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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Massachusetts Institute of Technology Reactor LEU Fuel Element Flow Test Preliminary Design

As part of the U.S. National Nuclear Security Administration’s (NNSA) mission to minimize the civilian use of weapon-grade highly enriched uranium (HEU) fuels, the NNSA Office of Material Management and Minimization Conversion Program is collaborating with six U.S. High Performance Research Reactors (USHPRR), including one critical facility, to convert from the use of HEU to low-enriched uranium (LEU) fuel. The conversion objectives for the USHPRR are to develop LEU fuel element designs that will ensure safe reactor operations and maintain the existing experimental facilities performance. The work is being conducted through many interrelated activities that are being completed by stakeholders across organizations. Within the Reactor Conversion (RC) Pillar of the USHPRR Project, four of the USHPRR, including the Massachusetts Institute of Technology Reactor (MITR-II, also referred to as MITR), have progressed through preliminary fuel element design using the proposed monolithic alloy of uranium- 10 wt% molybdenum (U-10Mo). Preliminary fuel element design and safety analyses have been completed for MITR. This work has relied on preliminary data for properties, performance, and fabrication tolerances for the fuel systems that have been produced by the Fuel Qualification (FQ), Fuel Fabrication (FF), and RC Pillars of the USHPRR Project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Effects of ZrN Coating and Heat Treatment on U-Mo Dispersion Fuel under Irradiation

The stability of U-Mo fuel particles in Al matrix can be improved using ZrN coating and heat treatment. This study investigated irradiation behaviors of fuel plates containing U-Mo fuel particles fabricated with different conditions including heat treatment and the thickness of ZrN coating. Additionally, different fission densities were applied to each fuel plate to understand how varying fission densities affect its irradiation behaviors. This study showed that more recrystallization of grains and development of high burnup structure (HBS) occurred in irradiated fuel particles with higher fission densities. Heat treatment reduced the accumulation of fission gas bubbles in the fuel particles by coarsening their grains. The fuel particles with a ZrN coating thickness of 1.2-1.8 µm showed less significant development of U-Mo/Al interaction layers than the particles with a coating thickness of 0.5 µm after irradiation. This might indicate the existence of a critical thickness of ZrN coating to minimize the formation of the interaction layers. Fission gas bubbles were observed at grain boundaries of irradiated U-Mo fuel particles at low fission densities, and the formation of dendritic features with a darker contrast was identified in backscattered electron images. These features stemmed from grain boundaries with fission gas bubbles or HBS, and some of them contained the bubbles, indicating that they might represent an early stage of HBS propagation in the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Fabrication of surrogate oxide spent fuel with various cracking patterns and design of an axial gas transport apparatus

In this article, understanding gas transport behavior in nuclear fuel rods is important for the design, performance, and safety of nuclear fuels. Surrogate materials help enable efficient research by reducing both the costs and the amount of time required. A parametric study using Darcy’s law is completed that demonstrates the feasibility of observing pressure decay over short 13-to-15-cm specimens to enable full characterization of the fabricated specimens using x-ray computed tomography. This paper demonstrates that thermally shocked and mechanically compressed alumina pellets produce surrogate samples whose various cracking patterns are representative of the severity of cracking observed as a function of burnup in irradiated nuclear fuels. Furthermore, image analyses of the cracking patterns—in conjunction with gas transport testing using surrogate samples—affords a valuable accelerated basis for developing gas transport simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Nuclear fuels for transient test reactors

Transient test reactors with the ability to test fissile specimens under extreme conditions have been crucial tools in the development of nuclear technologies. Less than 10 unique facility designs have ever been constructed, most of which remain operational today and still use the original nuclear fuel constructed for them more than 40 years ago. Historic fuel systems for transient test reactors vary in significant ways which have marked influences on reactor capabilities. Eventually, new fuel will be needed to support the longevity of transient test reactor missions. This paper reviews precedent transient reactor fuel systems in the context of their unique requirements. A few key conclusions are illustrated by comparing and contrasting these transient test reactors. Fuel composites which are mostly graphite can enable transient reactors with very high neutron fluence capability (>2E16 n/cm 2 ) and are amenable to longer “shaped” transients but cannot achieve pulses <10 ms in duration. Reducing the graphite-to-uranium ratio can yield a very narrow pulse capability but delivers less fluence and requires cores with considerably more fissile material. Designs based on uranium dioxide (UO 2 ) make use of readily available materials to create compact cores with narrow pulse width capabilities but with moderate neutron fluence capabilities (~2E15 n/cm 2 ). Uranium zirconium hydride (U-ZrHx) is a well-established fuel system for pulsing reactors which has been intermittently manufactured throughout the decades. U-ZrHx offers similar capabilities to UO 2 designs in terms of nuclear kinetics, but with about half the fluence capability (~1E15 n/cm 2 ). An evolution of the UO 2 system, termed “ternary ceramic” fuel, shows that dispersing UO 2 in zirconium oxide and calcium oxide can increase fluence capability greatly (~7E15 n/cm 2 ), but is not presently a commonly available fuel form. A unique composite of UO 2 and beryllium oxide (UO 2 -BeO) can be used to create a core with similar kinetics and compact core geometry as U-ZrHx designs, but with significantly higher fluence capability (~6E15 n/cm 2 ). Like ternary ceramic fuel, newly fabricated UO 2 -BeO would require reestablishing its historic manufacturing process which would be further complicated by the health hazards associated with beryllium. In conclusion, like most engineering problems, there is no perfect solution, but this paper outlines the advantages and disadvantages of candidate fuel options to help guide detailed evaluations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Design of a separate effects MiniFuel irradiation experiment investigating microstructure evolution in high burnup UO 2

The microstructural evolution of UO 2 fuel pellets during commercial operation in light water reactors (LWRs) is known to vary significantly across the pellet radius due to spatial variations in local temperature and burnup. The primary obstacle to extending LWR refueling cycles to 24-month intervals is the susceptibility of certain high burnup fuel microstructures to fuel fragmentation, relocation, and dispersal (FFRD) during a loss of coolant accident (LOCA). Although FFRD of the high burnup structure in the rim region of a pellet is well studied, the fine fragmentation that has been observed in a second region, near the midradius of the pellet (termed the “dark zone”) following mock LOCA testing of high burnup commercial fuel rods is less understood. This paper describes the design, analysis, and execution of a separate effects MiniFuel irradiation experiment that aims to identify the specific temperature and burnup regimes under which FFRD-susceptible dark zone microstructures form. The small disc specimens (3 mm diameter by ∼0.3 mm thick) enable more precise control of the relatively uniform temperature and burnup conditions. A total of 42 specimens were fabricated with typical LWR fuel densities (∼96%–98% of theoretical density) and grain sizes (∼12 μm) and are being irradiated over a range of temperatures (600°C–1000°C) and discharge burnups (50–72 MWd/kg-U) that bound the midradius region of high burnup LWR fuel. Fuel specimens with identical 235 U enrichments were inserted in two irradiation locations in the High Flux Isotope Reactor and are currently undergoing irradiation to further evaluate the impact of rate effects (fission rate, time at temperature) on the microstructural evolution. The fuel fabrication and the thermal and neutronic simulations used for designing the experiment are detailed in this paper. A secondary objective of the experiment is to observe fission gas release (FGR) under the various irradiation conditions, and this work provides first-order predictions of FGR from all fuel specimens. The insights gained from these experiments will inform future high burnup core designs that could minimize the formation of susceptible microstructures and ultimately enable 24-month refueling cycles while minimizing the fraction of the fuel susceptible to FFRD.

FFRD↗
DOE OSTI2020

α-U and ω-UZr 2 in neutron irradiated U-10Zr annular metallic fuel

Here, to develop metallic fuel with ultra-high burnup of 30%-40%, an annular U-10Zr fuel with 55% smear density was fabricated through a casting route and irradiated at the Advanced Test Reactor at Idaho National Laboratory. The annular fuel design also serves as a demonstration of the feasibility to replace sodium bond with a helium bond to benefit the geological disposal of irradiated fuel, cut the cost of fuel fabrication, and boost the overall metallic fuel economy. This paper reports the results from transmission electron microscopy (TEM) based post-irradiation examination of this fuel type irradiated to a burnup of 3.3% fissions per initial heave metal atoms for initial screening purpose. After irradiation, the initial U-10Zr separated into an a-U annular region and an UZr 2+x center region with nanoscale spinodal decomposed microstructure. Because of the provided large amount of coherent interface areas, the fission gas atoms and vacancies generated in UZr 2+x phase are possibly pinned at the interface areas, leading to 20 times smaller fission gas bubbles than those in the neighboring a-U. The large bubbles in a-U become connected and merged into large pores that provide fast release paths for fission gas which prevents further fuel swelling. The fuel center still has open space to accommodate further fuel swelling from solid fission products at higher burnup. Other neutron irradiation induced phase and microstructure change are also characterized and compared with traditional solid fuel designs.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Bayesian estimation of cross-section and errors of experiment and calculation for accurate prediction of neutronic characteristics

In some experiments, there might be a correlation of experimental errors such as an error caused by fabrication. The correlation of calculation errors also might exist among analyses of experiments. In the present study, considering these correlations, the cross-section and errors of experiment and calculation are adjusted for the accurate prediction of neutronic characteristics. The adjustment method is derived based on Bayesian theory. The 'a posteriori' distributions of cross-section and errors of experiment and calculation are obtained by the derivation. The case study is performed by using the benchmarks of light-water moderated experiments reported in ICSBEP. The error from fuel rod fabrication and measurement are included in the experiment error, and the error from the simplification of the benchmark model is included in the calculation error. Since these errors have a positive correlation among the benchmarks, the prediction accuracy is expected to be improved by considering these correlations. For the sensitivity analysis of the correlation among benchmarks, the correlation coefficient of experiment error and that of calculation error is assumed from 0.0 to 0.9. The results indicate that the prediction accuracy of the multiplication factor can be improved by increasing the correlation coefficient of experiment error and that of calculation error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Molten pool dynamics and humping suppression in high-speed laser welding via tailored beam configurations

High-speed laser welding is essential for increasing the production rate of fuel cell fabrication. However, when the welding speed exceeds a critical limit, humping occurs and reduces the weld quality. In this study, two tailored beam configurations, including an adjustable ring mode and a dual-beam configuration, were employed to suppress humping. Computational fluid dynamics simulations were performed to elucidate the underlying suppression mechanisms. Here, the results show that, in the adjustable ring mode, humping mitigation arises from a reduced backward cross-sectional melt flow rate and a more stable molten pool. In the dual-beam configuration, humping suppression is attributed to the deceleration of melt flow, the conduction-mode behavior of the trailing beam, and the widening of the molten pool induced by the trailing laser. Furthermore, because the dual-beam configuration directly modifies the trailing molten pool dynamics, it achieves more effective humping suppression, extending the welding speed limit to 1.50 m/s, compared with 1.00 m/s for the adjustable ring mode.

08 HYDROGEN↗
DOE OSTI2025

Effects of shielding gas and spot reduction on humping suppression in high-speed laser welding

High-speed laser welding is a crucial process to enhance the production rate of fuel cell fabrication. However, exceeding a critical welding speed leads to defect formation such as humps and root concavities, which degrade weld quality and mechanical performance. This study employed shielding gas and spot reduction approaches to suppress these defects. The relationships between individual defect types and their influence on weld strength were systematically examined. Moreover, in situ high-speed synchrotron X-ray imaging was employed to comprehend the mechanisms of humping mitigation by shielding gas through analysis of instantaneous keyhole and melt pool dynamics. The observations reveal that shielding gas mitigates humping by shortening the molten pool through enhanced surface heat convection and by altering surface tension to prevent melt flow collisions. The effects of shielding gas and spot reduction were further incorporated into the dimensionless humping index (π h ) that describes the humping formation tendency. These findings provide mechanistic insights and quantitative guidance for defect control strategies in high-speed laser welding.

08 HYDROGEN↗
DOE OSTI2021

Enhancing isotope mixing in U-10Mo downblend castings with electromagnetic stirring

Uranium alloyed with 10 wt percent molybdenum (U-10Mo) is the proposed fuel for use in the United States’ high-performance research reactors. The U-10Mo fuel is fabricated by a two-step casting process that downblends highly enriched uranium with depleted uranium (DU) and/or natural uranium (NU) and subsequently alloys the resulting high-assay low-enriched uranium (HALEU) with Mo. Currently, a two-step casting process is needed to meet the ingots’ U-235 enrichment homogeneity specifications. This work demonstrates the ability to provide more homogeneous U-235 distributions during downblend casting by using zirconia for crucibles rather than graphite. When graphite is used as a crucible, the electromagnetic field produced by the induction heater couples directly with the graphite. Zirconia is nonconducting in this study, and therefore does not couple with the field. The induction field couples directly with the metal, and causes electromagnetic stirring (EMS) in the molten metal pool. Eight downblend casting experiments were carried out in this work. Four were performed with zirconia crucibles and four with graphite crucibles. The U-235 enrichment was measured at nine discrete points in each casting using laser ablation multi-collector inductively coupled mass spectrometry (LA-MC-ICP-MS). The enrichment homogeneity was approximately eight times better in zirconia crucible castings than in graphite crucible castings as measured by the plate enrichment range and enrichment coefficient of variation. The results show a single casting step could be used to meet USHPRR enrichment specifications if electromagnetic stirring is present during casting. A related investigation was carried out to determine whether it is appropriate to measure enrichment on as-cast ingots or whether heat treated specimens should be used to accurately characterize enrichment. The results show that the as-cast microstructure plays a significant role in enrichment homogeneity. It is recommended that all enrichment measurements be done after the homogenization heat treatment typical of U-10Mo processing.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Unveiling mechanisms and onset threshold of humping in high-speed laser welding

The fabrication of fuel cells relies on a rapid laser welding process. However, challenges arise with the occurrence of humping when the welding speed surpasses a critical threshold, which poses difficulties in achieving a smooth surface finish and a consistent weld strength. This study aims to elucidate the humping mechanisms by analyzing the morphology of molten pool and the characteristics of melt flow at varying welding speeds via in situ synchrotron high-speed X-ray imaging and computational fluid dynamics simulations. Our findings indicate that the short keyhole rear wall, the high backward melt velocity, and the prolonged tail of molten pool are the primary factors contributing to the onset of humping. Furthermore, a dimensionless humping index (π h ) was introduced, which successfully captured the onset threshold of humping across different literatures. This index not only provides a quantitative description of the humping formation tendency but also serves as a valuable tool for optimizing the laser welding process.

42 ENGINEERING↗
DOE OSTI2023

Low-Power Nuclear Heating Tests of UN Fuel Particles in Hydrogen Environment for Nuclear Thermal Propulsion

Recently, there has been a reemergence of interest within the space-exploration community regarding nuclear thermal propulsion (NTP) for interplanetary travel. NTP, as conceptualized, is a compact propulsion source, is relatively light because there is no need for an oxidizing agent and has a very high specific impulse compared to traditional forms of propulsion. These factors would reduce first-leg orbital-rocketry costs (by the source’s low weight and volume) and decrease overall travel time for astronauts with higher transit speeds. Idaho National Laboratory was provided bare, uncoated, and tungsten-coated uranium nitride fuel particles as part of a preliminary investigation. The fuel was fabricated by BWX Technologies and irradiated at the Massachusetts Institute of Technology research reactor. X-ray computed tomography and scanning electron microscopy and associated chemical analysis techniques were performed on these particles at Idaho National Laboratory. Both coated and uncoated forms were examined in the as-fabricated and as-irradiated conditions. The tungsten coating was found to be intact following the irradiation, indicating positive attributes for performance under these low-power-irradiation conditions. The irradiation of uncoated particles resulted in nitrogen enrichment at the edge of the particle.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Artificial Intelligence in Nuclear Safeguards; Evaluating Safeguards and Security Risks and Benefits for Advanced and Small Modular Reactor Deployments

Rapidly growing interest in advanced and small modular reactor (A/SMR) technologies presents challenges as well as opportunities for implementing international safeguards and security. A/SMR deployments are expected to be more numerous, more geographically dispersed, and more varied in their designs, placing new demands on the data systems and analytical tools used to support oversight (Alberti et al., 2023; Canadian Nuclear Safety Commission et al., 2024). Because of this variability, the importance and reliance on data systems for A/SMR deployments is expected to be higher than for previous reactor generations. Artificial Intelligence and Machine Learning (AI/ML) offer potential capabilities to address the high variability inherent in A/SMR technology. The beneficiaries of AI-assisted tools include facility operators, government regulators, IAEA inspectors, and A/SMR vendors. This report analyzes how AI/ML-assisted technologies can strengthen the implementation of IAEA safeguards and security measures. It also identifies AI-assisted tools to strengthen operator, facility, and regulator knowledge management practices and examines the potential risks AI/ML-based tools may introduce to IAEA safeguards and security efforts. It concludes with a set of hypothetical, standards-style requirements for AI/ML systems used in safeguards contexts, grounded in an inspector-centric view of system verification. Despite the potential benefits of AI/ML systems, understanding potential intentional and unintentional failure modes is critical for ensuring adequate protection of nuclear materials and facilities. Unique features of A/SMRs including sealed cores, remote and novel paradigms of operation, off-site reactor fabrication, novel fuel forms, and varied refueling requirements, introduce challenges for traditional safeguards technological approaches (Pensado et al., 2024; Federation of American Scientists, 2025). AI/ML systems deployed to address these challenges may introduce new risks requiring systematic evaluation rooted in both AI-specific risk frameworks, such as the NIST AI Risk Management Framework (NIST AI RMF), and established cyber risk management standards such as NIST SP 800-30 (National Institute of Standards and Technology [NIST], 2023; NIST, 2012).

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Chloride-Induced Stress Corrosion Cracking of Friction Stir-Welded 304L Stainless Steel: Effect of Microstructure and Temperature

Dry storage canisters of used nuclear fuels are fabricated using SUS 304L stainless steel. Chloride-induced stress corrosion cracking (CISCC) is one of the major failure modes of dry storage canisters. The cracked canisters can be repaired by friction stir welding (FSW), a low-heat input ‘solid-phase’ welding process. It is important to evaluate the ClSCC resistance of the friction stir welded material. Stress corrosion cracking (SCC) studies were carried out on mill-annealed base materials and friction stir welded 304L stainless U-bend specimens in 3.5% NaCl + 5 N H2SO4 solution at room temperature and boiling MgCl2 solution at 155 °C. The engineering stress on the outer fiber of the FSW U-bend specimen was ~60% higher than that of the base metal (BM). In spite of the higher stress level of the FSW, both materials (FSW and BM) showed almost similar SCC failure times in the two different test solutions. The SCC occurred in the thermo-mechanically affected zone (TMAZ) of the FSW specimens in the 3.5% NaCl + 5 N H2SO4 solution at room temperature, while the stirred zone (SZ) was relatively crack-free. The failure occurred at the stirred zone when tested in the boiling MgCl2 solution. Hydrogen reduction was the cathodic reaction in the boiling MgCl2 solution, which promoted hydrogen-assisted cracking of the heavily deformed stirred zone. The emergence of the slip step followed by passive film rupture and dissolution of the slip step could be the SCC events in the 3.5% NaCl + 5 N H2SO4 solution at room temperature. However, the slip step height was not sufficient to cause passivity breakdown in the fine-grained SZ. Therefore, the SCC occurred in the partially recrystallized softer TMAZ. Overall, the friction-stirred 304L showed higher tolerance to ClSCC than the 304L base metal.

304L austenitic stainless steel↗