Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “ID”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,567 records · Page 87
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Manufacturability and performance response of developed Ni-based alloys for reactor environments

Under the Advanced Materials and Manufacturing Technologies program, two Ni-based alloys fabricated by laser powder bed fusion (LPBF) at Oak Ridge National Laboratory and by laser powder directed energy deposition (LP-DED) at Idaho National Laboratory have been evaluated: γ′-strengthened Haynes 282 and solution-strengthened Inconel 625. Printing defects were observed in the LPBF 282 alloy fabricated using a Renishaw AM250 machine, likely due to particle spattering during printing. For both the LPBF and LP-DED 282 alloys, annealing at 1,180°C for 1 h resulted in partial recrystallization and a bimodal grain distribution, and subsequent aging for 4 h at 800°C led to the formation of a high density of nano-size γ′-strengthening precipitates. Creep testing performed at 750°C revealed lower creep life and ductility for the LPBF and LP-DED 282 compared with wrought 282. X-ray computed tomography combined with optical and scanning electron microscopy microstructural characterization revealed crack formation in the LPBF 282 alloy during creep testing, initiated either from printing defects or from creep cavitation at grain boundaries. Creep cavitation and cracking at grain boundaries were also observed for the LP-DED 282 alloy, and optimization of the printing strategies and postheat treatments will be needed to fabricate high-performance 282 components for high-temperature nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Cyber-Informed Engineering Workbook: SCADA (VoltVAR)

This case study workbook provides a hypothetical project to support discussion and application of Cyber-Informed Engineering principles. Participants in the workshop are encouraged to use the workbook to capture insights and lessons learned.

42 - ENGINEERING↗
DOE OSTI2025

Report on Preliminary Detailed Experimental Plan for Neutron Irradiation of A709 at ATR and HFIR

Advanced nuclear power technologies will use higher temperatures as a means to extract energy at a higher efficiency than current plants and therefore put a larger demand on the structural materials. Improved performance of structural materials could enable greater safety margins, longer plant lifetimes, and reduce maintenance costs. Alloy 709 is championed as the next generation of austenitic alloys for advanced nuclear reactors. In parallel to the ASME code case pursued, the AMMT program is initiating a neutron irradiation campaign to provide first-of-a-kind engineering data to establish operational design parameters and how the mechanical response is modified by environmental factors. This document refines the AMMT neutron irradiation campaign to a 4 year program to support the generation of creep knockdown factors for Alloy 709 and welded Alloy 709. The campaign is divided among two national laboratories, Oak Ridge National Laboratory and Idaho National Laboratory, to use the strengths of each laboratory. Through a cooperative plan, time-independent properties and time-dependent properties will be obtained across a large temperature window, nominally 300°C to 800°C, damage levels up to 10 dpa, and with and without the impacts of transmutation produced helium.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Energy Requirements for Integration of Nuclear Reactors with Iron and Steel Plants

This report identifies energy needs of heavy energy users within the domestic iron and steel industry and suggests solutions for integrating nuclear energy. The iron and steel industry, composed of several types of plants which perform different processes with varied energy demands and vectors, is a heavy consumer of electric power and fossil fuels including coke and natural gas. Almost all major process temperatures exceed the temperatures of direct heat available from advanced reactors, and so electricity and hydrogen were considered instead. Reference units were adopted and estimated energy demands computed for the blast furnace (BF), direct reduced iron (DRI) unit, electric arc furnace (EAF), and reheat furnaces. By utilizing production capacity data from industry reports, the ranges of power demands were estimated, including for hydrogen production by high temperature steam electrolysis (HTSE). For the EAF and DRI unit, more detailed integration studies with thermodynamic modeling were also conducted and determined a possible solution with a specific reactor design and number of modules. Furthermore, because many unit processes are co-located, entire plants were considered by adding the energy demands of the unit processes to form four hypothetical reference plants. The range of power needs for the reference plants is compatible with multi-unit banks of microreactors at the low end, and would create a need for multiple larger-capacity SMRs at the high end (1 GWe plus 0.14 GWt). Although the overall power need at the high end is well-matched with one present-day large reactor offering (1.1 GWe), redundancy considerations may require a minimum of two reactors, potentially eliminating the single large reactor from consideration. Auxiliary or house loads would increase the reference plant estimates. Finally, the report provides total estimated energy needs under integration of all U.S. units of each process (BF, DRI, EAF, and reheat furnaces), representing a national potential for nuclear energy in the industry. U.S. iron and steel plants may be candidates for integration with nuclear reactors via electricity and hydrogen, and many sites have energy requirements that correspond well to the capacities of several advanced nuclear power designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Report on Preliminary Detailed Experimental Plan for Neutron Irradiation of A709 at ATR and HFIR

Advanced nuclear power technologies will use higher temperatures as a means to extract energy at a higher efficiency than current plants and therefore put a larger demand on the structural materials. Improved performance of structural materials could enable greater safety margins, longer plant lifetimes, and reduce maintenance costs. Alloy 709 is championed as the next generation of austenitic alloys for advanced nuclear reactors. In parallel to the ASME code case pursued, the AMMT program is initiating a neutron irradiation campaign to provide first-of-a-kind engineering data to establish operational design parameters and how the mechanical response is modified by environmental factors. This document refines the AMMT neutron irradiation campaign to a 4 year program to support the generation of creep knockdown factors for Alloy 709 and welded Alloy 709. The campaign is divided among two national laboratories, Oak Ridge National Laboratory and Idaho National Laboratory, to use the strengths of each laboratory. Through a cooperative plan, time-independent properties and time-dependent properties will be obtained across a large temperature window, nominally 300°C to 800°C, damage levels up to 10 dpa, and with and without the impacts of transmutation produced helium.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Manufacturability and performance response of developed Ni-based alloys for reactor environments

Under the Advanced Materials and Manufacturing Technologies program, two Ni-based alloys fabricated by laser powder bed fusion (LPBF) at ORNL and laser powder directed energy deposition (LP-DED) at INL have been evaluated: γ′-strengthened Haynes 282 and solution-strengthened Inconel 625. Printing defects were observed in the LPBF 282 alloy fabricated using a Renishaw AM250 machine, likely due to particle spattering during printing. For both the LPBF and LP-DED 282 alloys, annealing at 1,180°C for 1 h resulted in partial recrystallization and a bimodal grain distribution, while subsequent aging for 4 h at 800°C led to the formation of a high density of nano size γ′-strengthening precipitates. Creep testing performed at 750°C revealed lower creep life and ductility for the LPBF and LP-DED 282 compared with wrought 282. X-ray computed tomography combined with optical and scanning electron microscopy microstructural characterization revealed crack formation in the LPBF 282 alloy during creep testing, initiated either from printing defects or from creep cavitation at grain boundaries. Creep cavitation and cracking at grain boundaries were also observed for the LP-DED 282 alloy, and optimization of the printing strategies and post heat treatments will be needed to fabricate high performance 282 components for high temperature nuclear applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

HDG-1 Fiber Bragg grating data analysis

The main goal of the High dose graphite 1 Advanced test reactor experiment was to study nuclear grade graphite at high fluences. Additional supplementary optical fiber instrumentation was added to this long duration experiment for instrumentation development purposes. The supplementary instrumentation consisted of two pure silica core, fluorine doped cladding optical fibers each etched with 9 fiber Bragg gratings, one fiber being heat treated for 9 hours at 750 C and 16 hours at 750 C, the other being heat treated for 24 hours at 550 C and 48 hours at 650 C. Fiber Bragg gratings are known to have issues of measurement drift when in high temperature and high radiation environments like what is encountered in the Advanced test reactor. At the culmination of this experiment, the optical fibers saw ~1.3E21 n/cm2 total fluence, which is at the highest fluences that fiber Bragg gratings have been studied to date. Reported here is the analysis of this data including radiation induced shift and changes in sensitivity.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2025

Development of Powder Production Methods for Advanced LEU Fuel Concepts

A set of novel fuel concepts has been proposed for use in advanced low-enriched uranium (LEU) systems that utilize powder metallurgical methods for fabrication of the fuel forms. Preliminary tests demonstrated the ability to produce powder but did not yield the desired quality to be used as feedstock in these applications. This study seeks to establish an improved powder production method for uranium-based alloys and evaluate the parameters required to refine the particles produced by size and morphology. Powder samples were fabricated via atomization and analyzed for particle morphology utilizing sieving methods and scanning electron microscopy (SEM). Surrogate testing displayed improved performance in producing small (<250 µm) spherical particles, and initial tests demonstrated the initial capability to atomize uranium. Further work is required to refine the atomization process to produce high quality uranium microspheres.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Application of Electrochemical Methods to Molten Salt Reactors: Draft TLR Documenting Assessment of Electrochemical Monitoring

Nuclear Regulatory Commission (NRC) is developing the regulatory framework and technical expertise to support regulatory review of advanced non-water reactors, including molten salt reactors (MSRs). In an MSR, it is essential that the salt chemistry be maintained in a desired range in terms of redox potential for reliable operation and mitigation of corrosion to structural materials in the reactor, particularly the reactor vessel and heater exchanger. Measuring the redox potential of the salt in the reactor would also allow for the material lifetimes to be predicted more accurately, and chemical issues to be diagnosed more quickly. In addition to the chemical composition analysis by ICP-MS, electrochemical methods including potentiometry and linear wave scanning (LSC) were also used in molten salt reactor experiment (MSRE) for redox potential monitoring purposes. Electrochemcial methods offers unique advantages such as quick turnaround in results and unique capability of in-line monitoring of redox potential, and are considered popular electroanalytical techniques that can be used to monitor redox potentials and salt chemistry including impurities. The last two decades have seen significant advances in science and engineering of electrochemical methods for potential application to molten salts. The primary goal of this report is to assist NRC in understanding the monitoring of the salt chemistry by electrochemical methods and provide NRC reviewers with necessary information and tools to support regulatory review of MSR designs. This reports consists of two majors parts—Part 1, chemical potential of molten salts and effects by fission process in MSR; Part 2, assessment of electrochemical methods for application to MSRs. The TRLs of the typical relevant electrochemical methods for molten salts were evaluated based on the DOE TRL guidelines and upon a review of the current status of the electrochemical methods for two typical salt systems, fluoride and chloride, for MSRs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Data Quality Assessment of Optiwatt Vehicle Telematics Data

In October 2024, the Idaho National Laboratory (INL) received data from Optiwatt (Compass Global, Inc.) describing the driving and charging behavior of electric vehicle (EV) drivers. The data shared had been collected from approximately 10,000 vehicles and included vehicle specifications, driving information like odometer readings at the beginning and end of origin-destination pairs (i.e., trips with identification of home for trip start and end for Tesla vehicles), and charging information such as charging energy consumed per charge session and if the charge occurred at home. The vehicle data were provided from 9 EV makes and 18 EV models, with production years ranging from 2012–2024, but more than 9,500 of the vehicles were Tesla EVs. The data includes more than six million trips and more than three million charging events that occurred between June 2023 to Aug 2024 and collected from California and the Eastern United States. The purpose of this report is to review the quality of the data received from Optiwatt and the feedback INL received from Optiwatt after data concerns were shared with them.

33 - ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2025

Steam Generators: Thermal Hydraulic Measurements and Instrumentation

Steam generators are integral components of both conventional and advanced reactors utilizing steam power cycles, as they influence the power conversion performance. Steam generator designs are highly influenced by the type of reactor coolant used, which mainly governs the different control, monitoring, and safety system actuation requirements. Therefore, the present work reviews the thermal hydraulics instrumentation within steam generators designed for pressurized water reactors (PWRs), high-temperature gas-cooled reactors (HTGRs), sodium fast reactors (SFRs), and molten salt cooled reactors (MSRs). Emphasis is given to measurement techniques that are either unique or particularly important to specific reactor types. Lastly, a discussion is provided on the instrumentation and measured parameters. The work identifies standard parameters of interest and is expected to aid the development of test beds and novel measurements techniques.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

AGR-5/6/7 Thermal Model with Non-uniform Gas Gaps

The daily as-run thermal analysis for the Advanced Gas Reactor (AGR)-5/6/7 experiment was documented in ECAR-5633 (Hawkes 2021). A finite element model was created for the entire test train, which consisted of five capsules. The fuel compacts, graphite holders, stainless-steel capsule walls, and all other major components were individually modeled, including all thermocouples (TCs). Daily heat rates for each compact and component in the model were input from daily as-run physics analyses. Daily gas compositions for each capsule were also input. The thermal conductivity of the compacts and graphite holders varied with fast neutron fluence. Gas mixture thermal conductivity was implemented using experimentally attained values from the literature. Fluence and temperature-dependent thermal conductivity was used for the graphite components and the fuel compacts. Radiation heat transfer was implemented. The model was tuned in an attempt to match the TC readings during the first cycle by adjusting the Neolube (graphite lubricant) thickness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Analysis of temperature effects on self-powered neutron detectors in a gamma field

Self-powered neutron detectors (SPND) are small, robust, in-core radiation sensors that have the potential to operate in high temperature environments (above 600°C). However, previous experiments indicated temperature related effects in the form of displacement currents. Given the recent irradiation at The Ohio State University Research Reactor using a delayed-response rhodium-based SPND alongside a prompt-response tantalum-based SPND, it was concluded that only low-energy electrons from gamma interactions were the cause of these temperature effects. A series of heated gamma irradiations with temperatures up to 750°C of SPNDs were performed at the Advanced Test Reactor Gamma Facility in order to better characterize these displacement currents. The results from the experiments demonstrated a strong relation between the displacement currents that occurred during temperature ramps to thermoluminescence (TL) glow peaks—the SPND with Al2O3 insulation has one glow peak within 750°C and the SPND with MgO has three. A demonstration was performed to calculate the trap energy levels of the displacement currents through TL glow peak analysis. Unfortunately, the methods employed provided inconsistent results, requiring an updated model based on ongoing irradiation instead of post-irradiation measurements.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2025

Digitally Optimized Autonomous Robotic System for Hanford Waste Tank Handling [Slides]

Idaho National Laboratory (INL) is developing a robotic system and digital twin to perform monitoring, inspection, and mapping of underground tank pits at the Hanford Site, operated by the U.S. Department of Energy (USDOE). The Autonomous Pit Exploration System (APES) will consist of four primary components: an electric vehicle truck with a robotic system developed by INL to deploy payloads into the tanks through small diameter preexisting risers, a robotic arm developed by Florida International University (FIU) for the purposes of deploying a variety of payloads into the tank pit environment, and a small robotic COTS crawler platform for examining the bottom of the pits.

42 - ENGINEERING↗
DOE OSTI2025

Report on Potential Challenges and Impacts of Advanced Nuclear-Maritime Applications in the U.S.

This report is the third deliverable of a series of reports set forth by the Department of Energy (DOE), led by the American Bureau of Shipping (ABS), for the research award, “Accelerating Commercial Maritime Demonstration Projects for Advanced Nuclear Reactor Technologies.” The report highlights the challenges that are facing both the nuclear and maritime industry as technology seeks to integrate.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Acceleration of Thermochemistry Solves in MOOSE and Pronghorn

This work focuses on the development and implementation of strategies to accelerate thermochemical calculations within MOOSE-based multiphysics simulations, particularly for applications in MSRs. We highlight the inherent complexity of nuclear materials, which require a multiscale approach to accurately model their behavior across various physical domains, including mechanical, chemical, and thermal phenomena. Thermochemical equilibrium calculations are crucial for predicting material properties and enhancing the fidelity of these simulations. The integration of Thermochimica, a Gibbs energy minimizer, into MOOSE allows for the direct minimization of Gibbs energy at every point on the mesh. However, the computational cost of such integration is significant. To address this, we explored acceleration strategies such as multi-threading support and the use of a thermodynamic ValueCache to reduce redundant calculations. Additionally, we investigated modifications to Thermochimica to enable phase constraints and improve its coupling with phase-field models, which are essential for simulating microstructural evolution and corrosion in MSR. These efforts aim to optimize the computational efficiency and accuracy of multiphysics simulations, thereby supporting the development of reliable and efficient nuclear materials for next-generation reactor technologies.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Advanced Test Reactor Neutron Dosimetry Report – 174 A & B

Results of neutron dosimetry measurements in ATR for cycle 174 A B are presented. Cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗