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In-situ thermomechanical measurements during burst testing of Zr claddings using digital image correlation and infrared thermography techniques

Burst testing has long been a relied-on method for assessing fuel cladding materials in light-water reactoraccident scenarios. However, burst testing historically provided minimal in-situ observations beyond a few thermocouple readings, pressure from transducers, and post-test deformation measurements. Acquisition of full cladding thermomechanical data throughout a burst test would be invaluable to fuel system qualification, material model development and fuel performance code validation, such as the BISON fuel performance code. Recently, methodologies for simultaneously applying 2-dimensional digital image correlation and infrared thermography techniques to quantify the relationship between temperature, internal cladding pressure, and deformation during burst testing were developed. This work details results from the application of these techniques to burst testing of Zircaloy-4 cladding segments subjected to loss-of-coolant accident transients in an air environment. Two types of tests were performed with: (a) cladding segments connected to a fixed pressure reservoir as well as (b) closed systems where internal pressure increased during transients. In-situ thermomechanical measurements including strain and thermal gradient data are presented, demonstrating the effectiveness of applying combined surface measurement techniques to burst testing

Bell, Sam↗

Research, Development and Demonstration of Advanced Vehicle Technologies in Clean Motor Fuels (CRADA Final Report)

NREL/DOE has been funding research, development, and demonstration (RD&D) projects to facilitate the deployment of alternative fuels into the marketplace in order to achieve three objectives: 1) enhance the reduction of mobile source emissions in California and the U.S.; 2) increase U.S. competitiveness and economic prosperity; and 3) preserve the environment through the reduction of emissions and toxins from the motor vehicle population. The Lewis-Presley Air Quality Management Act requires the South Coast Air Quality Management District to establish a program to encourage participation in projects to increase the utilization of clean-burning fuels. The parties seek to collaborate on a joint RD&D project to perform a field test on a small vehicle fleet operating on Fischer-Tropsch synthetic diesel fuel. Recently the 106th U.S. Congress designated "gas-to-liquid" (GTL) transportation fuels that are produced from domestic products as alternative fuels under the Energy Policy Act of 1992 (EPACT). Niche markets could develop quickly for these fuel products, perhaps especially for the use of Fischer-Tropsch synthetic fuels in centrally fueled vehicle fleets in California. NREL and its project partners from the U.S. Department of Energy and West Virginia University have recently conducted what we believe to be the first-ever controlled tests of Fischer-Tropsch synthetic fuels in heavy vehicles. These tests have shown that Fischer-Tropsch fuels can be substituted in unmodified trucks and buses without any detectable loss in drivability and performance. Compared to a California diesel fuel baseline, neat Shell Fischer-Tropsch fuel emitted about 12% lower NOx and 24% lower particulate matter over a five-mile driving route while maintaining the same fuel economy on an energy equivalent basis. Whereas these exhaust emissions reductions from these previous studies have been impressive, far greater exhaust emissions reductions can be achieved if the vehicles are retrofitted with emissions control technologies that are enabled by Fischer-Tropsch fuels. For this program, a test fleet would be selected and retrofitted with continuously regenerating traps (CRT’s) or diesel particulate filters (DPF's). The University of West Virginia, under a separate subcontract to SCAQMD will have the responsibility to test and measure the emissions from the test fleet of vehicles.

33 ADVANCED PROPULSION SYSTEMS↗

ATF-2 Ramp Conceptual Design Report

The following report documents the conceptual design for the ATF-2 Ramp experiment. The experiment aims to fill the in-pile irradiation testing gap to conduct integral ramp testing, which was created by the closure of R2, Osiris, and Halden test reactors. The concept involves the simultaneous ramping of three fuel pins using a power axial locator mechanism in the Loop-2A testing facility in the center flux trap of the advanced test reactor (ATR). The three fuel pins are ramped in individual coolant channels containing a prototypic PWR environment. Rod failure will be detected using a fuel rod elongation sensor attached to the fuel pin upper end cap. An axial stack of concentric hafnium and zirconium shrouds will be used to shape the flux around the test pins to create different power levels in each pin and to ensure the peak power location of each pin remains in the center of the test rod. Monte Carlo simulations are used to demonstrate the viability of this design concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Drying of Spent Nuclear Fuel: Considerations and Examples

A recently revised American Society for Testing and Materials consensus standard guide for drying of spent nuclear fuel (SNF) provides considerations and guidance for preparing SNF for its safe storage in a sealed dry storage system. The standard discusses (1) needs for drying, (2) techniques to dry, and (3) demonstration of adequate dryness. No specific approach is prescribed since the effective techniques and needs for drying depend on the specifics of the fuel and dry storage system. Here, this paper discusses these topics using examples for both Zr-alloy-clad commercial SNF and for aluminum-alloy-clad research reactor SNF. Residual water can include free water (liquid and/or vapor), physisorbed water bound to internal surfaces, and chemisorbed water incorporated into surface films, such as (oxy)hydroxides. The potential impacts of these residual waters are corrosion/oxidation, radiolytic breakdown into gaseous and/or reactive species, and canister pressurization. For commercial SNF, inadvertent free water, even up to large amounts (e.g., 10+ mol), is not predicted to cause adverse corrosion degradation, except possible fuel oxidation for breached SNF. For aluminum-clad SNF, the production of radiolytic hydrogen with contribution from the chemisorbed water in its hydrated oxides is a primary consideration. For both SNF systems, canister pressurization is predicted to be well within the canister design, and flammability would not pose a safety concern using an oxygen limit of 5 vol % criterion. (Flammability control can be achieved by limiting either hydrogen or oxygen, and an oxygen limit is expected to be easier to meet in the presence of radiolytic H 2 generation.) The two primary technologies for SNF drying, vacuum drying and forced-gas dehydration, are described herein, and drying tests and campaigns using these methods are cited. Dryness criteria and the methods used to detect and measure residual (free) water are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-Cell Deployment and First Use of Digital Image Correlation for In-Situ Strain Analysis of Irradiated Nuclear Fuel Rods During LOCA Transient

Digital image correlation (DIC) is a noncontact, optical method increasingly used across industries and research environments for acquiring multidimensional strain data. At Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS), DIC has been extensively applied to study the thermomechanical response of nuclear fuel claddings, yielding fundamental insights into material behavior. However, these efforts have focused exclusively on unirradiated materials and relied on the SATS out-of-cell infrastructure. Efforts over the past two years have been made to extend these capabilities to ORNL’s Irradiated Fuels Examination Laboratory SATS system within the hot-cells to enable testing of irradiated fuel cladding materials. Integrating DIC into the hot-cell SATS infrastructure presents unique challenges, including enabling remote operation of optical equipment and adapting auxiliary hot-cell systems for DIC implementation. This report discusses design, stand-up testing, and application of DIC to an irradiated nuclear fuel cladding segment. A DIC testing rig was successfully built and validated out-of-cell through extensive surrogate tests and calibrations. The rig and specialized DIC furnace were installed in the hot-cell, and a DIC test was successfully conducted. Results from that test correspond to expectations for Zr-based alloys and showed similar uncertainties compared to out-of-cell tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Leveraging the High Flux Isotope Reactor for nuclear fuel development: a review of experiments, facilities, and capabilities

Materials testing reactors (MTRs) have been used to develop in-core nuclear fuels and materials since the outset of the nuclear power industry. However, the closure of prominent MTRs worldwide and protracted construction timelines for new facilities have increased reliance on existing infrastructure for near-term irradiation testing needs. One facility that can support these needs is the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. HFIR boasts the highest steady-state neutron flux in the Western Hemisphere and, among other roles, has been used to rapidly administer high fluences on fuels and materials for fission and fusion reactor applications. This paper reviews HFIR facilities and infrastructure, fuel-bearing irradiation experiments conducted in HFIR, and select nonfueled experiments that demonstrate advanced techniques transferable to fuels experiments. Collectively, these examples underscore HFIR's potential role as a nuclear fuels testbed supporting both the existing reactor fleet and advanced reactor fuel development.

Fuel qualification↗

In-situ strain behavior and BISON simulations of Zircaloy cladding subjected to temperature cycling separate-effects tests in a steam environment

Understanding fuel system performance during anticipated transients without scram (ATWSs) in boiling water reactors (BWRs) is necessary for refining current and future safety limits. High-fidelity material models and simulations are fundamental to rigorous assessment of zirconium-based cladding performance. However, experimental thermomechanical data during simulated ATWSs to validate these modes are limited. To provide relevant in-situ data, Zircaloy-4 cladding was subjected to cyclic heating in a steam environment to simulate an out-of-pile BWR ATWS. Digital image correlation was used to capture the cladding strain behavior in-situ for comparison against simulations using the BISON finite element code. Conventional high-temperature models were compared using multiple schemes to gain a better understanding of the applicability of three BISON models to BWR ATWS: (1) the default combination of creep models in BISON, (2) the high-temperature Erbacher model alone, and (3) the low-temperature Limback-Andersson model alone. The cases run with the Limback-Andersson model alone produced the lowest root mean square error (RMSE). The lowest RMSE for the Limback-Andersson model alone was 0.659%, and the highest RMSE reported was 4.22%. A data gap within the model in the temperature regime of interest was also identified, and to account for this gap, the current model in BISON is linearly interpolated between two separate datasets. In conclusion, this evaluation highlights the need to either develop a new model or to improve the existing model to capture transient creep effects resulting from a cyclic temperature transient.

Anticipated transient without SCRAM↗

Real-time Canister Welding Health Monitoring and Prediction System for Spent Fuel Dry Storage

In this report, the salt mist corrosion testing at controlled temperature was carried out successfully on 304L SS submerged arc welded coupons by Unifog Dispersion Tower at ORNL. Corrosion in-situ monitoring system from Intelligent Automation, Inc. (IAI) was setup in the corrosion test and running smoothly to record signals during the corrosion test. Parts and device selection, sensors attachment, and experimental procedures are discussed. There wasn’t any visible corrosion or crack propagation on the weld coupon after about a month of corrosion testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The Novel Charfuel® Coal Refining Process 18 TPD Pilot Plant Project for Co- Producing an Upgraded Coal Product, and Commercially Valuable Co- Products: Area of Interest #3 – Coal Beneficiation Pilot Plant Testing (Final Report)

Operation of Carbon Fuels, LLC’s (“CF”) existing, permitted 18 TPD pilot plant located in Golden, Colorado using two individually ranked (ASTM D 388) coal types (two campaigns), employing the novel Charfuel® coal refining process to produce an upgraded coal product and a number of high-valued organic and inorganic coproducts (for which there presently exists large commercial markets) in order to produce engineering and product data which will then be utilized toward the design of a commercial scale integrated facility (pre-feed document). Carbon Fuels, LLC has developed the Charfuel® Coal Refining Process which refines domestically abundant, raw coal (in the same manner as crude oil is refined) to produce the identical, high value co-products that are refined from crude oil. Thus, gasoline, jet fuel, “green diesel”, fuel oil, and marine fuels, as well as petrochemicals such as benzene, toluene, xylene, and methanol are refined from raw coal using this process. The Charfuel® Coal Refining Process is not a coal conversion process, like pyrolysis, or indirect liquefaction. Nor is it an alternative energy system. Rather it is a coal refining process that has the ability to economically produce products traditionally associated with the refining of crude oil but using only abundant, raw coal as the refinery feed stock. The Charfuel® Coal Refining Process is more economical than crude oil refining and is environmentally benign. Therefore, this value added process yields a return on investment well above 50% for a commercial facility. Furthermore, the Charfuel® process, unlike alternatives such as ethanol and hydrogen, can utilize the existing transportation, delivery, and other petroleum based systems. Hence, there is no need for new engines, pipelines, tankers, or product acceptance. As a result, the profitability of the process is increased. Objectives: (1) Operation of the integrated 18 tpd pilot plant, using two coal types (ranks); (2) Demonstration of process flexibility in being able to produce different products (gas, liquid, and char), as well as determination of operating parameters for identifying scale up criteria for two coal types (ranks); (3) Generation of engineering and design information (process specifications) for use in designing a commercial scale plant (scale-up); (4) Determination of important environmental issues surrounding the process and the products such as fate of trace elements (mercury and other heavy metals) and distributions of SO2, NOx, and CO 2 by analysis of effluent streams; (5) Production of sufficient product to allow reliable commercial economic evaluation of both the refined coal product and the coproducts; and, (6) Assessment of longer-term reliability of unit operations. Period 1: reconfiguration of the 18 TPD plant to meet specific FOA requirements and to qualify the facility for operation; and, Period 2: operation of the 18 TPD plant for two campaigns using two coals types (ranks) which are widely commercially used and abundant - the first being a subbituminous (Powder River Basin (“PRB”)) coal, and the second a bituminous (Illinois #6) coal.

01 COAL, LIGNITE, AND PEAT↗

Idaho National Laboratory Advanced Reactor Technologies Quality Assurance Program Plan

This QAPP covers INL ART R&D, licensing, and program support activities. QA implementing procedures will be identified for all activities to ensure work is documented and the results are capable of successfully completing a technical review by DOE. For work performed by other organizations under contract or other agreements with INL, applicable requirements will be flowed down through the contract or other agreements, but the organizations will be responsible to implement the flowed down requirements through their own specific implementing procedures. This QAPP also covers implementation of software QA (SQA) requirements for software used exclusively by INL ART. Software that is used to support multiple INL programs/projects is not covered by this QAPP; the responsibility for implementing SQA requirements for that software resides with the owners and respective managers of the software. This QAPP does not cover activities for installing and operating experiments in the Advanced Test Reactor (ATR) at INL, nor does it cover ATR facility modifications designed to meet INL ART testing needs (such as temperature, pressure, gas flow, and control systems). Those activities are controlled by the INL QAP and specific ATR implementing procedures. ATR facility management and INL ART management jointly approve experiment designs and functional system requirements. This QAPP does apply to INL ART R&D experiments that will be inserted into ATR and any support activities associated with those R&D activities. This QAPP does not cover activities for installing and operating general post irradiation examination (PIE) test equipment in Materials and Fuels Complex (MFC) facilities. Those activities are controlled by the INL QAP and MFC implementing procedures. However, this QAPP does apply to MFC facility modifications designed to meet specific INL ART testing needs. It also applies to INL ART experiments that will have PIE performed on them and any support activities associated with the PIE activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods

At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

ROADRUNNER uranium nitride MiniFuel: Experimental design, fabrication and pre-irradiation baseline characterization for accelerated burnup testing

Uranium nitride (UN) is a promising fuel candidate for advanced reactor systems owing to its high uranium density and thermal conductivity; however, its qualification remains constrained by the scarcity of well-controlled irradiation performance data. Here, to address this limitation, the ROADRUNNER (Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments) campaign employs the MiniFuel platform in the High Flux Isotope Reactor (HFIR) to enable accelerated burnup irradiation testing under tightly controlled and largely isothermal conditions. This paper presents the experimental design, fuel fabrication, and pre-irradiation baseline characterization of the ROADRUNNER UN MiniFuel campaign. Thirty-six UN minidisc specimens were fabricated with systematically varied as-fabricated density (86–96% of theoretical density), carbon impurity content (961–5240 ppm), oxygen content (≤ ∼2000 ppm), and grain size (2.5–24 μm). The irradiation matrix spans nominal fuel temperatures of 873 K, 1173 K, and 1473 K and target burnups of 3.75%, 6.0%, and 7.5% fissions per initial metal atom (FIMA). Neutronic and thermal analyses were performed to define specimen-specific burnup accumulation and temperature histories, establishing the boundary conditions for subsequent in-pile behavior. Comprehensive pre-irradiation characterization—including dimensional metrology, density verification, impurity analysis, X-ray diffraction, Raman spectroscopy, scanning electron microscopy, X-ray computed tomography, and confocal profilometry—provides a detailed baseline for post-irradiation examination. Pre-irradiation data were further used to generate predictive estimates of fission gas release and swelling using existing empirical correlations. This quantitative comparison reveals substantial inter-model divergence at intermediate and elevated temperatures that exceeds propagated input uncertainties, highlighting structural gaps in the historical irradiation database. The ROADRUNNER irradiation campaign is currently underway in HFIR, with initial firs cycle completed in late 2025 and remaining targets scheduled through 2027. The experimental design and baseline dataset presented here establish the framework needed to interpret forthcoming post-irradiation measurements and to provide discriminating data for the validation and refinement of physics-based UN fuel performance models.

Lopes, Denise Adorno [Oak Ridge National Laborator↗

FY23 Post-LOCA Characterization of Irradiated Fuel (NSUF Award 17-12985)

This report summarizes the sample preparation and data collection on fuel pins following integral LOCA testing in the Severe Accident Test Station at Oak Ridge National Laboratory. The test sample cross-sections show post-transient fragmentation behavior of UO 2 commercially irradiated above 67 MWd/kgU. Ring compression testing and hydrogen measurements were collected on the cladding segments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Identifying Challenges in Safeguards for Metallic Fuel Fabrication Facilities

As new advanced reactors gain popularity, there is an increasing interest in metallic fuel fabrication for fast reactors. While metallic fuels themselves are not a new idea, as many of the first reactors employed metallic fuels, new designs, compositions, and fabrication methods are appearing throughout the nuclear community. As the interest grows and facilities are constructed, both domestic and international safeguards will need to be heavily involved to support safeguards-by-design (SBD) measures from the start. This work compiles a review of historical and modern fuel types and fabrication methods, fabrication processes, safeguards gaps, and potential safeguards solutions. Metallic nuclear fuel types have been around for many decades and were included in some of the first reactors including the Experimental Breeder Reactor (EBR)-I and -II, the Fermi 1 reactor, the Integral Fast Reactor (IFR), and the Dounreay Fast Reactor (DFR). These reactors used various compositions including pure uranium (U) metal, U-zirconium (Zr) alloys, plutonium (Pu)-aluminum (Al) alloys, U-fissium (Fs) alloys, U-Pu-Zr alloys, and U-molybdenum (Mo) alloys [1, 2, 3, 4, 5]. These small alloying additions are included to improve the material properties of the pure U metal. The alpha-phase U (stable below 661C) suffers elongation in one direction causing grain boundary cracking and increasing creep rate due to irradiation growth, thermal cycling, and preferential crystal orientation. It is ideal to utilize the gamma-phase U (typically stable above 769C) by adding small amounts of alloying elements such as Zr or Mo to stabilize this phase down to room temperature [3]. Additionally, some research has been focused on U with transuranic (TRU) elements present, typically coming from the used fuel recycling process. Including these elements in fast reactor fuel can aid in the reduction of nuclear waste by burning minor long-lived actinides. However, the additions of TRU elements can cause concerns to arise when trying to fabrication or safeguard metallic fuels. A typical metallic fuel element is shown in Figure 1. Sodium is added into the cladding to create a thermal bond between the fuel slug and cladding wall. The fuel slug is then inserted and the end plug is welded on to the top of the fuel element. A gas plenum is left to create a headspace for gaseous fission products to escape rather than continue to build in the fuel itself [1, 5]. Other fuel element geometries exist as well, such as the Lightbridge twisted cruciform geometry shown in Figure 2 [6]. This design allows for better cooling performance and provides room for fuel rod swelling without impacting the fuel rod diameter. There are many different fabrication methods for metallic fuels, which is one of the many benefits of these fuel types. Many of these fabrication methods are relatively easy and cost-efficient. The most popular fabrication method is injection casting, sometimes called vacuum induction melting (VIM), shown in Figure 3 [4, 8, 9, 7, 10]. This method was largely used for EBR-II fuel fabrication. The injection casting system is contained inside of a vessel consisting of a Y2O3-coated graphite crucible surrounded by an induction coil with ZrO2-coated quartz molds suspended above the crucible. The fuel feedstock is placed inside of the graphite crucible and melted using the induction furnace. The induction furnace utilizes a dual frequency with the high frequency melting the feedstock and the low frequency causing stirring of the melted feedstock to form a homogeneous mixture. The mixture is heated to approximately 1600C in an argon environment. The vessel is evacuated and then the quartz molds are lowered into the graphite crucible containing the molten metal and the vessel is repressurized to inject the metal fuel upwards into the molds. The molds are removed and then shattered to release the fuel slugs. This fabrication method was used to fabricate 39,000 metallic fuel pins for EBR-II. While injection casting has been the most common metallic fuel fabrication method throughout the decades, many other methods have been explored including low-pressure gravity casting, microwave casting, continuous casting, centrifugal casting, coextrusion, and many others [11, 12, 8, 13, 14, 15]. Some of these methods aim to mitigate challenges that arise with americium (Am) volatilization during the casting process for TRU-containing fuel feedstocks, an issue with injection casting. Coextrusion is one of the methods explored at the Idaho National Laboratory (INL) and has been utilized for the initial fabrication tests of Lightbridge's unique fuels, as well as other metallic fuels with cladding coextruded. In this process, large billets are formed and machined and then inserted into a molten salt bath for approximately 30 minutes. The billets are then loaded into the extrusion press and extruded. This process can be seen in Figure 4 [15].

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Effects of Synthesis Gas Concentration, Composition, and Operational Time on Tubular Solid Oxide Fuel Cell Performance

There is tremendous potential to utilize the exhaust gases and heat already present within combustion chambers to generate electrical power via solid oxide fuel cells (SOFCs). Variations in system design have been investigated as well as thorough examinations into the impacts of environmental conditions and fuel composition/concentration on SOFC performance. In an attempt to isolate the impacts of carbon monoxide and hydrogen concentration ratios within the exhaust stream, this work utilizes multi-temperature performance analyses with simulated methane combustion exhaust as fuel combined with dilute hydrogen baseline tests. These comparisons reveal the impacts of the complex reaction pathways carbon monoxide participates in when used as an SOFC fuel. Despite these complexities, performance reductions as a result of the presence of carbon monoxide are low when compared to similarly dilute hydrogen as a fuel. This provides further motivation for the continued development of SOFC-CHP systems. Stability testing performed over 80 h reveals the need for careful control of the operating environment as well as signs of carbon deposition. As a result of gas flow disruption, impacts of anode oxidation that may normally not hinder power production become significant factors in addition to coarsening of the anode material. Thermal management and strategies to minimize these impacts are a topic of future research.

54 ENVIRONMENTAL SCIENCES↗

Depletion Benchmark of the AFIP-7 Experiment in the Advanced Test Reactor

Reactor physics depletion benchmarks for low-enriched uranium fuel are limited in number. In particular, there is very limited data for LEU benchmarks for U-10Mo (Uranium-10% Molybdenum) plate fuel developed for use in U.S. high-performance research reactors (USHPRR). USHPRR includes the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), University of Missouri Research Reactor (MURR), Massachusetts Institute of Technology Reactor (MITR), and National Bureau of Standards Reactor (NBSR) at the National Institute of Science and Technology. These reactors are fueled with high-enriched uranium dispersed fuel in a silicon/aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiated fueled specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This report documents the development of a benchmark model of the irradiation of the ATR Full -size plate In center flux trap Position 7 (AFIP-7) experiment.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗