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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 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Combining four-point bending and corrosion to map stress-dependent corrosion susceptibility of 316H stainless steel in FLiNaK

Understanding the effect of stress on the corrosion of steels in molten salts is important for material screening and safety assessment of molten salt reactors. We present a method that combines four-point bending with corrosion testing, enabling the mapping of corrosion susceptibility as a function of local stress from a single specimen. Guided by finite element analysis, a four-point bending setup was used to bend a 316H stainless-steel bar under controlled elastic stress during a 100-hour exposure in FLiNaK at 700 °C. Post-exposure characterization using scanning electron microscopy and energy-dispersive X-ray spectroscopy revealed a pronounced correlation between stress and corrosion. Regions under tensile stress exhibited significantly deeper attack depths and greater chromium depletion along grain boundaries, whereas compressive zones showed shallower corrosion cracks. The stress effect is attributed to stress concentration under tensile loading, which promotes chromium dissolution and crack propagation. This technique can be readily applied to other structural alloys, enabling quantitative measurement of corrosion susceptibility as a function of local stress.

316H↗
DOE OSTI2022

TEM Specimen Manufacturing to Support Analysis of IASCC Mechanisms in Irradiated Austenitic Steels – Light-Water– Relevant Materials

The present report describes the manufacturing of samples for advanced microstructure studies of mechanisms of irradiation-assisted stress corrosion cracking (IASCC) in the materials of light-water reactor (LWR) internal components. The samples were manufactured via FIB (focused ion beam milling) from 4-point-bend specimens subjected to oxidation and mechanical loading in a simulated LWR environment (primary water, PW). The corrosion tests were conducted at the University of Michigan (UM), and the tested specimens were transported to ORNL to perform FIB work. Finally, the manufactured FIB specimens were shipped back to the UM for microstructure characterization. Section 1 in the report provides a brief introduction; Section 2 describes materials, their in-service and irradiation history, and the geometry of specimens for IASCC testing. Section 3 documents typical work steps (specimen mapping) and reports the number of manufactured FIB objects. Section 4 briefly describes the preliminary results of microstructure analysis performed at the University of Michigan and highlights the key findings. The summary and conclusions section discusses future possible future steps and generalizes the obtained results, their unique character, and their importance for the nuclear industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Prediction of Creep-Induced Strain Using a Symbolic Regression-Based Model

Material creep under high-temperature conditions limits the lifetime and safety of structural systems such as advanced nuclear reactors. Conventional creep testing is slow and often produces inconsistent results across nominally identical experiments, making lifetime prediction uncertain. Here, to address these challenges, this work develops a data-driven symbolic regression (SR) model that consolidates results from duplicate creep tests and predicts the remaining strain-time curve of an ongoing experiment. The method uses piece-wise multi-objective SR with physical constraints to generate analytic, interpretable functions describing transient creep strain. Applied to Inconel Alloy 617 data, the approach achieved relative mean absolute errors of 1.0–9.5%, providing closed-form predictions of strain evolution. These results demonstrate a first step toward reducing the duration and cost of long-term creep testing while retaining physically interpretable model forms.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1975

Beryllium reflected cavity reactor for UF6 critical experiments

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diam by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials are available. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗
NASA NTRS2006

Bench-Scale Monolith Autothermal Reformer Catalyst Screening Evaluations in a Micro-Reactor With Jet-A Fuel

Solid oxide fuel cell systems used in the aerospace or commercial aviation environment require a compact, light-weight and highly durable catalytic fuel processor. The fuel processing method considered here is an autothermal reforming (ATR) step. The ATR converts Jet-A fuel by a reaction with steam and air forming hydrogen (H2) and carbon monoxide (CO) to be used for production of electrical power in the fuel cell. This paper addresses the first phase of an experimental catalyst screening study, looking at the relative effectiveness of several monolith catalyst types when operating with untreated Jet-A fuel. Six monolith catalyst materials were selected for preliminary evaluation and experimental bench-scale screening in a small 0.05 kWe micro-reactor test apparatus. These tests were conducted to assess relative catalyst performance under atmospheric pressure ATR conditions and processing Jet-A fuel at a steam-to-carbon ratio of 3.5, a value higher than anticipated to be run in an optimized system. The average reformer efficiencies for the six catalysts tested ranged from 75 to 83 percent at a constant gas-hourly space velocity of 12,000 hr 1. The corresponding hydrocarbon conversion efficiency varied from 86 to 95 percent during experiments run at reaction temperatures between 750 to 830 C. Based on the results of the short-duration 100 hr tests reported herein, two of the highest performing catalysts were selected for further evaluation in a follow-on 1000 hr life durability study in Phase II.

Tomsik, Thomas M.↗
NASA NTRS2006

NASA Tech Briefs, February 2006

Topics discussed include: Nearly Direct Measurement of Relative Permittivity; DCS-Neural-Network Program for Aircraft Control and Testing; Dielectric Heaters for Testing Spacecraft Nuclear Reactors; Using Doppler Shifts of GPS Signals To Measure Angular Speed; Monitoring Temperatures of Tires Using Luminescent Materials; Highly Efficient Multilayer Thermoelectric Devices; Very High-Speed Digital Video Capability for In-Flight Use; MMIC DHBT Common-Base Amplifier for 172 GHz; Modular, Microprocessor-Controlled Flash Lighting System; Generic Environment for Simulating Launch Operations; Modular Aero-Propulsion System Simulation; X-Windows Socket Widget Class; Infrastructure for Rapid Development of Java GUI Programs; Processing Raman Spectra of High-Pressure Hydrogen Flames; X-Windows Information Sharing Protocol Widget Class; Simulating Humans as Integral Parts of Spacecraft Missions; Analyzing Power Supply and Demand on the ISS; Polyimides From a-BPDA and Aromatic Diamines; Making Plant-Support Structures From Waste Plant Fiber; Large Deployable Reflectarray Antenna; Periodically Discharging, Gas-Coalescing Filter; Ion Milling On Steps for Fabrication of Nanowires; Neuro-Prosthetic Implants With Adjustable Electrode Arrays; Microfluidic Devices for Studying Biomolecular Interactions; Studying Functions of All Yeast Genes Simultaneously; Polarization Phase-Compensating Coats for Metallic Mirrors; Tunable-Bandwidth Filter System; Methodology for Designing Fault-Protection Software; and Ground-Based Localization of Mars Rovers.

Source record↗
DOE OSTI2022

The Nuclear Fuel Cycle: Safe Management of Spent Nuclear Fuel

The aim for storage of spent nuclear fuel (SNF) either in wet or in dry storage systems is to ensure general safety objective s are met throughout a desired storage period. Staff at the Savannah River National Laboratory (SRNL), in collaborations with partners at other national laboratories, industry research organizations, and the University of South Carolina (UofSC), have performed materials aging testing and analyses, and have established nuclear materials aging management programs to support extended periods of safe storage of research reactor (RR) SNF and of commercial power reactor (PR) SNF pending ultimate disposal. Several example challenges include susceptibility of aluminum SNF from research reactors to corrosion in poor quality water (wet storage). In dry storage, aluminum SNF can release hydrogen via radiolysis of the hydrated oxides on the aluminum cladding. Austenitic stainless steel canisters used for dry storage are susceptible to chloride-induced stress corrosion cracking (outside-in attack) that threaten the confinement boundary provided by the canister. This paper further describes these challenges, among others, and the formulated solutions to support extended safe storage of SNF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

The objective of this study is to develop a 3D/NK model to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA). This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following in gas reactors a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode an extensively study has been carried out. Some tests with linear coolant mass flow rate change demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Machine learned interatomic potential for dispersion strengthened plasma facing components

Tungsten (W) is a material of choice for the divertor material due to its high melting temperature, thermal conductivity, and sputtering threshold. However, W has a very high brittle-to-ductile transition temperature, and at fusion reactor temperatures (≥1000 K), it may undergo recrystallization and grain growth. Dispersion-strengthening W with zirconium carbide (ZrC) can improve ductility and limit grain growth, but much of the effects of the dispersoids on microstructural evolution and thermomechanical properties at high temperatures are still unknown. We present a machine learned Spectral Neighbor Analysis Potential for W–ZrC that can now be used to study these materials. In order to construct a potential suitable for large-scale atomistic simulations at fusion reactor temperatures, it is necessary to train on ab initio data generated for a diverse set of structures, chemical environments, and temperatures. Further accuracy and stability tests of the potential were achieved using objective functions for both material properties and high temperature stability. Additionally, validation of lattice parameters, surface energies, bulk moduli, and thermal expansion is confirmed on the optimized potential. Tensile tests of W/ZrC bicrystals show that although the W(110)–ZrC(111) C-terminated bicrystal has the highest ultimate tensile strength (UTS) at room temperature, observed strength decreases with increasing temperature. At 2500 K, the terminating C layer diffuses into the W, resulting in a weaker W–Zr interface. Meanwhile, the W(110)–ZrC(111) Zr-terminated bicrystal has the highest UTS at 2500 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Mechanical Testing of Additively Manufactured Materials

This report summarizes research activities conducted at Argonne National Laboratory in support of the development, qualification and certification of additively manufactured (AM) metallic components to allow for innovative reactor design and licensing. In this program, AM 316L stainless steel (SS) was evaluated in both light water reactor (LWR) environments as well as at higher temperatures expected in advanced reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AGR-5/6/7 Irradiation Disassembly and Metrology First Look

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of high-temperature gas-cooled reactors (HTGRs), which are graphite-moderated nuclear reactors cooled with helium. This work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching program goal is to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials (e.g., graphite), safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product transport and source term determination. The ART AGR program has conducted four distinct fuel irradiation experiments in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The first of these irradiation tests, designated AGR-1, began in ATR in December of 2006 and ended in November 2009. This experiment was primarily to act as a shakedown test of the multi-capsule test train design and to provide early data on fuel performance that would be used in fuel fabrication process development. AGR-1 fuel kernels were produced on an engineering scale, but the TRISO coatings and cylindrical fuel compacts were fabricated on a laboratory scale. The AGR-1 PIE was completed and the final report was published in 2015. The second irradiation test, AGR-2, started in ATR in June 2010 and ended in October 2013. The AGR-2 irradiation test was designed to provide fuel performance data for coated particles fabricated on an engineering-scale pilot line using a coater with an internal chamber diameter of 150 mm (6 in.). The final PIE report was published in 2021. AGR-3/4, a single irradiation that combined what were originally conceived as the third and fourth tests, was to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport. PIE of the AGR-3/4 experiment is still in progress as of this writing. The subject of this report is AGR-5/6/7, the final qualification test of AGR TRISO fuel made entirely at the engineering scale.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

An Evaluation and Qualification of U.S.-Based Research Reactors for Irradiation Capabilities Supporting Advanced Nuclear Systems

Irradiation experiments are a prerequisite for evaluating nuclear reactor system designs, analyzing the performance of these systems, and obtaining licenses. Likewise, irradiation facilities are necessary for producing the radioisotopes used in industrial and medical applications. Recent developments in modeling and simulation capabilities and advancements in computational resources have further enabled the design of irradiation experiments for evaluating radiation-induced phenomena and determining nuclear fuel, material, and system design and safety criteria pertaining to both normal and accident scenarios. These computational tools and models require comprehensive experimental datasets acquired under prototypic radiation conditions—for exploring material and system performance under the uniquely harsh environments found in nuclear reactors—to enable verification and validation for qualification and licensing purposes. However, qualification of irradiation experimental facilities, primarily research and test reactors (RTRs), necessitates that their performance be evaluated based on the irradiation environment (e.g. flux, power, testing capabilities) using an appropriate scoring matrix. Although many university campus RTRs are available for research and development (R&D) activities and initiatives, this study focuses on evaluating and qualifying the irradiation facilities (mostly RTRs) within the United States that are suitable for advanced nuclear fuel, material, and system irradiation experiments aimed at establishing operational-performance limits and informing component and fuel designs so as to improve operational efficiencies and mitigate proliferation vulnerabilities, as well as for radioisotope production aimed at multipurpose applications. As a result, the findings of the present study support the acceleration of nuclear fuel and material qualifications, thus hastening new and advanced nuclear energy system demonstrations and radioisotope production efforts by using extended R&D.

irradiation experiment↗
DOE OSTI2020

Sampling and Analysis Plan for Investigation of Undeclared Nuclear Activities - 20104

The nuclear-centered energy policy in South Korea is moving towards a nuclear-free era. It also facilitates an action or progress on the denuclearization of the Korean peninsula. The Nuclear Safety and Security Commission (NSSC) launched a new R and D project which attempts to verify nuclear activities via analyzing environmental samples. Environmental samples possibly contain significant radiological information relevant to past undeclared activities. In order to precisely verify them, it is essential to provide 1) the concrete information of target facilities, 2) appropriate methodologies and procedures, and 3) well-suited instrumentation and equipment corresponding to sample signatures. It is also important to provide scientific reliability or validity in any process such as sample collection, handling, transportation and laboratory analysis. Therefore, the information of nuclear facilities (fuel fabrication plant, graphite moderated reactor, reprocessing plant) in North Korea are reviewed in order to determine parameters for appropriate sampling methods representing a facility's characteristics. Secondly, a sampling procedure for radioactive elements is being developed based on international and domestic standards such as ASTM (American Society for Testing and Materials), KS (Korean industrial Standards), and International Atomic Energy Agency (IAEA) manuals. The procedure contains some specific information including sampling process, sampling tools, amount of sample, sample identification, and other related information. Developing the qualified national guidelines is most important since sampling and analyzing results can be used as evidences of nuclear activities. The guideline developed through this work will clearly define various methodologies such as document verification, sampling procedures, collection methods, sample storage criteria, chain of custody, and transportation conditions. The final objective of this work is to establish the infrastructure and regulatory framework for assessment of nuclear activities, and provide a national guideline for sampling and analysis. This will bring confidence in the results and play a critical role in complete denuclearization progress in South Korea. (authors)

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Conceptual Design for a Blanket Tritium Extraction Test Stand

Future fusion reactors must be able to breed the tritium they will consume. Several breeding and tritium extraction technologies are under investigation internationally. PbLi is of particular interest as a breeder material. A new forced convection PbLi loop, the Tritium Extraction eXperiment (TEX), is being designed and constructed to investigate tritium extraction from PbLi. Specifically, TEX will serve to investigate the extraction efficiency of various vacuum permeator membranes and configurations. The major components of the loop include a moving magnet pump, reverse permeator, furnace test section, analysis chamber, supply tank, and plenum. As significant hazards are present in such experiments, safety is an integral focus of the experimental design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Oklo and BWXT Short Decks

Two abstracts, one for each deck: Oklo is planning on producing fuel using spent EBR-II material and demonstrating an NRC licensed reactor. This presentation gives a high level overview of Oklo's progress in this work. The BWXT Advanced Nuclear Reactor Experiment 1 (BANR-1) is an irradiation experiment at INL's Advanced Test Reactor (ATR) that will irradiate novel Tristructural Isotropic (TRISO) fuel for potential future use by BWXT. This presentation gives a program overview and progress.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

STARTR: An Open-Source MARVEL model for the NRIC Virtual Test Bed [Poster]

The National Reactor Innovation Center (NRIC) seeks to improve the understanding of microreactor physics in industry and academia through the development of a Microreactor Applications Research Validation and Evaluation (MARVEL) reactor-based model, published on the Virtual Test Bed (VTB). To achieve this goal, the Sodium-cooled Thermal-spectrum Advanced Research Test Reactor (STARTR) model was built using publicly available MARVEL specifications where possible and approximations where applicable, and was optimized for fast runtimes for researchers to receive rapid simulation feedback. STARTR will fill a gap between stakeholder interest and available models, as the first Sodium-cooled Thermal Reactor (STR) hosted on the VTB with baseline performance sanctioned by INL. This project involved the definition of all materials used in the reactor, geometry and all reactor subcomponents, and assertion of tallies and simulation settings within OpenMC 0.13.3. This poster details a small subset of the overall reactor physics testing: the two-dimensional power peaking factors and the flux energy spectrum, as well as plots of the created geometry. Future work includes code-to-code verification between the OpenMC-based model and a separately designed MCNP 6.2-based model.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

Survey Corrosion Testing Methodologies for Additively Manufactured Materials – PNNL

The Advanced Materials and Manufacturing Technology (AMMT) Program intends to develop cross-cutting technologies in support of a broad range of nuclear reactor technologies, and to maintain U.S. leadership in materials and manufacturing technologies for nuclear energy applications. The overarching vision of the AMMT Program is to accelerate the development, qualification, demonstration, and deployment of advanced materials and manufacturing technologies to enable reliable and economical nuclear energy.

36 MATERIALS SCIENCE↗