeguards

Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuel therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or impossible to access location. The primary driver for suggesting this shift in approach is the change in robotic technology and information technology in general. It should be possible, with minimal impact to the facility, to measure each assembly every time that it is moved in the pool, with the first measurements being made at discharge. The following conclusions were reached: The total neutron count rate can be accurately predicted at any future moment in time based upon the measured count rate at discharge, provided the initial enrichment and burnup of the assembly is known at discharge. It is expected that the total neutron count rate measured at discharge will be indicative of the initial enrichment and burnup of that assembly. If the automated robot were to focus on measuring the assemblies in the rack without moving them, the time available would increase immensely.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

TRITON11 - Westinghouse 11*11 BWR fuel design

Westinghouse has developed an 11*11 BWR fuel design referred to as the TRITON11 fuel design. Given the current economic pressure on the utilities, the key development objective was to significantly reduce the fuel cycle costs. Careful considerations were also made to meet the varying requirements between the different BWR utilities. In particular, the fuel was optimized for both short and long cycle operation, as well as for the up-rated cores and higher burnups. In addition to these advances, careful considerations were given to the overall reliability of the new design to further reduce the risk of fuel failures. Further improved fuel economy requires improved thermal margins since both SVEA-96 Optima2 and Optima3 are already highly optimized for reactivity performance at the fuel lattice level. It was realized that better thermal margins must come from adding more fuel rods, thereby reducing the linear heat generation rate and surface heat flux. This requires going beyond the 10*10 fuel lattice. An 11*11 fuel design was chosen as optimum, considering benefits versus manufacturing cost and added complexity. The numbers, dimensions and positioning of fuel rods and internal water channels (passing non-boiling water) were analyzed intensively. The resulting TRITON11 design has 109 fuel rods and 3 central water rods for improved moderation. The paper presents the TRITON11 fuel design. The development objectives are laid out and the resulting performance and reliability improvements are described. The development is in the final phase of testing and validation. Examples of recent tests are presented. The advances in BWR core analysis codes in support of TRITON11 fuel are briefly summarized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Verification of diesel spray ignition phenomenon in dual-fuel diesel-piloted premixed natural gas engine

Dual-fuel (DF) engines, in which premixed natural gas and air in an open-type combustion chamber is ignited by diesel-fuel pilot sprays, have been more popular for marine use than pre-chamber spark ignition (PCSI) engines because of their superior durability. However, control of ignition and combustion in DF engines is more difficult than in PCSI engines. In this context, here we focus on the ignition stability of n-heptane pilot-fuel jets injected into a compressed premixed charge of natural gas and air at low-load conditions. To aid understanding of the experimental data, chemical-kinetics simulations were carried out in a simplified engine-environment that provided insight into the chemical effects of methane (CH4) on pilot-fuel ignition. The simulations reveal that CH4 has an effect on both stages of n-heptane autoignition: the small, first-stage, cool-flame-type, low-temperature ignition (LTI) and the larger, second-stage, high-temperature ignition (HTI). As the ratio of pilot-fuel to CH4 entrained into the spray decreases, the initial oxidization of CH4 consumes the OH radicals produced by pilot-fuel decomposition during LTI, thereby inhibiting its progression to HTI. Using imaging diagnostics, the spatial and temporal progression of LTI and HTI in DF combustion are measured in a heavy-duty optical engine, and the imaging data are analyzed to understand the cause of severe fluctuations in ignition timing and combustion completeness at low-load conditions. Images of cool-flame and hydroxyl radical (OH*) chemiluminescence serve as indicators of LTI and HTI, respectively. The cycle-to-cycle and spatial variation in ignition extracted from the imaging data are used as key metrics of comparison. The imaging data indicate that the local concentration of the pilot-fuel and the richness of the surrounding natural-gas air mixture are important for LTI and HTI, but in different ways. In particular, higher injection pressures and shorter injection durations increase the mixing rate, leading to lower concentrations of pilot-fuel more quickly, which can inhibit HTI even as LTI remains relatively robust. Decreasing the injection pressure from 80 MPa to 40 MPa and increasing the injection duration from 500 µs to 760 µs maintained constant pilot-fuel mass, while promoting robust transition from LTI to HTI by effectively slowing the mixing rate. This allows enough residence time for the OH radicals, produced by the two-stage ignition chemistry of the pilot-fuel, to accelerate the transition from LTI to HTI before being consumed by CH4 oxidation. Thus from a practical perspective, for a premixed natural gas fuel–air equivalence-ratio, it is possible to improve the “stability” of the combustion process by solely manipulating the pilot-fuel injection parameters while maintaining constant mass of injected pilot-fuel. This allows for tailoring mixing trajectories to offset changes in fuel ignition chemistry, so as to promote a robust transition from LTI to HTI by changing the balance between the local concentration of the pilot-fuel and richness of the premixed natural gas and air. This could prove to be a valuable tool for combustion design to improve fuel efficiency or reduce noise or perhaps even reduce heat-transfer losses by locating early combustion away from in-cylinder walls.

42 ENGINEERING↗
DOE OSTI2019

Modeling a fast spectrum molten salt reactor in a systems dynamics fuel cycles code

Molten salt reactor (MSR) concepts are being actively developed by several private companies, with some touting the fuel cycle benefits of this technology. Simulating the deployment of MSRs and their effect on fuel cycle transitions is the role of a system dynamics fuel cycles tool. However, many current fuel cycles tools have difficulty modeling MSR isotopic variations because they were designed for modeling existing reactors based on discrete fuel elements with batch fueling schemes (i.e., charge and discharge). This work presents ORION fuel cycle modeling results using a specific high power density fast spectrum MSR design. For code verification, a single-stage MSR fuel cycle was set up in ORION and results were compared with the results from the SCALE-based reactor physics model (ChemTriton). A transition scenario was set up in ORION from the current light water reactor fleet to a future fleet of MSRs to study the characteristics of an MSR that affect its performance during transition. Several transition scenarios were set with holds and delays in place that are representative of a sodium fast reactor as modeled in Evaluation Group 23 for the Fuel Cycle Options Campaign Evaluation and Screening Study. Several MSR characteristics lend to its high performance and ease in transition from the current fuel cycle: fissile material is only loaded at initial startup; all fissile material is available as it is bred due to online fuel processing; and fissile material recirculating through the core is briefly held in tanks for processing and has very minimal out-of-core time (on the order of a minutes to a few days). These characteristics are in contrast to a solid-fueled reactor, which requires discharge, cooling, and fabrication of fuel all of which could take a few months to a few years and repeated loading of fresh fuel assemblies. In conclusion, a solid-fueled reactor only discharges a fraction of a core every year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molybdenum Sleeve Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.55 cm Pitch)

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM-096, LEUCOMP-THERM-097, LEU-COMP-THERM-101, and LEU-COMP-THERM-102. The purpose of these experiments was to measure the effects of molybdenum in nearly-critical systems. The molybdenum was introduced into the fuel arrays as tubular sleeves that surrounded some of the fuel rods in the fuel arrays measured. Four hundred molybdenum tubes nominally 12.7 mm outside diameter, 498 mm long, with 0.762 mm wall thickness were provided for the experiments by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). Small polyethylene adapters at each end of the tubes were used to center each tube on a fuel rod in the assembly. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 1261 fuel rod positions on a pitch of 0.610 in (1.5494 cm) in a series of 20 hexagonal rings surrounding the central fuel rod. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The five critical experiments in this series were performed in August through December 2022, in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility. Case 1 had no molybdenum sleeves, Case 2 had 208 molybdenum sleeves clustered at the center of the array, Case 3 had 397 molybdenum sleeves clustered at the center of the array, Case 4 had 175 molybdenum sleeves in the central position and in five alternating hexagonal rings, and Case 5 had 331 molybdenum sleeves in the central position and in seven alternating hexagonal rings. All five critical experiments are judged to be acceptable as benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Potential Availability of Alternative Fuel to Supply Maritime Activities in Pacific Northwest Ports

The international shipping sector represented 3% of global greenhouse gas emissions in 2023 (Office of Energy Efficiency & Renewable Energy 2024). International shipping has been classified as a difficult-to-decarbonize industry (IRENA 2024). In an effort to drive decarbonization, the U.S. Department of Energy has partnered with Mission Innovation to co-lead the Zero-Emission Shipping Mission, which launched in 2021(Office of Energy Efficiency & Renewable Energy 2021). In addition, the U.S. Department of State partnered with Norway to launch the Green Shipping Challenge in 2022 (Office of the Spokesperson 2022b). As part of the ZESM and Green Shipping Challenge, the United States is collaborating with the Republic of Korea (ROK) to develop a green shipping corridor (U.S. Mission Korea 2023). The United States and ROK have conducted a pre-feasibility study as the first step in developing a green shipping corridor between the countries. The ports included in the study are Seattle, Tacoma, and Everette in the U.S. Pacific Northwest (PNW) and Busan, Ulsan, and Masan in ROK. The National Renewable Energy Laboratory's role in the study was to analyze the availability and technical potential of alternative marine fuels in proximity to U.S. PNW ports. The findings show most of the existing alternative fuel capacity within the region is from renewable diesel, biodiesel, and sustainable aviation fuel facilities. The largest growth in fuel capacity in the region by 2030 is projected to be in renewable diesel and hydrogen. The overall technical readiness of non-drop-in alternative fuel production and conversion technologies is more developed than alternative-fueled ships and associated fueling infrastructure. However, much of the fuel capacity in the region is comprised of drop-in fuels, making it technically possible to use existing infrastructure for transporting and bunkering to the existing fleet. Data to inform regional alternative fuel quantity estimations were collected from an extensive review of databases, reports, announcements, and other publicly available resources. A maturity index and sector competition factor were applied to announced fuel projects to determine the quantity of alternative fuel available to the marine sector in the region by 2030. Demand data were collected from fuel bunkering logs covering the PNW seaports (State of Washington 2021). Both supply and demand data were converted to very-low sulfur fuel oil gallon equivalents (VLSFO-GE) for better comparison. Qualitative data were gathered through interviews with stakeholders, project developers, and industry experts. Total alternative fuel capacity available to the marine sector in the region is estimated to be 824 million VLSFO-GE per year by 2030. This is sufficient to cover the requirements of a green shipping corridor between the United States and ROK. The findings from this report are being used to inform detailed feasibility studies for several U.S. PNW -ROK green shipping corridors. Updates from the U.S. PNW - ROK feasibility studies will continue to be published on Mission Innovation's green corridor tracking website (Zero Emission Shipping Mission, n.d.). In addition, this report has helped to inform further work on shipping decarbonization in the U.S. PNW, including the Pacific Northwest to Alaska Green Corridor focused on cruise vessels.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2014

Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Mo fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to the irradiation performance of monolithic U-10Mo nuclear fuel. There are several issues or concerns that warrant more detailed study, such as precipitation along cladding-to-cladding bond line, chemical banding, uncovered fuel-zone edge, and interaction layer between U–Mo fuel meat and zirconium. Future post-irradiation examination results will focus, among other things, on identifying in-reactor failure mechanisms and, eventually, directing further fresh fuel characterization efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Understanding Fission Gas Bubble Distribution and Zirconium Redistribution in Neutron-irradiated U-Zr Metallic Fuel Using Machine Learning

U-10wt.% Zr (U-10Zr) based metallic fuel is the leading candidate for next-generation sodium cooled fast reactor in United States. Currently, Idaho National Laboratory (INL) has been the leading national laboratory for research, development, and demonstration (RD&D) on metallic fuel. Advanced post-irradiation characterization will help to understand fuel microstructure and property change during irradiation, benefiting fuel qualification for commercial application. Characterization capabilities ranging from sub-nanometer to micrometer, such as scanning electron microscopy (SEM), focused ion beam (FIB) sampling, transmission electron microscopy (TEM) characterization, and local thermal conductivity microscopy (TCM), have been utilized recently on irradiated U-10Zr fuel samples to gain a better understanding of nuclear fuel microstructure and property evolution inside a reactor. The FIB/SEM coupled with energy dispersive X-ray spectroscopy (EDS) can capture the essential information to achieve better understanding of fuel behaviors. Inside a nuclear reactor, the phase and microstructure of U-10Zr is constantly changing under neutron bombardment. For example, the gaseous fission product atoms have a limited solubility inside fuel matrix and tend to precipitate out in bubble form, which not only contribute to fuel thermal conductivity degradation but also provide a shortcut for movement of fission products, i.e. lanthanides. The resultant deposition of lanthanides at the cladding inner surface will potentially trigger a chemical reaction/interaction between nuclear fuel and cladding at reactor operational conditions, threatening fuel integrity and safety. FIB/SEM coupled with EDS can provide the fission bubble information as well as probe into phase separation or Zr redistribution, which is fundamental to predict the fuel performance. With high velocity image data generating method, such as FIB/SEM, an automatic way to extract the microstructural information quantitively can better serve the needs from post irradiation characterization. A trained machine learning model, named Decision Tree, is employed to generate a bubble classifier and to categorize bubbles into three categories: isolated bubble, connected without lanthanides, and connected with lanthanides bubbles[3]. This work presents a showcase of this approach on six regions of a fuel cross-section along the radial temperature gradient. We obtained distributions of bubble categories and porosity rates along the six regions. Moreover, a secondary phase U-Zr2 was determined and found on regions 5 and 6. The secondary phase fraction was increasing from 15.61% in region 5 to 34.79% in region 6 based on this approach . This quantitative data offers insights into the lanthanide migration and potentially thermal conductivity degradation. This information from machine learning will be fed into fuel design code for better prediction of fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Discrete Modeling of Early-Life Thermal Fracture in Ceramic Nuclear Fuel

Fracturing of ceramic fuel pellets heavily influences performance of light water reactor (LWR) fuel. Early in the life of fuel, starting with the initial power ramp, large thermal gradients cause high tensile hoop and axial stresses in the outer region of the fuel pellets, resulting in the formation of radial and axial cracks. Circumferential cracks form due to thermal gradients that occur when the power is ramped down. These thermal cracks cause the fuel to expand radially, closing the pellet/cladding gap and enhancing the thermal conductance across that gap, while decreasing the effective conductivity of the fuel in directions normal to the cracking. At lower length scales, formation of microcracks is an important contributor to the decrease in bulk thermal conductivity that occurs over the life of the fuel as the burnup increases. Because of the important effects that fracture has on fuel performance, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and abnormal conditions. Modeling fracture within the context of the finite element method, which is based on continuous interpolations of solution variables, has always been challenging because fracture is an inherently discontinuous phenomenon. Work is underway at Idaho National Laboratory to apply two modeling techniques model fracture as a discrete displacement discontinuity to nuclear fuel: The extended finite element method (XFEM), and discrete element method (DEM). XFEM is based on the standard finite element method, but with enhancements to represent discontinuous behavior. DEM represents a solid as a network of particles connected by bonds, which can arbitrarily fail if a fracture criterion is reached. This paper presents initial results applying the aforementioned techniques to model fuel fracturing. This work has initially focused on early life behavior of ceramic LWR fuel. A coupled thermal-mechanical XFEM method that includes discontinuities in both temperature and displacement fields at crack locations has been developed and is being applied to thermal fracture of LWR fuel. A DEM model of coupled heat conduction and solid mechanics has been developed and used to simulate random initiation and propagation of thermally driven cracks during initial power cycles. This DEM model predicts the formation of realistic radial cracking patterns during power rise and circumferential cracks as power is ramped down. These initial results are very encouraging, and these techniques are expected to provide improved understanding of fuel behavior in a wide variety of conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutronic feasibility analysis of fully ceramic microencapsulated fuel for commercial PWRs

The fully ceramic microencapsulated (FCM) fuel is based on tri-structural isotropic (TRISO) coated particles embedded in silicon carbide (SiC) matrix, which is among the accident tolerant fuel concept options. With the strength of high fission product retention, the FCM fuel provides potential superior safety performance. Compared to the conventional standard UO{sub 2} fuel, the deficiency of FCM fuel is its considerable lower fissile inventory, which results in a reduced reactor power or/and life. Supposing the FCM fuel is used as direct replacement of the fuel in commercial PWRs, the spectrum is softened due to the usage of SiC as matrix, and the moderator temperature coefficient is possible to be positive at the beginning of life, leading to high risk during operation. A neutronic feasibility study of FCM fuel utilized in commercial PWRs is performed in the paper. The fuel enrichment and under-moderated lattice are evaluated by analyzing the typical TRISO particle with UO{sub 2} kernels which is widely used in high temperature gas reactors and that with enlarged UN fuel kernels. It was found that under the mandatory constraint of low enriched uranium, the FCM fuel with UO{sub 2} kernels, although relatively mature, is only capable of small reactor designs, and the assembly lattice needs to be modified. The FCM fuel with UN kernels, when directly loading in the commercial PWRs, can yield comparable reactor power and life; however, a higher fuel enrichment and the fuel design and fabrication need to be optimized. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗