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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

NREL Hydrogen Sensor Testing Laboratory

The NREL Sensor Laboratory supports the HFTO's mission for the safe and efficient implementation of hydrogen as an energy carrier. Sensors are a critical element for a facility safety system by providing an early indication of unintended releases. Sensors can be used in process control applications (e.g., FQ verification). The Sensor Laboratory provides developers and end-users with an unbiased resource to validate sensor performance to application requirements. The Sensor Laboratory facilitates deployment through support of codes and standards development and safety research. The Sensor Laboratory supports advanced strategies for early leak detection.

detection↗
DOE OSTI2022

Implementation and testing of the GPT sensitivities in TRIPOLI-4

Computing the sensitivity of given reactor parameters with respect to nuclear data is mandatory in key applications such as the assessment of reactor safety or core design. In recent years, the development of Monte Carlo methods capable of estimating sensitivity coefficients based on the Standard or Generalized Perturbation Theory (SPT and GPT, respectively) has attracted intensive research efforts, especially in view of having reference golden standard simulations to be compared to faster but approximate deterministic calculations. Following our previous implementation and validation of the SPT functionalities in the Monte Carlo TRIPOLI-4, devoted to the sensitivity of the k eigenvalue, in this work we detail the development of GPT methods for ratios of reaction rates, and validate them against a few relevant benchmark results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Development of a forced advection sampling technique (FAST) for quantification of methane emissions from orphaned wells

Abstract. Orphaned wells, meaning unplugged and non-producing wells lacking responsible owners, pose a significant and undersampled environmental challenge due to their vast number and unknown associated emissions. We propose, develop and test an alternative method for estimating emissions from orphaned wells using a forced advection sampling technique (FAST) that can overcome many of the limitations in current methods (cost, accuracy, safety). In contrast to existing ambient Gaussian plume methods, our approach uses a fan-generated flow to force advection between the emission source and a point methane (CH4) sensor. The fan flow field is characterized using a colocated sonic anemometer to measure the 3D wind profile generated by the fan. Using time-series measurements of CH4 concentration and wind, a simple estimate of the CH4 emission rate of the source can be inferred. The method was calibrated using outdoor controlled-release experiments and then tested on four orphaned wells in Lufkin, TX, and Osage County, OK. Our results suggest that the FAST method can provide a low-cost, portable, fast and safe alternative to existing methods with reasonable estimates of orphaned well emissions over a range of leak rates below 40 g h−1 and within certain geometric and atmospheric constraints.

Dubey, Mohit L↗
DOE OSTI2023

Multi-Scale Characterization of Porosity and Cracks in Silicon Carbide Cladding after Transient Reactor Test Facility Irradiation

Silicon carbide (SiC) ceramic matrix composite (CMC) cladding is currently being pursued as one of the leading candidates for accident-tolerant fuel (ATF) cladding for light water reactor applications. The morphology of fabrication defects, including the size and shape of voids, is one of the key challenges that impacts cladding performance and guarantees reactor safety. Therefore, quantification of defects’ size, location, distribution, and leak paths is critical to determining SiC CMC in-core performance. This research aims to provide quantitative insight into the defect’s distribution under multi-scale characterization at different length scales before and after different Transient Reactor Test Facility (TREAT) irradiation tests. A non-destructive multi-scale evaluation of irradiated SiC will help to assess critical microstructural defects from production and/or experimental testing to better understand and predict overall cladding performance. X-ray computed tomography (XCT), a non-destructive, data-rich characterization technique, is combined with lower length scale electronic microscopic characterization, which provides microscale morphology and structural characterization. This paper discusses a fully automatic workflow to detect and analyze SiC-SiC defects using image processing techniques on 3D X-ray images. Following the XCT data analysis, advanced characterizations from focused ion beam (FIB) and transmission electron microscopy (TEM) were conducted to verify the findings from the XCT data, especially quantitative results from local nano-scale TEM 3D tomography data, which were utilized to complement the 3D XCT results. In this work, three SiC samples (two irradiated and one unirradiated) provided by General Atomics are investigated. The irradiated samples were irradiated in a way that was expected to induce cracking, and indeed, the automated workflow developed in this work was able to successfully identify and characterize the defects formation in the irradiated samples while detecting no observed cracking in the unirradiated sample. These results demonstrate the value of automated XCT tools to better understand the damage and damage propagation in SiC-SiC structures for nuclear applications.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Uncertainty Quantification Enabled by Automatic Differentiation for Hydrodynamic Simulation of Shock‐to‐Detonation Transition in High Explosives

Quantifying the effects of uncertainty in a reactive burn model on the run-to-detonation time in high explosives (HEs) provides a robust methodology for assessing the probability of an HE failing the IHE qualification standard. Moreover, uncertainty quantification helps evaluate whether the model calibration accurately represents data outside the calibration set. This study uses a specialized hydrodynamic simulation code for modeling detonation to determine the run-to-detonation time of the HE PBX 9502 for various impact velocities. To quickly approximate uncertainties in the model, a surrogate was constructed using a Taylor series expansion centered at the mean of the input parameters. To obtain the sensitivities required for constructing the Taylor series, HYP-percomplex Automatic Differentiation (HYPAD) was implemented. HYPAD is a methodology for infusing existing codes with automatic differentiation capabilities by augmenting variables with one or more imaginary units to compute step-size independent partial derivatives. These derivatives are accurate to machine precision with respect to the implemented numerical algorithm, meaning their accuracy reflects that of the underlying method (e.g., integration or discretization schemes). Using reduced order modeling techniques, the mean and standard deviation of the run-to-detonation time of a shock within PBX 9502 were computed for a number of initial impact velocities. A weighted least squares regression was then performed to obtain a best fit curve and prediction interval for the computed statistics. Historical data points from explosively driven wedge tests were utilized to validate the prediction interval, ensuring its reliability in predicting future outcomes. With this prediction interval and a known safety constraint curve, the most probable point of failure and the probability of failure for the HE PBX 9502 were determined.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Design of separate-effects In-Pile transient boiling experiments at the $\mathrm{TREAT}$ Facility

The cladding-to-coolant heat transfer during rapid power excursions, such as reactivity-initiated accidents, remains a crucial area of uncertainty in nuclear reactor safety. This uncertainty impacts the ability to accurately predict fuel performance behavior for these conditions. Improving our understanding of transient cladding-to-coolant heat transfer will enhance our ability to model design basis accidents and could increase design and improve safety margins for the current commercial fleet and advanced reactors. To address these issues, the Critical Heat Flux Static Environment Rodlet Transient Test Apparatus (CHF-SERTTA) experiment uses a novel approach with a borated stainless-steel heater rodlet that can replicate cladding temperatures experienced during a design-basis reactivity-initiated accident. The rodlet and experiment is instrumented to provide temperature and thermal-hydraulic conditions throughout the transient. The design of the rodlet allows for separate effect testing that eliminates complexities with a fuel and cladding specimen. We report the experiments have provided data that will be used to improve our understanding of boiling behavior, specifically critical heat flux, under transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Systemwide Planning with a Branch-and-Price Algorithm for Pavement-Marking Assessment Data Collection via the Mobile Retroreflectivity Unit Routing Model

The visibility of pavement markings is one of the most critical factors for traffic safety, and a periodical assessment plan is crucial for maintaining this function. Traditional assessment methods, such as visual windshield surveys or manual testing using handheld devices, are unsafe, time-consuming, and labor-intensive. In recent years, transportation agencies have begun to adopt the use of mobile retroreflectivity units (MRUs) for condition assessment of pavement markings. MRUs, different from other manual methods, can be utilized to collect large-scale retroreflectivity data in an efficient manner. However, no relevant research has yet proposed a mathematical optimization model for arranging the evaluation schedule and paths of MRUs. This study aims to propose a MRU routing model, and an efficient solution methodology. A branch-and-price algorithm, including column generation and branch-and-bound, was implemented. Computational experiments have been conducted based on actual tasks from the Florida MRU program for validation. In conclusion, results show that the proposed solution methodology with a set partitioning model in this study not only finds the optimal solution for problems with tasks less than 60, but also effectively narrows the solution gap to be within 1.0% for problems with tasks less than 131.

42 ENGINEERING↗
DOE OSTI2020

Progress in extending high poloidal beta scenarios on DIII-D towards a steady-state fusion reactor and impact of energetic particles

To prepare for steady-state operation of future fusion reactors (e.g. the International Thermonuclear Experimental Reactor and China Fusion Engineering Test Reactor (CFETR)), experiments on DIII-D have extended the high poloidal beta ( β P ) scenario to reactor-relevant edge safety factor q 95 ~ 6.0, while maintaining a large-radius internal transport barrier (ITB) using negative magnetic shear. Excellent energy confinement quality ( H 98y2 > 1.5) is sustained at high normalized beta ( β N ~ 3.5). This high-performance ITB state with Greenwald density fraction near 100% and q min ≥ 3 is achieved with toroidal plasma rotation V tor ~ 0 at ρ ≥ 0.6. This is a key result for reactors expected to have low V tor . At high β P (≥1.9), large Shafranov shift can stabilize turbulence leading to a high confinement state with a low pedestal and an ITB. At lower β P (<1.9), negative magnetic shear in the plasma core contributes to turbulence suppression and can compensate for reduced Shafranov shift to continue to access a large-radius ITB and excellent confinement with low V tor , consistent with the results of gyrofluid transport simulations. These high- β P cases are characterized by weak/no Alfvén eigenmodes (a.e.) and classical fast-ion transport. At high density, the fast-ion deceleration time decreases and Δ β fast is lower; these reduce a.e. drive. The reverse-shear Alfvén eigenmodes are weaker or stable because the negative magnetic shear region is located at higher radius, away from the peaked fast-ion profile. Resistive wall modes can be a limitation at simultaneous high β N , low internal inductance, and low rotation. Analysis suggests that additional off-axis external current drive could provide a more stable path at reduced q 95 . Based on a DIII-D high- β P plasma with large-radius ITB, two scenarios are proposed for CFETR Q = 5 steady-state operation with ~1 GW fusion power: a lower-$l_i$($l_i$ ~ 0.66) and a higher-$l_i$($l_i$ ~ 0.75) case. Using a Landau closure model, multiple energetic particle (EP) effects on the a.e. stability are analyzed modifying the growth rate of the a.e.s triggered by the neutral-beam-injection EPs and alpha particles, although the stabilizing/destabilizing effect is weak for the cases analyzed. The stabilizing effects of the combined EP species β , energy, and density profile in CFETR need further investigation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Modeling immunity in microphysiological systems

There is a need for better predictive models of the human immune system to evaluate safety and efficacy of immunomodulatory drugs and biologics for successful product development and regulatory approvals. Current in vitro models, which are often tested in two-dimensional (2D) tissue culture polystyrene, and preclinical animal models fail to fully recapitulate the function and physiology of the human immune system. Microphysiological systems (MPSs) that can model key microenvironment cues of the human immune system, as well as of specific organs and tissues, may be able to recapitulate specific features of the in vivo inflammatory response. This minireview provides an overview of MPS for modeling lymphatic tissues, immunity at tissue interfaces, inflammatory diseases, and the inflammatory tumor microenvironment in vitro and ex vivo. Broadly, these systems have utility in modeling how certain immunotherapies function in vivo, how dysfunctional immune responses can propagate diseases, and how our immune system can combat pathogens.

Research & Experimental Medicine↗
DOE OSTI2020

Status Report on ASME Code Development for Nonmetallic Core Components in 2020

The purpose of this report is to provide a status update on the progress and ongoing activities of the current ASME Boiler Pressure Vessel (BPV) Code Section III, Division 5, on nonmetallic core components and assemblies which includes the design rules for graphite and ceramic composite materials. Section III is concerned with the design and construction of nuclear reactor components and Division 5 focuses on high temperature reactors (HTR). The ASME Section III Committee’s function for nuclear items other than pressure-retaining components, is to establish rules of safety related to structural integrity. The term “construction” refers to the all-inclusive effort comprising of materials, design, fabrication, examination, inspection, testing, certification, and pressure relief. The Department of Energy (DOE) supports industry codes and standards development through focused research providing the technical basis for new or modified codes and standards and the participation of subject matter experts in codes and standards committees. The activities for the nonmetallic core components and assemblies committees are the pursuit of two working groups that meet quarterly during ASME BPV Code Week meetings. The two working groups are General Requirements for Graphite and Ceramic Composite Core Components and Assemblies (GR GCCCCA) and Nonmetallic Design and Materials (NDM). The rules on general requirements for nonmetallic core components are discussed in the articles of Subsection HA subpart B. Ceramic composites were recently incorporated within the rules, and the GR GCCCCA working group is now in the process of aligning the requirements with the subsections and articles of ASME BPV Code Section III NCA, the general requirements for Division 1 and 2 of the code. The NDM working group is concerned with the design rules for graphite and composite materials for core components and assemblies discussed in Subsection HH (Class SN Nonmetallic Core Components) subpart A and subpart B which reference graphite and composite materials respectively. The bulk of the progress and work is done within this working group. A significant undertaking, with discussion that started in 2015, is to gain endorsement of ASME Section III, Division 5, by the U.S. Nuclear Regulatory Commission (NRC). In June 2018, ASME formally recommended endorsement to the NRC. The NRC has since initiated the endorsement review process, which is to be concluded early in 2021. ASME task groups on metallic structures and components, as well as non-metallic core support structures, have been formed to define potential pathways and schedules for NRC endorsement of Division 5. NRC determined to perform a review on the 2017 edition of Section III Division 5 as a baseline. As a result, the use of graphite as nonmetallic core components formed part of the review, but it excluded composite materials (which were first introduced in the 2019 edition). This report documents progress in FY 2020 directed toward ASME’s efforts to develop design codes for graphite and composites. Progress includes a major code review by NRC, the first code enquiry, and its solution, an overhaul of the treatment of Weibull Statistics, and the addition of Nonmandatory Appendices to the composites code relating to carbon-carbon composite materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Empirical Formula for Puncture Energy of Flat Metal Plates by a Cylindrical Flat Punch

Impact problems of plate-like parts by punch-like objects with relatively large mass moving at slow speeds of a few feet per second constitute a subset of impact problems of interest at Sandia. This is in contrast to small objects moving in the range of hundreds or thousands of feet per second or higher. The objective of this work is to develop a simple formula that can be used to estimate a lower bound for the puncture energy of metal plates impacted by cylindrical, essentially rigid punches of circular cross-section and at nose. Such geometry is used as a basis in the design of puncture mitigation barriers or procedures. This was accomplished by deriving an expression using non-dimensional analysis and then calibrating it based on tests results in the range of speeds of interest. Lower bounds can then be determined based on confidence intervals or factors of safety.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Considerations for Fire Service Response to Residential Battery Energy Storage System Incidents

Renewable energy sources (e.g., rooftop photovoltaics, wind turbines) are capable of addressing our grid-level energy challenges by reducing environmental impacts, increasing resiliency, and increasing the supply of input energy. However, renewable energy sources alone are not a complete solution. The generation of power from major renewable energy sources fluctuates with the weather, creating significant challenges in matching electrical power generation to electrical power consumption. The value of renewable power generation sources is multiplied when paired with battery energy storage systems (ESS), which can store excess power generated from renewable energy sources when it is not consumed and deliver that power later when demand exceeds base power generation. Residential ESS units are frequently installed to support renewable energy initiatives by storing energy from intermittent sources such as photovoltaic panels. Residential ESS units also offer an alternative to gas-powered generators or other backup power means. Lithium-ion batteries are the most common residential ESS technology due to their affordability and energy density. Though lithium-ion systems come with benefits, there is also a risk of thermal runaway within this technology which can result in flammable gas release, fire, and explosion. As the installations of these products increase, the frequency of response to fire incidents involving these products will increase. In response to this new and evolving hazard, UL Solutions (ULS) and UL Fire Safety Research Institute (FSRI) have partnered with the International Association of Fire Fighters (IAFF) to conduct a series of large-scale tests sponsored by the US Department of Energy to characterize the challenges for fire fighters responding to fires involving residential energy storage systems. The project focuses on developing size-up and tactical considerations to support the fire service in navigating the evolving modern fireground.

25 ENERGY STORAGE↗
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 OSTI2025

A Novel Manufacturing Process of Lightweight Automotive Seats (Integration of Additive Manufacturing and Reinforced Polymer Composite)

Lightweight automotive seats offer multiple benefits to original equipment manufacturers in terms of cost savings from various aspects, including less material usage, more integrated processes, and compliance with Corporate Average Fuel Economy Standards. Original equipment manufacturers have been focusing on innovative ways to produce light weight automotive seats. The commercially available automotive seats are currently made of multiple metal components combined through welding and fasteners. The use of additive manufacturing and composite structures is particularly useful for light weighting the automotive components. Additive manufacturing (AM) offers multiple advantages over traditional manufacturing processes such as freedom of design thereby enabling complex structural geometries, mass customization and waste minimization, and control over the fiber alignment through deposition in a predetermined pattern. Combining metal inserts with polymer composites through a novel manufacturing process allows design of lightweight and high-performance materials for automotive components. However, fabricating these metal polymer composite structures through traditional manufacturing processes limits their mechanical properties due to limited design freedom, lack of control over fiber orientation in composite parts, and poor interfacial bonding between the constituent materials. It is essential to develop a novel manufacturing process to enable high throughput production of lightweight automotive seats using metal and polymer composites. As such it is important to design the automotive seat suitable for manufacturing via this process and perform mechanical characterization on various subcomponents of the seat to ensure that the design and performance requirements provided by the auto manufacturer are met. The aim of this project is to develop a novel manufacturing technique to produce lightweight automotive seat by combining AM with conventional manufacturing processes. The car seat back panel will be designed via topology optimization and numerical simulations to minimize the overall weight while ensuring it meets all the performance requirements. The optimization of the seat back structure will be based on computational stress analysis to maximize the stiffness and minimize the weight. Materials currently used by Ford Motor Company will be adopted for a few subcomponents while the in-house composite materials will be used for the rest of the seat back. The composite and metallic materials will be tested to determine their mechanical properties as these are necessary for simulations. A novel manufacturing process will be developed to integrate AM metal inserts with discontinuous reinforced composite through large scale additive manufacturing and compression overmolding processes. The developed manufacturing technique will be used to fabricated various subcomponents suitable for the seat back design and mechanically tested to determine their properties. The manufacturing of the lightweight seat back design through this process involves integrated AM metal inserts with the composite structure for recliner connection. The manufacturing of the entire seat back which is lightweight through the novel manufacturing process will be discussed. The performance of the designed seat back will be investigated through numerical simulations and shown to meet all the requirements provided by the auto manufacturer. The final goal of developing a novel manufacturing process for lightweight automotive seats is met through design optimization of seat back, manufacturing of subcomponents, mechanical characterization, and validation through numerical simulations. The routes to achieve the final goal of the project and the depth in which they were investigated changed throughout the project due to personnel changes and the COVID-19 pandemic. The project resulted in the development of a novel manufacturing process to integrate metal inserts with tailored polymer composite preforms through overmolding. Leveraging this proven manufacturing process, a lightweight seat back was designed through topology optimization and numerical simulations. The designed seat back uses AM metal inserts and compression overmolding of tailored polymer composite preforms obtained via large scale additive manufacturing. The metal polymer composite structures fabricated through this process exhibited enhancement in stiffness and improved ductility upon testing. Overall, the project provided an alternative design and manufacturing technique for automotive seat back that enables weight saving while meeting the safety and performance requirements.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2021

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Autonomous Hydrogen Fueling Station

This project “Autonomous Hydrogen Fueling Station” covered the autonomous refueling with both gaseous hydrogen and liquid hydrogen. The part on gaseous hydrogen focused on the development of an autonomous robotic fueling arm that would couple to a fuel cell engine for hydrogen refueling without guidance from the forklift operator and budget period. Research was also covered for the robotic fueling with a commercial vehicle. The second phase of the project created the baseline for an autonomous liquid hydrogen transfer system that would minimize boil off losses by operating at thermodynamically efficient state points. For the development of the robotic fueling arm, testing was conducted to establish a baseline measurement of the accuracy and repeatability of a human operator positioning a lift truck in front of a dispenser. The goal was to establish the range of motion required for an autonomous fueling mechanism to mate a hydrogen nozzle with a receptacle on a fuel cell system installed in a forklift. The final design comprised a selective compliance articulated robot arm (SCARA)-type mechanism with two arms for horizontal motion and a ball screw for vertical movement and color and LIDAR cameras were used for marker identification and proximity awareness to guide the robotic arm to its target receptacle. Initial tests resulted in 199 out of 200 successful attempts at autonomous coupling of the dispensing coupler and a fuel cell engine, without hydrogen. The dispenser prototype was modified to include tubing for both hydrogen fuel and air purge lines, but subsequent tests were confounded by the shoulder motor over current errors which limited the robot from getting to the fully inserted position to achieve a positive latch. Robotic hydrogen refueling was successfully demonstrated over 1.5 hours of testing, Plug completed 29 successful latches with an average number of 4 sequential latches before failure. However, a robot capable of placing the nozzle with more force is required for higher reliability. For budget period two, a small scale (10 kg / transfer) automated control system was designed that would operate valves to control pressure and flow of liquid nitrogen between a source and receiving tank with an aim to minimize boil off losses by operating at the most thermodynamically efficient state points. Control system logic flow and a P&ID were developed prior to system safety characterization via HAZOP. A control narrative and system state points were defined. Delays in approval for a change of project objective and procurement issues precluded the construction and test of the final prototype system.

08 HYDROGEN↗
DOE OSTI2020

Spent Nuclear Fuel Exploratory Roadmaps - 20054

The U.S. Department of Energy manages nearly 2,500 metric tons of heavy metal (MTHM) of spent nuclear fuel (SNF) resulting from several decades of research, testing, and production reactors [DOE 1995]. This SNF is managed at the DOE Hanford Site in Washington State (Hanford), Idaho National Laboratory (INL), and Savannah River Site in South Carolina (SRS). From 1995 to 2004, the Department of Energy (DOE) made several key programmatic decisions, supported by appropriate documentation in accordance with the National Environmental Policy Act (NEPA) for the management of SNF [DOE 1995]. These decisions have provided an overarching framework for SNF management, as well as site-specific and SNF-related management decisions for the past two decades. In the years since these decisions were made, with the notable exception of the successful drying and packaging of the production reactor fuel at Hanford, a majority of the decisions have been largely unimplemented. Also, since these decisions were made, a number of changes that bear on considerations relative to the path forward have occurred. A SNF Exploratory Road map activity, identified reasonable alternate pathways for DoE's inventory of SNF. Three fundamental pathways were identified in this activity for the long-term management and disposition of DOE SNF: Direct Disposal Pathway, Existing Processing Pathway and Alternate Processing Pathway. Further actions would be required (e.g., evaluation of NEPA analysis, technology maturity evaluations, refinement of cost and risks, benefits and advantages relative to other alternate paths) to define and inform many key decisions that will result in the selection of the disposition pathways that are most beneficial to the US Government in dispositioning DoE's SNF inventory. DOE is currently managing all SNF safely; however, the age of DOE-owned SNF and facilities for storing and processing SNF, coupled with the uncertainty of the storage duration, necessitates decisions and actions to ensure that the infrastructure will be in place to ensure continued safe and effective long-term management and eventual disposal. The identification of reasonable alternate pathways took these considerations into account in an effort to proactively manage conditions that could challenge the safety of storing and managing DOE SNF over the time periods now contemplated and also to ensure that flexibility is preserved to ensure that the DOE SNF remains compatible with final disposition pathways when they become available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Fabrication of MiniFuel Compacts for High-Power Irradiation Testing of TRISO Fuel

MiniFuel compacts containing 20 individual tristructural isotropic (TRISO)-coated fuel particles were characterized to provide supporting preirradiation characterization data. The preirradiated characterization data supports irradiation capsule design and safety analysis, as well as downstream postirradiation examination (PIE) for a planned MiniFuel irradiation to explore high particle powers. The analysis included dimensional inspection, x-ray radiography and tomography, as-fabricated defect fraction analysis, and a matrix impurity analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗