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At least 1,207 records · Page 67
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

High-Speed and Dynamic Diaphragm Compressor for High-Capacity Fueling (Final Technical Report)

Goal of the project is to scale the NEL hydrogen compressor technology from 1.1 kg/minute to 5.5 kg/minute capacity (scale 1:5). This will enable achievement of 330 kg/hour capacity, sufficient to achieve fast high-capacity heavy-duty vehicle (HDV) fueling. The technology readiness level (TRL) maturity objective is to develop and test a scaled-up compressor that is to advance from TRL3 to TRL7.

08 HYDROGEN↗
DOE OSTI2020

Experimental assessment of antimony (Sb) in pure uranium for immobilizing fission product lanthanides

The use of fuel additives is one of the concepts to mitigate fuel cladding chemical interaction (FCCI) for metallic fuel because fission product lanthanides are expected to be immobilized by the additive. Antimony (Sb) has been discovered to be a good candidate in UZr fuel. The present study focuses on its mechanism for immobilization in pure uranium, U–4Sb alloy was fabricated to understand the Sb behavior, while U–4Sb–4Ce was fabricated to simulate the case when lanthanides are generated. Both of the as-cast and annealed samples were characterized by scanning electron microscope (SEM) and energy dispersive spectrometer (EDS). U–Sb precipitates are formed in U–4Sb alloy, while U–Sb and Ce–Sb were found in U–4Sb–4Ce alloy, thermal exposure does not change the Sb-precipitation morphologies or chemical composition. Furthermore, diffusion couple tests between those alloys and cladding materials (Fe or HT9) under 650 °C for 500 h were performed and analyzed using SEM/EDS. Diffusion couple tests demonstrate the reaction between cladding and uranium, while no reaction between Sb-precipitations and cladding materials was found.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Ducted Fuel Injection Provides Consistently Lower Soot Emissions in Sweep to Full-Load Conditions

Earlier studies have proven how ducted fuel injection (DFI) substantially reduces soot for low- and mid-load conditions in heavy-duty engines, without significant adverse effects on other emissions. Nevertheless, no comprehensive DFI study exists showing soot reductions at high- and full-load conditions. This study investigated DFI in a single-cylinder, 1.7-L, optical engine from low- to full-load conditions with a low-net-carbon fuel consisting of 80% renewable diesel and 20% biodiesel. Over the tested load range, DFI reduced engine-out soot by 38.1–63.1% compared to conventional diesel combustion (CDC). This soot reduction occurred without significant detrimental effects on other emission types. Thus, DFI reduced the severity of the soot–NO x tradeoff at all tested conditions. While DFI delivered considerable soot reductions in thepresent study, previous DFI studies at low- and mid-load conditions delivered larger soot reductions (>90%) compared to CDC operation at the same conditions. Therefore, the DFI configuration used here has been deemed nonoptimal (in terms of parameters such as the injector-spray and piston geometries), and several improvements are recommended for future studies with high-load DFI. Further, these improvements include employing better spray-duct alignment, a deeper piston bowl with a smaller injector umbrella angle, and a fuel injector that opens and closes faster. The study also suggests future research to make DFI ready for commercialization, such as metal-engine tests to ensure desirable DFI performance over an engine’s complete speed/load map. Overall, this study supports the continued development and commercialization of DFI to meet upcoming emissions regulations for heavy-duty vehicles. Specifically, multicylinder engine experiments and CFD simulations should be utilized to optimize the performance and clarify the full potential of DFI.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Single-Bar SMT Creep-Fatigue Testing on Alloy 617 with Software Controls on elastic Follow-Up Feedback in Support of New Creep-Fatigue Design Methodology

The Simplified Model Test (SMT) is one of the alternative ways to calculate the creep-fatigue damage of elevated temperature components. The SMT approach unifies the creep and fatigue damage, simplifying the damage calculation process with improved accuracy. Traditionally, SMT tests has have been conducted through the two bars: first is the driver bar and second is the test bar. The driver bar stays elastic and imposes the confinement to the second bar. This experimental procedure requires two test frames and a large test specimen, which limits the test parameter range and accessibility of this testing procedures. Hence, a single-bar SMT (SB-SMT) tests has been developed to simplify complexity in SMT experiment setup. The software-controlled SB-SMT test process has been introduced which completely replaces the driver bar. This paper discusses the single-bar SMT (SB-SMT) test procedure with software controls. The challenges and critical parameters in the software-controlled SB-SMT procedure are addresses addressed and recommendations are provided. A scoping test with a set of wider strain range, elastic follow-ups and dwell time validated the proposed test procedure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Tantalum Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.02 cm Pitch)

The United States (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 (SNL). 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, LEU- COMP-THERM-097, LEU-COMP-THERM-101, LEU-COMP-THERM-102, and LEU-COMP-THERM-111. Applied interests in neutron-absorbing elements at epithermal and intermediate neutron energy ranges at DOE’s Hanford Site and Savannah River Site were identified through DOE Office of Environmental Management (EM) workshops and needs surveys performed in conjunction with the National Nuclear Security Administration/Nuclear Criticality Safety Program. In response, a team lead by Oak Ridge National Laboratory (ORNL) performed a study evaluating the use of 7uPCX for testing the epithermal cross sections of materials at SNL. This marked the inception of the set of experiments described here. The purpose of these experiments was to measure the effects of tantalum in nearly-critical systems. The tantalum was introduced into the fuel arrays as experiment rods within a central test region. The central test region was designed to target the epithermal neutron energy range by providing a dry cylindrical cavity constructed of aluminum and lined with 0.04 in (0.1016 cm) thick cadmium sheet. The central test region has an outer diameter of 3.75 in (9.525 cm) length of 31 in (78.74 cm) and can hold 85 tantalum rods nominally 0.25 in (0.635 cm) outside diameter and 31.25 in (79.375 cm) long. 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 2016 fuel rod positions on a pitch of 0.4 in (1.016 cm) in a series of 21 hexagonal rings positioned around the central test region. 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 eight cases in this experiment series were performed in 2024 in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility by varying the number of fuel rods in the critical assembly. The personnel involved in conducting these experiments included David Ames, Elijah Lutz, Beth Hanson, Patrick Ward, Jason Soares, and Augie Chapa from SNL, as well as Mathieu Dupont and B.J. Marshall from ORNL. Case 1 had no tantalum experiment rods. Case 2 had seven tantalum rods in the center of the central test region. Case 3 had 18 tantalum rods in the third hexagonal ring of the central test region. Case 4 had 19 tantalum rods in the center of the central test region. Case 5 had 30 tantalum rods in the outer most positions of the central test region. Case 6 had 37 tantalum rods in the center of the central test region. Case 7 had 61 tantalum rods in the center of the central test region. Case 8 had 85 tantalum rods filling all positions in the central test region. All eight critical experiments are judged to be acceptable as benchmark experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Neutronic Safety Analysis of Pu-238 Production at Idaho National Laboratory

This analysis was completed to support the irradiation of plutonium fuel services (PFS) targets in the NEFT in the Advanced Test Reactor (ATR) as a part of the campaign to restart domestic production of plutonium-238 used in radioisotope power systems (RPS) by the National Aeronautical and Space Administration (NASA) and Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Nuclear Infrastructure Program (NE-3). Referred to as the PFS-ATR-GEN1-NEFT experiment, the assembly was designed to hold 46 PFS targets in the NEFT. The scope of this paper is to outline the MOPY method used to calculate the heat generation rates (HGRs), flux, fission density, and quantify the viability of the target design for Pu-238 production in ATR.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Integrated simulation of U-10Mo monolithic fuel swelling behavior

Here, a separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to take into account various processes occurring in RTR fuels during irradiation. In order to accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions vs. fission density) predicted using the phase-field method. The remainder of fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed in order to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterizing the seismic response of a molten salt nuclear reactor

Abstract One Generation IV nuclear reactor, which uses a fluoride salt as a coolant, graphite reflector blocks as a moderator, and circulating buoyant TRISO pebbles as fuel is at an advanced stage of development. To characterize the seismic behavior of components of this reactor, validate numerical models for analysis, and develop recommendations for design, a set of earthquake‐simulator experiments on a scaled model of the reactor vessel and its internals was executed on a six‐degree‐of‐freedom earthquake simulator. The model was seismically isolated at its base using two types of spherical sliding bearings. The scaled model involved representations of the prototype reactor vessel, core barrel, reflector blocks, coolant, and spherical fuel pebbles. The material and geometric properties of different test components were selected based on a dynamic similitude scaling analysis and an approximate length scale of 0.4. Four sets of three‐component earthquake motions were used as inputs for testing. Instrumentation on the test specimen recorded the dynamic responses of the outer vessel, core barrel, and reflector‐block assembly, the hydrodynamic responses (sloshing and hydrodynamic pressure) of the liquid coolant, pebble consolidation under earthquake shaking, and the behavior of the isolation systems. This paper describes the design of the experiments and presents key results from the tests. The dynamic responses of the outer vessel, core barrel, and the reflector blocks revealed that the components responded as a unit for the intense shaking used in the experiments. The sloshing response of the fluid in a thin annulus near the perimeter of the vessel was heavily damped. The change in the packing fraction of the pebble bed under repeated, intense 3D earthquake shaking was less than 3%. Seismically isolating the vessel substantially reduced demands on its internal components.

Engineering↗
DOE OSTI2026

RuKY Catalyst‐Packed Permeation Membrane for Quantitative Ammonia and d3‐Ammonia Dehydrogenation to Ultrapure Hydrogen

Ammonia is a promising carbon-free hydrogen carrier, but incomplete ammonia dehydrogenation (cracking) generates atmospheric emissions of NO x , a potent greenhouse gas. Additionally, incomplete cracking of ammonia leads to regulatory challenges in nuclear and fusion power, where tritiated ammonia (NT 3 ) emissions are strictly controlled. Therefore, we report the use of low-temperature ammonia dehydrogenation catalysts (3%Ru/1%Y/12%K/Al 2 O 3 ) in a palladium alloy H 2 permeation membrane for quantitative conversion of ammonia into hydrogen and nitrogen at industry-relevant conditions. This catalytic membrane reactor system achieved an astonishing effluent concentration of <1 ppm at 450°C under a 100% NH 3 stream, which is far beyond the 99.6% conversion target required for the adoption of ammonia as a vehicle fuel. The low-temperature ammonia dehydrogenation catalyst was tested in a packed bed reactor with NH 3 and ND 3 to both elucidate the reaction mechanism and to quantify the kinetic isotope effect of the membrane reactor. The rate-limiting step at temperatures relevant to the palladium membrane are isotope independent, indicating that the isotopologue content will not modify the desired reaction kinetics. By reducing emissions to below-trace levels with no additional separation, this work provides a path to greatly simplified and miniaturized ammonia cracking processes.

ammonia decomposition↗
DOE OSTI2023

Multifidelity uncertainty quantification with models based on dissimilar parameters

Multifidelity uncertainty quantification (MF UQ) sampling approaches have been shown to significantly reduce the variance of statistical estimators while preserving the bias of the highest-fidelity model, provided that the low-fidelity models are well correlated. However, maintaining a high level of correlation can be challenging, especially when models depend on different input uncertain parameters, which drastically reduces the correlation. Existing MF UQ approaches do not adequately address this issue. In this work, we propose a new sampling strategy that exploits a shared space to improve the correlation among models with dissimilar parameterization. We achieve this by transforming the original coordinates onto an auxiliary manifold using the adaptive basis (AB) method (Tipireddy and Ghanem, 2014). The AB method has two main benefits: (1) it provides an effective tool to identify the low-dimensional manifold on which each model can be represented, and (2) it enables easy transformation of polynomial chaos representations from high- to low-dimensional spaces. This latter feature is used to identify a shared manifold among models without requiring additional evaluations. Here we present two algorithmic flavors of the new estimator to cover different analysis scenarios, including those with legacy and non-legacy high-fidelity (HF) data. We provide numerical results for analytical examples, a direct field acoustic test, and a finite element model of a nuclear fuel assembly. For all examples, we compare the proposed strategy against both single-fidelity and MF estimators based on the original model parameterization.

42 ENGINEERING↗
DOE OSTI2022

Simulation of natural circulation cartridge loop experiments and application to molten salt reactors

This work uses the TRAC/RELAP Advanced Computational Engine (TRACE) thermal hydraulics (TH) code to model natural circulation cartridge loop experiments previously conducted at Oak Ridge National Laboratory (ORNL) using water and compares the simulated and experimental results. TRACE is also used to characterize natural circulation in the cartridge loop vehicle using FLiNaK as the working fluid. The experimental vehicle is a buoyancy-aided, annular cartridge loop, referred to as a thermosyphon, and is designed to aid in qualifying liquid–fueled and/or liquid–cooled irradiation experiments for the Versatile Test Reactor (VTR), which is currently being designed in the United States. Out-of-pile water experiments have been conducted using the cartridge and the Thermosyphon Test Loop facility at ORNL, and future experiments are anticipated that would use other molten salt surrogates as the working fluid, followed by eventual insertion of a similar cartridge into VTR. Additionally, this work aims to determine how well TRACE can replicate the natural convection conditions that were observed experimentally; this serves as an initial step for validating the modeling tool for design and safety calculations to support future irradiation experiments in VTR. Initial predictions of potential experiments were made using FLiNaK as the natural circulation fluid to demonstrate the relevance of the cartridge design to molten salt reactors (MSRs). Results from this study indicate that TRACE can accurately capture natural convection phenomena in the thermosyphon and that several design changes to the current cartridge vehicle are necessary to achieve hydraulic conditions similar to those expected in MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Extraction, Characterization, and Stability Studies of Bistriazinyl-Derived Carboxylic Acids

Separation of An(III) and Ln(III) ions will benefit the recycling of used nuclear fuel (UNF). For this purpose, many ligands have been tested over the years, and several separation processes have been successfully demonstrated on the laboratory scale. Current research aims at the development of new ligands that are built only with carbon, hydrogen, oxygen, and nitrogen (CHON), as they can be incinerated completely without secondary waste production. Here, we tested a new class of water-soluble ligands, the bistriazinyl-octa-carboxylic acids. One member, in particular, 2,6-bis-[5,6-di(3,4-dicarboxyphenyl)-1,2,4- triazin-3-yl]-pyridine (BTPOA), was found to be suitable for the selective separation of Am(III) and Cm(III) ions from Ln(III) ions and may act as a CHON alternative to its sulfonated analogue (SO 3 -Ph-BTP). BTPOA exhibited good extraction results and a high selectivity for Am(III) over Eu(III) ions. This ligand’s complexation of metal ions was further studied using potentiometric spectroscopy, as well as time-resolved laser-induced spectroscopy with Cm(III) and Eu(III) in aqueous HClO 4 and HNO 3 media. Conditional stability constants of each formed species were determined. In the HClO 4 system, Cm(III) formed three species (1:1, 1:2, and 1:3) through the stepwise addition of a single BTPOA molecule. On the other hand, in HNO 3 , Cm(III) formed two 1:2 complexes and one 1:3 complex, while the stepwise formation of three species was observed for Eu(III). The stability constants are comparable to the values for SO 3 -Ph-BTP. The radiolytic behavior of BTPOA was also investigated using electron pulse irradiation measurements to determine absolute rate coefficients (k) under ambient temperature conditions for the reaction of BTPOA with typical UNF reprocessing radical radiolysis products the hydrated electron (e aq − , k = (1.60 ± 0.02) × 10 10 M −1 s −1 ), the hydrogen atom (H • , k = (2.17 ± 0.03) × 10 9 M −1 s −1 ), and hydroxyl ( • OH, k = (6.95 ± 0.06) × 10 9 M −1 s −1 ) and nitrate (NO 3 • , k = (0.37 ± 0.02) × 10 7 M −1 s−1) radicals. These rate coefficients indicate that the radiolytic longevity of BTPOA should increase with HNO 3 concentration, owing to the consumption of e aq − /H • , by nitrate anions, and the replacement of • OH by the less reactive NO 3 • .

Diaz Gomez, Laura [Forschungszentrum Juelich (Germ↗
DOE OSTI2020

Sulfur Pellets Responses to a Bare and Steel Reflected Pulse of the Oak Ridge National Laboratory Health Physics Research Reactor

The experiments analyzed in this report were conducted at the Health Physics Research Reactor (HPRR), also known as the $\textit{Fast Burst Reactor}$. The reactor was designed and built at Oak Ridge National Laboratory (ORNL) in 1961. The HPRR was an unmoderated, unshielded fast reactor that used highly enriched uranium and molybdenum alloy as fuel. The reactor was initially sent to the Nevada Test Site in 1962, where it was used to evaluate radiation doses received as a result of the Hiroshima and Nagasaki bombings during World War II. A few years later, the reactor was sent back to ORNL to be part of the Dosimetry Application Research (DOSAR) facility shown in Figure 1, which included a reactor building shown on the left (west) of the picture and a control and laboratory building in the upper right corner (northeast). The critical assembly was used for numerous technical studies, including systems calibration, dosimetry, radiobiology of plants and animals, testing of radiation alarms, as well as teaching and training in radiation dosimetry and nuclear engineering. Between 1963 and 1987, the HPRR was operated for thousands of hours, achieving criticality close to 10,000 times and motivating many publications. The HPRR was decommissioned in 1987. The goal of this effort was to use historical data from operation of the HPRR to create a criticality accident alarm system (CAAS) benchmark to be included in the $\textit{International Handbook of Evaluated Criticality Safety Benchmark Experiments}$ (ICSBEP Handbook). A thorough inspection was performed of all available documentation and information available. The most promising experiments that were selected for evaluation were those described in the 1987 ORNL report entitled $\textit{Health Physics Research Reactor Reference Dosimetry}$, ORNL-6240. The report includes reference dosimetry results of the shielded and unshielded configurations of the HPRR after burst operations. Because of changes to the reactor positioning and storage systems that were made in 1985, the previous dosimetry reports became obsolete, and the newly designed experiments were needed to create the HPRR’s adjusted dosimetry data. The various results reported in ORNL-6240 include reference doses and dose equivalents from different conventions at different distances and elevations as determined using the detected neutron fluence and conversion factors. The HPRR neutron fluence was obtained through different methods, including sulfur pellet analysis and threshold detector unit data. Information about the HPRR spectrum was also obtained through Bonner sphere measurements. This benchmark is focused on a part of the measured sulfur fluences reported in Appendix H of ORNL-6240. Standard commercial sulfur pellets were placed at different distances from the HPRR centerline during burst operation and were activated due to the 32 S(n,p) 32 P reaction. The resulting 32 P activity was then measured and the information about the corresponding sulfur fluence and/or neutron dose could be extracted. Many of those measurements have 7 been performed with the HPRR in its bare configuration or with different shields (combinations of Lucite, concrete, steel). All the necessary, precise information about material and/or dimensions of the different shields was not found, so it was decided to focus only on the unshielded and steel-shielded configurations to minimize the benchmark uncertainty. A total of 31 cases (24 unshielded and 7 shielded cases at different positions) of sulfur fluence were selected before evaluation to develop the benchmark.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Verification of Spent Fuel Inside Dry Storage Casks by Cask Top Fast Neutron Mapping (FY2023 Mid-Year Report)

This project is developing a prototype scanner array verification system for detection of missing fuel assemblies in spent-fuel storage casks. The prototype consists of six fast-neutron scintillator detectors mounted to a linear actuator frame that is placed on the top of a spent fuel cask to scan across all fuel assembly positions. The scanner array was assembled and tested at LLNL in FY2022. A field test schedule has been requested at the Idaho National Laboratory (INL) Cask Farm site for FY2023. Note that the Cask Farm contractor determines this scheduling and not INL directly. Further system automation will be designed and implemented with the goal of obtaining a level of system operation that meets IAEA needs. This includes integration of the scanner array and data-acquisition control software into a single interface for operator use. In addition, commercial operators and the IAEA may have special requirements for portability, shipping, lifting, and installation. Prior to the Field Test at INL, the system will be operated at LLNL to exercise lifting procedure and linear actuators, monitor stability of detector energy and pulse-shape discrimination calibration, and test system software integration efforts. Following the Field Test, we will present results and discuss the technology with the IAEA. We will incorporate additional improvements to the system based on lessons learned from the field test and feedback from the IAEA. If successful, the technology can be transferred to the IAEA or other stakeholders for assessment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Sludge Batch 11 Assembly: Tank 15

Savannah River Mission Completion Nuclear Safety and Engineering Integration (SRMC-E) has requested that Savannah River National Laboratory (SRNL) perform Tank 15 characterization analyses in support of Sludge Batch 11 (SB11) assembly. This report provides important characterization of the slurry in Tank 15 prior to transfer from Tank 15 to Tank 13 (prior to sending to Tank 51) in order to demonstrate that the transfer is "Low Rem" and ensure the sludge concurs with the estimated transfer mass for the SB11 recipe. A total of 3 sets of Tank 15 samples were delivered to SRNL from December 2022 to August 2023. The individual sets of composite samples were analyzed for density, weight percent solids, chemical composition, radionuclides, and/or supernate corrosion control tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Dual-pump Coherent anti-Stokes Raman scattering measurements in H 2 /CH 4 Counterflow Flames

Temperature and major species concentration measurements are reported in a Williams Seshadri counterflow burner using dual-pump coherent anti-Stokes Raman scattering (DPCARS). Hencken burner H2/Air flames with equivalence ratio of 0.8-1.4 were used to characterize the accuracy of the system. The mean errors for temperature, XH2/XN2 mole fraction ratio, and XO2/XN2 mole fraction ratio are 1.0%, 14.7%, and 5.5%, respectively. The counterflow burner uses a CH4, H2, and N2 mixture as fuel and synthetic air as oxidizer. Measurements were performed along the counterflow burner central axis for two different fuel compositions: 4% H2/86% CH4/10% N2 and 87%H2/3% CH4/10% N2. The measured temperature and species mole fractions agree well with the 1-D opposed-flow diffusion flame simulation. One observation is that the fuel inlet temperature is higher than 300 K, which suggests the possibility of the stagnation plane residing closer to the fuel side. Lastly, these measurements provide valuable data for testing and validation of advanced 2-D flame models.

Murdock, Benjamin K.↗
DOE OSTI2025

Ceramography and Thermal Modelling of Irradiated TRISO Particles using BISON

This study presents a comprehensive analysis of irradiated TRISO fuel particles from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program conducted at the Advanced Test Reactor (ATR) as part of my internship. Failure Analysis using high-resolution microscopy was performed to characterize mechanisms in TRISO particle layers, including the kernel, buffer, inner pyrolytic carbon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon (OPyC) coatings. Failure modes were systematically classified. Thermal conductivity properties of each TRISO layer were evaluated through experimental measurements and computational models for use in BISON finite element code. Experimental thermal conductivity values ranged from 0.5 W/m·K for the buffer layer to 168 W/m·K for SiC, while BISON simulations provided averaged values of 0.57 W/m·K (buffer), 4.0 W/m·K (PyC layers), and 15.95 W/m·K (SiC). Temperature-dependent correlations from PARFUME were implemented for kernel thermal properties.

BISON↗