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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Advancing International Integration and Strengthening Responsible Peaceful Uses of Nuclear Applications Through Specialized Curriculums

Nuclear technology has been pivotal in addressing some of the most pressing global challenges, ranging from energy production to combatting infectious diseases to agricultural security. Oak Ridge National Laboratory (ORNL), as a leader in nuclear research and development including in the field of radioisotopes, is well situated to share lessons learned and enhance global collaboration from years of discoveries in the field. Accordingly, the U.S. Department of Energy’s National Nuclear Security Administration (NNSA) sponsors specialized educational programs at ORNL, with support from the IAEA, focused on advancing peaceful nuclear applications and associated industries while upholding strong nuclear safety, security, and safeguards standards. The Joint U.S./IAEA International School on Peaceful Uses of Nuclear Applications, launched in 2024, is a cornerstone of this collaboration. The school provides an opportunity for early-career professionals from around the globe to gain practical knowledge and skills in utilizing nuclear technologies for peaceful purposes. This initiative demonstrates the United States’ commitment to its obligations under the Treaty on the Non-Proliferation of Nuclear Weapons (NPT), specifically Article IV, which calls on nuclear-weapon states to facilitate access to the peaceful uses of nuclear energy while guarding against the proliferation of nuclear weapons. In this paper, we explore the curriculum, objectives, and global impact of the program. Participants of the ICARST-2025 conference are invited to learn more about this program, contribute to its development, and explore opportunities for collaboration. This school exemplifies how strategic partnerships, and educational initiatives can drive the peaceful and beneficial use of nuclear technology worldwide

Raffo Caiado, Ana [ORNL] (ORCID:0009000239304805)↗
DOE OSTI2024

Multiscale Modeling of Silicon Carbide Cladding for Nuclear Applications: Thermal Performance Modeling

The complex multiscale and anisotropic nature of silicon carbide (SiC) ceramic matrix composite (CMC) makes it difficult to accurately model its performance in nuclear applications. The existing models for nuclear grade composite SiC do not account for the microstructural features and how these features can affect the thermal and structural behavior of the cladding and its anisotropic properties. In addition to the microstructural features, the properties of individual constituents of the composites and fiber tow architecture determine the bulk properties. Models for determining the relationship between the individual constituents’ properties and the bulk properties of SiC composites for nuclear applications are absent, although empirical relationships exist in the literature. Here, a hierarchical multiscale modeling approach was presented to address this challenge. This modular approach addressed this difficulty by dividing the various aspects of the composite material into separate models at different length scales, with the evaluated property from the lower-length-scale model serving as an input to the higher-length-scale model. The multiscale model considered the properties of various individual constituents of the composite material (fiber, matrix, and interphase), the porosity in the matrix, the fiber volume fraction, the composite architecture, the tow thickness, etc. By considering inhomogeneous and anisotropic contributions intrinsically, our bottom-up multiscale modeling strategy is naturally physics-informed, bridging constitutive law from micromechanics to meso-mechanics and structural mechanics. The effects that these various physical attributes and thermo-physical properties have on the composite’s bulk thermal properties were easily evaluated and demonstrated through the various analyses presented herein. Since silicon carbide fiber-reinforced SiC CMCs are also promising thermal–structural materials with a broad range of high-end technology applications beyond nuclear applications, we envision that the multiscale modeling method we present here may prove helpful in future efforts to develop and construct reinforced CMCs and other advanced composite nuclear materials, such as MAX phase materials, that can service under harsh environments of ultrahigh temperatures, oxidation, corrosion, and/or irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1970

Nuclear rocket applications

Nuclear rocket applications in Nuclear Flight Propulsion Module /NFPM/ and Reusable Nuclear Shuttle /RNS/

Gronich, S.↗
DOE OSTI2025

Peening Techniques for Mitigating Chlorine-Induced Stress Corrosion Cracking of Dry Storage Canisters for Nuclear Applications

Fusion-welded austenitic stainless steel (ASS) was predominantly employed to manufacture dry storage canisters (DSCs) for the storage applications of spent nuclear fuel (SNF). However, the ASS weld joints are prone to chloride-induced stress corrosion cracking (CISCC), a critical safety issue in the nuclear industry. DSCs were exposed to a chloride-rich environment during storage, creating CISCC precursors. The CISCC failure leads to nuclear radiation leakage. Therefore, there is a critical need to enhance the CISCC resistance of DSC weld joints using promising repair techniques. This review article encapsulates the current state-of-the-art of peening techniques for mitigating the CISCC in DSCs. More specifically, conventional shot peening (CSP), ultrasonic impact peening (UIP), and laser shock peening (LSP) were elucidated with a focus on CISCC mitigation. The underlying mechanism of CISCC mitigation in each process was summarized. Finally, this review provides recent advances in surface modification techniques, repair techniques, and developments in welding techniques for CISCC mitigation in DSCs.

Chemistry↗
NASA NTRS1968

Early nuclear rocket applications.

Nuclear rocket applications, discussing performance, lunar logistics, solar system missions, orbital operations and payload data

Johnson, P. G.↗
DOE OSTI2025

International database of reference gamma spectra for nuclear safeguards applications

Nuclear safeguards missions use gamma spectroscopy as a non-destructive measurement technique for examining nuclear materials. Despite advances in the development of detection equipment as well as software codes, one of the concerns is the lack of well-documented spectra needed to test and validate isotopic analysis codes for their applicability. To address this need, this work introduces IDB, an international database of the reference gamma spectra of uranium, plutonium and mixed oxide nuclear material samples. IDB provides access to well-characterized sets of gamma spectra described by rich metadata, including information on the sample, measurement configuration and detector specifications. These spectra are accessible in different formats, also compatible with analysis code standards, thus promoting their sustainability and maintenance.

None, Dipti [International Atomic Energy Agency (I↗
DOE OSTI2025

Laser Powder Directed Energy Deposition of Steels for Nuclear Applications

This comprehensive investigation examines the structure–property relationships in two nuclear alloy systems—Alloy 709 (A709) austenitic stainless steel and Grade 92 (G-92) ferritic/martensitic (F/M) steel—manufactured via directed energy deposition (DED) for sodium-cooled fast reactor applications. This study establishes the fundamental mechanisms for controlling microstructures for optimizing the mechanical performance of additively manufactured nuclear materials through systematic heat treatment optimization and multiscale characterization. As-deposited A709 steel develops a complex multiscale strengthening architecture consisting of a fine cellular solidification structure with diameter of 2-3 µm within10–50 µm grains, elevated dislocation densities from rapid thermal cycling, and grain boundary precipitates that activate concurrent Hall–Petch, dislocation, and precipitation hardening mechanisms to achieve exceptional properties [yield strength (YS): 603 MPa, ultimate tensile strength (UTS): 844 MPa, Vickers hardness: 220 HV] that achieve a 44% superior strength compared to that of the wrought material. Heat treatments produce different results. Solution annealing (SA) dissolves the cellular structure and reduces the hardness to 190 HV. Precipitation treatment (PT) keeps the cellular structure but adds carbides, allowing the hardness to reach 205 HV. The best approach combines both treatments (SA+PT) and creates uniform precipitate distributions with M 23 C 6 carbides at the grain boundaries and MX carbonitrides in the matrix, achieving a hardness of 195 HV. However, directional differences persist, with a 12%–15% strength variation between orientations due to the inherited layered microstructural architecture that survives aggressive heat treatment. While tensile testing at 550°C demonstrates 40%–50% thermal softening with dynamic strain aging, DED A709 steel still maintains a 71% higher YS than that of the wrought material. Ion irradiation studies (100–400 dpa) of DED A709 steel reveal progressive radiation damage with increasing void density and radiation-induced segregation causing nickel enrichment and chromium depletion, which will ultimately compromise mechanical properties. As-deposited G-92 exhibits exceptional strength (UTS: 1650–1700 MPa, 430 HV) through a complex microstructure containing both ferrite and martensite phases, a high geometrically necessary dislocation (GND) density (17.04×10 14 /m 2 ), and fine carbides. Heat treatments create distinct changes. Normalizing produces fresh martensite with the highest hardness (460 HV) and an increased GND density (20.23×10 14 /m 2 ). Tempering develops dual precipitation systems and reduces the hardness to 290 HV. The optimal approach uses sequential normalizing plus tempering, achieving balanced properties with the lowest hardness (250 HV) and a reduced GND density (11.01×10 14 /m 2 ). A processing-dependent anisotropy is observed: horizontal specimens achieve superior ductile behavior, while vertical specimens exhibit brittle failure. A tempering heat treatment successfully mitigates this anisotropic behavior by transforming the hard martensitic as-deposited structure into tempered martensite enabling both horizontal and vertical specimens to exhibit similar stress–strain characteristics with visible necking behavior. Remarkably, testing at 550°C reveals a reversal in the anisotropy, where as-deposited specimens achieve near isotropy with superior thermal stability (a 15%–20% strength reduction), while tempered specimens develop an orientation dependence with a 25%–30% strength reduction. Both alloy systems demonstrate that DED processing creates specimens with a superior strength through refined microstructural features, though with distinct strengthening mechanisms—austenitic through cellular structures and precipitates versus F/M through phase transformations and precipitates. Heat treatment optimization requires alloy-specific approaches, with A709 benefiting from controlled precipitation while G-92 requires careful phase transformation control. The results show that DED manufacturing can produce nuclear materials with exceptional performance, but directional effects and temperature-dependent behavior must be carefully considered for reactor component design and qualification.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Creep Property of Intermetallic Dispersive Steels for Nuclear Applications

To meet the materials performance demands of next-generation nuclear and high-temperature energy systems, a new class of intermetallic-dispersive steels (IDS) has been designed through integrated computational thermodynamics and alloy design strategies. The IDS alloys incorporate coherent L1₂ (γ′-Ni₃Al) nanoprecipitates within an Fe–Ni–Cr austenitic matrix, engineered to achieve both high temperature strength and thermodynamic stability while suppressing the formation of detrimental δ-Ni₃Nb and η-Ni₃Ti phases. Initial creep testing of Ti-IDS and TiTa-IDS alloys demonstrates rupture lives comparable to or exceeding those of Inconel 718 and ODS steels, despite being fabricated through conventional ingot metallurgy. Step-load creep tests identified a stress threshold near 300–350 MPa for the onset of tertiary creep, and in-situ neutron diffraction experiments on Ti-IDS revealed clear load partitioning between the matrix and γ′ precipitates: elastic strain is shared by both phases, while plastic strain localizes in the matrix. These results confirm that stable γ′–matrix interfaces play a dominant role in retarding dislocation motion and enhancing creep resistance. In FY26, the program will conduct repeat creep rupture testing of TiTa-IDS at 650 °C/400 MPa, initiate long-term rupture testing of TiNb-IDS, and perform TEM-based microstructural characterization to elucidate dislocation–precipitate interactions and microstructural stability. Collectively, these efforts will establish the mechanistic foundation for next-generation high-temperature structural materials based on the IDS concept.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Advanced Manufacturing Techniques and Compositions of High Entropy Alloys for Nuclear Applications

In line with the objectives of the Department of Energy, Office of Nuclear Energy, Advanced Materials and Manufacturing Technologies (AMMT) program, this work focuses on new materials development and qualification research and development for next-generation, high-temperature nuclear reactors. High entropy alloys (HEAs) have the potential to serve in these extreme environments of next-generation nuclear reactors because of their unique phase transformation pathways and nanoscale and mesoscale microstructures. The current work focuses on understanding such nuclear-energy-relevant HEAs through a detailed literature survey, selected experimental work, and developing a decision matrix with criteria for the identification of HEAs that may have the most impact and value for further examination.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗
DOE OSTI2024

Studies on Printability Methodologies and Directed-Energy-Deposition-Fabricated Iron Alloys for Nuclear Applications

This report provides results from a printability study of laser directed energy deposition (DED)-based additive manufacturing of nuclear-grade stainless steels as well as DED process parameter development for austenitic Alloy 709 (A709) and ferritic/martensitic Grade 91 (G91) and Grade 92 (G92) steels. The printability study includes the use of machine learning and physics-based modeling via commercial software such as FLOW-3D for insights into the impact of the alloy composition, particularly the carbon content, on the printability of stainless steels during the DED process. In the DED process development work, 1 cm 3 alloy blocks were deposited with broad ranges of laser powers, scan speeds, and hatch spacings to optimize the build quality, resulting in densities of more than 99.8% for all three alloys. The microstructure and mechanical properties were characterized using electron microscopy, X-ray diffraction, and Vickers hardness measurements. Further, tensile samples were extracted from DED-fabricated alloys utilizing the optimized process parameters. The present work provides guidance and progress towards the successful deployment of the DED process for the fabrication of structural components of nuclear reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Passive Temperature Sensors for Nuclear Applications

In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, this program supports basic and applied nuclear research to help address the nation's energy security needs. In support of this program, the Idaho National Laboratory (INL) established in-house capabilities to develop, fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. This effort is continuing today through the DOE?s Advanced Sensors and Instrumentation (ASI) program. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance passive sensors for irradiations where instrumentation leads cannot be included. These sensors include silicon carbide (SiC) monitors, melt wires and the sublime temperature monitor. SiC monitors are available to detect peak irradiation temperatures between 200°C and 800°C in reactor locations where instrumentation leads cannot be used. SiC monitors may be evaluated using specialized equipment installed at INL?s Measurement Sciences Laboratory (MSL). A melt wire inventory is also maintained at MSL. This inventory contains wires for specific use in irradiation experiments ranging in temperatures from 30°C to 1500°C. Melt wires and SiC monitors have had decades of research and application. Recent research has produced a passive monitor known as the sublime temperature monitor. This passive sensor has the capability of recording temperature gradients. This paper will discuss passive temperature sensors currently being researched and implemented under the ASI program.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Nuclear applications in manned space station

The zirconium hydride reactor, coupled to a thermo-electric or Brayton conversion system, and the Pu 238 isotope/Brayton system, are considered to be the viable nuclear candidates for the modular space station electrical power system. The basic integration aspects of these nuclear electrical power systems are reviewed, including unique requirements imposed by the buildup and incremental utilization considerations of the modular station. Also treated are the various programmatic aspects of nuclear power system design and selection.

Brooksbank, W. A., Jr.↗