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At least 1,639 records · Page 91
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Fuel cell cassette with compliant seal

A fuel cell cassette for forming a fuel cell stack along a fuel cell axis includes a cell retainer, a plate positioned axially to the cell retainer and defining a space axially with the cell retainer, and a fuel cell having an anode layer and a cathode layer separated by an electrolyte layer. The outer perimeter of the fuel cell is positioned in the space between the plate and the cell retainer, thereby retaining the fuel cell and defining a cavity between the cell retainer, the fuel cell, and the plate. The fuel cell cassette also includes a seal disposed within the cavity for sealing the edge of the fuel cell. The seal is compliant at operational temperatures of the fuel cell, thereby allowing lateral expansion and contraction of the fuel cell within the cavity while maintaining sealing at the edge of the fuel cell.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2019

U.S. HIGH PERFORMANCE RESEARCH REACTOR PRELIMINARY DESIGN MILESTONE FOR CONVERSION TO LOW ENRICHED URANIUM FUEL

The U.S. reactors that regularly refuel with highly enriched uranium (HEU) have undertaken conversion to low-enriched uranium (LEU) using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo) that is under development for the U.S. High Performance Research Reactors (USHPRR). The U.S. National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program is pursuing USHPRR conversion. The USHPRR consists of nine national laboratories and three universities working with a uranium processing plant and commercial fuel fabricator to convert five reactors and an associated critical assembly. Feasibility designs were completed in 2009 for all USHPRR. Since then, the USHPRR have completed key milestones leading to preliminary designs. Between 2007 and 2011, the Fuel Qualification (FQ) Pillar conducted seven irradiation campaigns consisting of 14 large size plates and over 60 mini plates with U-10Mo fuel with a zirconium interlayer. The Preliminary Report on U-Mo Monolithic Fuel for Research Reactors submitted for review to the U.S. Nuclear Regulatory Commission (NRC) in 2017 summarizes these irradiation tests. The Fuel Fabrication (FF) Pillar has installed and is operating a fabrication pilot line at BWX Technologies, Inc. In tandem, the Reactor Conversion (RC) Pillar, has been working with each USHPRR to mature conversion designs and to include recent data. As of 2018, all five USHPRR have completed LEU designs. This includes three Preliminary Safety Analysis Reports (PSARs) submitted between 2014 and 2017 to the NRC by the National Institute of Standards and Technology, University of Missouri, and Massachusetts Institute of Technology for their reactors NBSR, MURR, and MITR, respectively. In 2018, the two USHPRR reactors regulated by the U.S. Department of Energy (DOE) at Idaho National Laboratory (ATR) and Oak Ridge National Laboratory (HFIR) completed optimized conceptual or preliminary designs, with ATR confirming a ‘base fuel’ design of the same fabrication complexity as required for MURR, MITR and NBSR. Instead of available U-10Mo designs, HFIR is pursuing a longer silicide (U3Si2) fuel design with a fuel meat density of 4.8 gU/cm3. However, HFIR conversion will require additional fuel qualification due to the complex fuel design containing boron and since HFIR exceeds existing silicide qualification limits. These preliminary designs allow for a commercial fabrication demonstration, and subsequent irradiation qualification of these fuel element designs in order to complete final SARs that allow for USHPRR conversion to LEU fuel.

Wilson, E. H.↗
DOE OSTI2019

Impacts of Fast-Spectrum Molten Salt Reactor Characteristics on Fuel Cycle Performance

The System Analysis and Integration Campaign nuclear fuel cycle evaluation and screening study was conducted to outline the potential benefits and challenges of nuclear fuel cycle options. The study identified continuous recycle in fast critical reactors as a common characteristic shared among some of the most promising future nuclear fuel cycles. This study was technology agnostic to focus on the underpinning physics that drives fuel cycle outcomes. This process used an analysis example to generate a physics-based understanding of the fuel cycle, and it yielded performance metrics over ranges rather than comparing absolute numbers. While no fast-spectrum molten salt reactor (MSR) has ever been built, several concepts exist using different fuels and carrier salts, many of which target high-burnup objectives, such as lightwater reactor nuclear waste consumption. Putting the resulting fuel cycle impacts of these reactor choices (e.g., fuel and neutron spectrum) into perspective is informative for our current understanding of a given fuel cycle’s performance, and it also demonstrates the applicability of MSRs to different fuel cycles. It has been established that in a fuel cycle transition, performance of a given fast-spectrum MSR depends in part on the initial fuel loading requirements and processing methods.

Betzler, Benjamin↗
DOE OSTI2022

High-Assay Low Enriched Uranium Pyroprocessing and Electrometallurgical Treatment at the Fuel Conditioning Facility

Fuel Conditioning Facility (FCF) fulfills part of INL’s mission by reprocessing irradiated sodium-bonded fuel from the EBR-ll reactor and participates in research to further explore the fuel cycle for recovery of fuel from a variety of current reactors in use today. The fuel cycle of a nuclear reactor under the current non-proliferation act, leaves unused fuel in the spent fuel rods. The rods are safely stored until their fuel can be recovered. ? Fuel recovery from the EBR-II fuel pins has several steps to separate the uranium from the salts. The pyroprocessing technology involves high - temperatures, chemical and electrochemical methods for separating the unused uranium from the salts, fission products, and used fuel. The recovered uranium is then remixed and formed into ingots or Regulus with an enrichment of < 20% U-235. This final product is a source to fuel current nuclear reactors as well as research and development for future nuclear reactors

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Criticality Analysis of FSV Spent Nuclear Fuel in the DOE Standard Canister

The U.S. Department of Energy (DOE) is responsible for managing over 300 types of spent nuclear fuel (SNF). To manage this large variety of fuel types, DOE plans to employ standardized canisters for the transportation, long-term storage, and eventual disposal of SNF. Idaho National Laboratory is currently supporting DOE’s SNF Packaging Demonstration Project, in which Fort Saint Vrain (FSV) fuel assemblies will be loaded into a DOE Standard Canister. This paper presents criticality calculations demonstrating that all four or five FSV fuel assemblies loaded into the DOE Standard Canister will remain subcritical in any expected normal or credible abnormal conditions. Previous criticality analyses were performed for one FSV fuel assembly and 12 Peach Bottom Core 2 fuel elements loaded into a DOE Standard Canister. This paper covers the criticality analysis performed for loading both four and five FSV fuel assemblies into a DOE Standard Canister. Various intact and degraded mode configurations were modeled in conducting the criticality calculations. This analysis encompassed three different configurations: (1) a single DOE Standard Canister loaded into a concrete storage overpack, (2) seven DOE Standard Canisters loaded into a concrete storage overpack, and (3) nine DOE Standard Canisters loaded into a concrete storage overpack. The overpack dimensions were varied for each of the three configurations, and transport, storage, and disposal scenarios were analyzed for each configuration. For transport scenarios, a pair of degradation cases was analyzed. In the first case, the fuel compacts became degraded and were removed from the fuel block, then deposited at the bottom of a horizontally placed canister, thereby simulating a drop event. The canister was considered to remain intact. In the second case, the spacing between horizontally placed canisters in a nine-canister overpack was reduced such that the canisters were piled on top of each other, simulating a drop event. For this case, no degradation of the canister internals or fuel was considered. For storage scenarios, the water moderator location in the system was varied to enable identification of the most reactive configurations. Dry and wet conditions were analyzed for the fuel materials, canister, and overpack. For disposal scenarios, two degradation cases were analyzed. In the first, the stainless-steel internals of the canister degraded to either hematite or goethite under both dry and wet conditions. In the second case, degraded FSV fuel formed a uranium-water slurry that filled the coolant/void holes. None of the cases presented exceeded the application specific upper subcritical limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS

Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.

Harp, Jason↗
DOE OSTI2019

Neutronic and thermal-hydraulic feasibility studies for High Flux Isotope Reactor conversion to low-enriched uranium silicide dispersion fuel

An iterative design process involving neutronic and thermal-hydraulic modeling and simulation has been employed to assess the feasibility of converting the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) silicide dispersion fuel. ORNL is funded by the National Nuclear Security Administration to evaluate HFIR conversion. Previous HFIR conversion studies focused on U-10Mo monolithic fuel; however, due to potential fabrication issues with the complex HFIR U-10Mo fuel design, ORNL is evaluating U 3 Si 2 -Al dispersion fuel as an alternative LEU fuel system.Fueled by 10.1 kg of HEU and operated at 85 MW, HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world. Retrofitting a compact, high-power density, HEU-based core with LEU is a challenging problem to solve, especially when considering the conversion requirements. Neutronic and thermal-hydraulic analyses were performed with Shift and HSSHTC, respectively, to predict reactor performance and thermal safety margins. A number of designs were proposed and evaluated using an iterative approach in an effort to show that reactor performance could match that obtained using HEU fuel and that thermal safety margins were adequate. Furthermore this study concludes that conversion of HFIR with U 3 Si 2 -Al LEU fuel is feasible if, among other requirements, the fuel meat region is centered and symmetric about the fuel plate thickness centerline, the active fuel zone length is increased from 50.80 cm to 55.88 cm, the proposed fabrication tolerances can be met, and the fuel can be qualified for HFIR conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Role of fuel properties and piston shape in influencing soot oxidation in heavy-duty low swirl diesel engine combustion

The presented study evaluated effects on soot emissions of important diesel fuel properties (e.g. volatility T95, hydrogen to-carbon ratio H/C and oxygen content O/C) in a heavy-duty, low-swirl, direct injection diesel engine operated with cooled exhaust gas recirculation to control the emissions of nitrogen oxides. Variation of injection pressure was conducted in three representative operation points in the single-cylinder engine including conventional and wave piston bowl types. Additionally, detailed optical combustion studies were performed in a high pressure/high temperature combustion chamber with an engine-like flame-wall interaction geometry. The fuel properties mainly influenced parameters relating to events near the nozzle, such as the ignition delay, flame lift-off distance and spray air entrainment. Oxygenation of the fuel reduced soot formation (a known effect of spray-jet dilution), but the fuels’ H/C and T95 also influenced soot emissions. Viscosity, heating value and density, treated as secondary fuel parameters, affected the injected fuel’s kinetic energy, and hence mean flow field and local turbulent mixing late in the combustion cycle. The results suggested that fuel effects on soot oxidation during late phases of combustion are mainly governed by parameters related to the fuel injection profile under the specific cylinder gas conditions. With an oxygenated fuel, flow structures (local turbulent side vortices) which are typical for the investigated engine type were found to form in a similar way as with a reference fuel. Hence, improved internal flow using a wave piston in combination with a low-sooting oxygenated fuel gave additional reductions in net soot emissions.

42 ENGINEERING↗
DOE OSTI2026

Redefining fuel heating value for engines: Accounting for heat of vaporization

Defining a fuel's heating value (i.e., energy content) is fundamental for calculating engine efficiency and for life cycle analysis comparisons between different fuels. Traditional definitions of lower heating value and higher heating value account for the effect of water vapor versus liquid water in the exhaust, which is important when the fuel is used in a furnace or boiler. In an engine, it is equally important to properly account for the energy required to vaporize liquid fuel. Heat of vaporization has a small effect for common hydrocarbon fuels, typically less than 1% of lower heating value, but the effect is much larger for other important fuels such as ethanol (3.4% of lower heating value) and methanol (5.9% of lower heating value). This paper defines a new type of fuel heating value that more accurately reflects the useful fuel energy content for engines. Vaporized heating value is defined as the heating value when starting with a vaporized fuel instead of a liquid fuel. It can be calculated by adding the fuel's heat of vaporization to the traditional lower heating value. This paper illustrates the rationale and benefits of using vaporized heating value using data from the literature.

09 BIOMASS FUELS↗
DOE OSTI2020

α-U and ω-UZr 2 in neutron irradiated U-10Zr annular metallic fuel

Here, to develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boost the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissions per initial heave metal atoms for initial screening purpose. After irradiation, the initial U-10Zr separated into an a-U annular region and an UZr 2+x center region with nanoscale spinodal decomposed microstructure. Because of the provided large amount of coherent interface areas, the fission gas atoms and vacancies generated in UZr 2+x phase are possibly pinned at the interface areas, leading to 20 times smaller fission gas bubbles than those in the neighboring a-U. The large bubbles in a-U become connected and merged into large pores that provide fast release paths for fission gas which prevents further fuel swelling. The fuel center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A Critical Review of High Burnup Fuel Fragmentation, Relocation, and Dispersal under Loss-Of-Coolant Accident Conditions

High burnup fuel in excess of ~6268 GWd/MTU has been observed to severely fragment when subjected to temperature transient conditions as in a loss-of-coolant accident. In the event of cladding failure, the severely fragmented fuel can relocate into the balloon region and potentially disperse through the burst opening into the reactor’s primary system. This observation has led to implementation of several independent test programs designed to focus on various aspects of fuel fragmentation, relocation, and dispersal. The purpose of this review is to perform a critical, holistic assessment of fuel fragmentation, relocation, and dispersal under loss-of-coolant conditions to identify data gaps in the experimental data base. Phenomena have been identified and well defined through the various testing programs, and general agreement regarding the governing parameters (temperature, burnup, heating rate, cladding deformation, etc.) has been established. However, there is a significant data gap connecting research to commercial application. Furthermore, previous research attempted to evaluate the problem holistically, but has yet to provide a comprehensive understanding of the problem. Major identified data gaps consist of (1) a comprehensive understanding of pretransient fuel rod conditions (fuel temperature, fuel stress, rod internal pressures, microstructure, etc.), (2) definition of prototypic loss-of-coolant accident conditions (heating rate, fuel temperatures, fuel stress), (3) identification of the differences between nuclear and electrical integral loss-of-coolant accident tests, and connection of fuel rod performance (cladding ballooning, burst opening, potential relocation, fragmentation susceptibility, etc.) to fuel fragmentation, relocation, and dispersal severity. Finally, the content of this manuscript provides a clear path for connecting testing and data acquisition to commercial application by providing concrete means for filling data gaps and prioritized testing regimes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Validating the BISON fuel performance code to integral LWR experiments

BISON is a modern finite element-based nuclear fuel performance code that has been under development at the Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON’s computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Our results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. In the initial rod diameter comparisons they were unsatisfactory and have lead to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Reactor performance and safety characteristics of ThN-UN fuel concepts in a PWR

The reactor performance and safety characteristics of mixed thorium mononitride (ThN) and uranium mononitride (UN) fuels in a pressurized water reactor (PWR) are investigated to discern the potential nonproliferation, waste, and accident tolerance benefits provided by this fuel form. In this paper, we present results from an initial screening of mixed ThN-UN fuels in normal PWR operating conditions and compares their reactor performance to UO 2 in terms of fuel cycle length, reactivity coefficients, and thermal safety margin. ThN has been shown to have a significantly greater thermal conductivity than UO 2 and UN. Admixture with a UN phase is required because thorium initially contains no fissile isotopes. Results from this study show that ThN-UN mixtures exist that can match the cycle length of a UO 2 -fueled reactor by using 235 U enrichments greater than 5% but less than 20% in the UN phase. Reactivity coefficients were calculated for UO 2 , UN, and ThN-UN mixtures, and it was found that the fuel temperature and moderator temperature coefficients of the nitride-based fuels fall within the acceptable limits specified by the AP1000 Design Control Document. Reduced soluble boron and control rod worth for these fuel forms indicates that the shutdown margin may not be sufficient, and design changes to the control systems may need to be considered. The neutronic impact of 15 N enrichment on reactivity coefficients is also included. Due to the greatly enhanced thermal conductivity of the nitride-based fuels, the UN and ThN-UN fuels provide additional margin to fuel melting temperature relative to UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Conversion from film to image plates for transfer method neutron radiography of nuclear fuel

This paper summarizes efforts to characterize and qualify a computed radiography (CR) system for neutron radiography of irradiated nuclear fuel at Idaho National Laboratory (INL). INL has multiple programs that are actively developing, testing, and evaluating new nuclear fuels. Irradiated fuel experiments are subjected to a number of sequential post-irradiation examination techniques that provide insight into the overall behavior and performance of the fuel. One of the first and most important of these exams is neutron radiography, which provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Results from neutron radiography are often the driver for subsequent examinations of the PIE program. Features of interest that can be evaluated using neutron radiography include irradiation-induced swelling, isotopic and fuel-fragment redistribution, plate deformations, and fuel fracturing. The NRAD currently uses the foil-film transfer technique with film for imaging fuel. INL is pursuing multiple efforts to advance its neutron imaging capabilities for evaluating irradiated fuel and other applications, including conversion from film to CR image plates. Neutron CR is the current state-of-the-art for neutron imaging of highly-radioactive objects. Initial neutron radiographs of various types of nuclear fuel indicate that radiographs can be obtained of comparable image quality currently obtained using film. This paper provides neutron radiographs of representative irradiated fuel pins along with neutron radiographs of standards that informed the qualification of the neutron CR system for routine use. Additionally, this paper includes evaluations of some of the CR scanner parameters and their effects on image quality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Progress and Prospective of Nitrogen-Based Alternative Fuels

Alternative fuels are essential to enable the transition to a sustainable and environmentally friendly energy supply. Synthetic fuels derived from renewable energies can act as energy storage media, thus mitigating the effects of fossil fuels on environment and health. Their economic viability, environmental impact, and compatibility with current infrastructure and technologies are fuel and power source specific. Nitrogen-based fuels pose one possible synthetic fuel pathway. In this paper, we discuss the progress and current research on utilization of nitrogen-based fuels in power applications, covering the complete fuel cycle. We cover the production, distribution, and storage of nitrogen-based fuels. We assess much of the existing literature on the reactions involved in the ammonia to nitrogen atom pathway in nitrogen-based fuel combustion. Furthermore, we discuss nitrogen-based fuel applications ranging from combustion engines to gas turbines, as well as their exploitation by suggested end-uses. Thereby, we evaluate the potential opportunities and challenges of expanding the role of nitrogen-based molecules in the energy sector, outlining their use as energy carriers in relevant fields.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Thermally-Conductive and Mechanically-Robust Graphene Nanoplatelet Reinforced UO 2 Composite Nuclear Fuels

Low thermal transport behavior along the radial direction of nuclear fuel pellets and pellet-cladding mechanical interaction significantly impact fuel performance and the safety of current nuclear energy systems. Here we report a new strategy of advanced fuel design in which highly thermally-conductive and mechanically-robust graphene nanoplatelets are incorporated into UO 2 fuel matrix to improve fuel thermal-mechanical properties. The 2D geometry of the graphene nanoplatelets enables a unique lamellar structure upon fuel consolidation by spark plasma sintering. The thermal conductivity along the radial direction of the sintered fuel pellets at room temperature reaches 12.7 and 19.1 wm –1 K –1 at 1 wt.% and 5 wt.% loadings of the graphene nanoplatelets, respectively, representing at least 74% and 162% enhancements as compared to pure UO 2 fuel pellets. Indentation testing suggests great capability of the 2D graphene nanoplatelets to deflect and pin crack propagation, drastically improving the crack propagation resistance of fuel matrix. Here, the estimated indentation fracture toughness reaches 3.5 MPa·m 1/2 by 1 wt.% loading of graphene nano-platelets, representing a 150% improvement over 1.4 MPa·m 1/2 for pure UO 2 fuel pellets. Isothermal annealing of the composite fuel indicates that the graphene nano-platelet is able to retain its structure and properties against reaction with UO 2 matrix up to 1150 °C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Emission factors of industrial boilers burning biomass-derived fuels

Boilers are combustion devices that provide process heat and are integral to many industrial facilities. Historically, outside of the pulp and paper industry, most boilers burned fossil fuels, although interest in decarbonization has been leading to an increased use of renewable fuels in boilers. These boilers, including those in the biorefineries, are often large sources of air pollutant emissions, and the characterization of these emissions is critical to obtaining air permits and ensuring protection of the surrounding air quality. Several industrial boilers and new biorefineries allow utilization of biomass-derived fuels (e.g. wastewater sludge, lignin, etc.) produced during their operation as a fuel for the boiler to meet process energy needs. However, there is limited empirical data on emission factors for the burning of unconventional fuels, such as solid-gas mixtures containing biomass residues. To fill this gap, we carry out a comprehensive data survey, collecting information on emission factors for boilers burning either a single or a mixture of solid and gaseous biomass-derived fuels. We review multiple hard-to-obtain and unconventional data sources, such as air permit applications, stack test data, and industry-sponsored data collection efforts, to compile emission factors for biomass-derived fuels. We then compare this data with wood residue emission factors from the U.S. Environmental Protection Agency’s AP-42 emission factor database. Our results indicate that the emission factors for boilers burning unconventional fuels vary widely depending on the fuel composition, boiler type, and fuel characteristics. Overall, we find that median emission factors of selected biomass-derived fuels are typically lower than those of wood residue boilers in AP-42. The information collected herein could be useful to permitting agencies and industries utilizing boilers who may want to reduce the carbon impact of their facilities by combusting biomass-derived wastes for process energy needs, for more accurate emission estimation for permitting.

09 BIOMASS FUELS↗
DOE OSTI2018

Irradiation of Miniature Fuel Specimens in the High Flux Isotope Reactor

The Advanced Fuels Campaign (AFC) within the US Department of Energy (DOE) Office of Nuclear Energy is considering several advanced ceramic fuels for use in light water reactors (LWRs) and/or advanced reactor concepts. Advanced fuels, such as uranium carbide (UC), uranium nitride (UN), and uranium silicide (U 3 Si 2 ) offer increased uranium density and enhanced thermophysical properties compared to conventional UO 2 fuel, while maintaining an acceptably high melting point. Enhancements to traditional UO 2 are also being considered that would increase the thermal conductivity of the fuel and fission product retention through various additives. However, some of these advanced fuels have very limited information available on their irradiation performance (microstructural evolution, swelling, fission gas release, etc.), particularly for the range of temperature and burnup that are relevant for LWR fuel pins. To address this issue, Oak Ridge National Laboratory has developed an experimental facility to irradiate miniature fuel specimens in the High Flux Isotope Reactor (HFIR). The small size of the fuel specimens simplifies the design, analysis, and post-irradiation evaluations. Post-irradiation examination will provide basic data on the stability and behavior of advanced fuels as a function of temperature and burnup. This report briefly summarizes the experiment facility design concept and the assembly of the first set of experiments. The experiment contains six different types of UN-based fuel kernels UN TRISO particles with varying densities, impurity levels, and burnable absorber contents. The experiments were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the assembly process are included in this report. The experiment is planned for insertion into the HFIR during cycle 480 (June 2018).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗