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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 271 records · Page 15
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DOE OSTI2026

Creep in multi-principal element materials –– A review

The ongoing push towards enhanced energy efficiency and reduced emissions has necessitated the creation of materials with superior performance, especially under extreme conditions. Modern industries, such as aerospace, energy production, and nuclear power, rely heavily on materials that can withstand elevated temperatures without compromising structural integrity. At these heightened temperatures, materials, even when subjected to mechanical stresses well below their yield strength, may experience slow deformation leading to eventual rupture — a phenomenon known as creep. With the expansive design space that comes with the high entropy concept and their reported excellent high temperature strength, multi-principal element materials (MPEMs) have attracted interest in the scientific community for high-temperature applications. Here, this review offers a comprehensive examination of existing studies on creep in MPEMs, which includes multi-principal element−alloys, −bulk metallic glasses, −ceramics, and −superalloys, comparing published findings on MPEMs with pure elements, traditional alloys, bulk metallic glasses, and superalloys. The sub-topics covered include a comparison among different creep-testing methods, creep mechanisms, creep exponents, creep strain rates, activation volume, and creep-activation energy. Modeling efforts for predicting creep behavior of MPEMs are also reviewed. Methods for improving creep resistance by performing heat treatments and/or modifying microstructures are discussed. Overall, the current state of MPEMs has not yet surpassed the creep performance of commercial alloys. Finally, directions for future efforts are suggested, such as experimenting in various controlled environments, expanding the number of compositions tested, exploring advanced manufacturing techniques, and using machine-learning to predict creep properties based on compositions and microstructures.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal neutron scattering cross sections for amorphous carbon

Carbon materials are commonly found in both nuclear reactors and experimental systems. Various carbon structures occur in nuclear applications ranging from crystalline and nuclear graphite to the amorphous carbon seen in next-generation advanced reactor designs. Amorphous carbon is based on a randomized graphite-like structure and offers the unique ability to disperse impurities throughout the bulk composition. A graphite-like amorphous carbon system was modeled using the classical molecular dynamics (MD) code LAMMPS (Large-scale Atomic/Molecular Massively Parallel Simulator). An improved version of the temperature-dependent Adaptive Intermolecular Reactive Empirical Bond Order (AIREBO) potential was used to model the carbon-carbon atomic interactions for the temperature at 300 K along with densities 1.60, 1.70, 1.85, and 2.23 g/cm{sup 3}. From the normalized velocity autocorrelation function (VACF), the phonon density of state (DOS) was then calculated as the Fourier transform of the normalized VACF. This DOS was then used as the primary input for the evaluation of the thermal scattering law (TSL, i.e. S(α,β)) and associated neutron thermal scattering cross sections. The TSL was analyzed using the Full Law Analysis Scattering System Hub (FLASSH). The amorphous structure results in shifts of the phonon DOS to lower energy modes than typically displayed for ideal crystalline graphite. This impact on the DOS is directly reflected in the TSL. Furthermore, the typical optical peak at 0.25 eV for the ideal graphite disappears for amorphous carbon, in good agreement with the expected structure. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Addendum to Capability Needs for Irradiated and Radioactive Materials Research (Ad Hoc Committee Summary Report)

Nuclear materials and fuels studies are challenging research and engineering targets because of their inherent radioactivity as well as their heterogeneous microstructure. Advanced light sources have made significant underpinning scientific contributions to the understanding of structural and fuel cladding material microstructure and properties and the development of structure-property relationships. These facilities have the potential to help advance the Department of Energy (DOE) Office of Nuclear Energy (DOE-NE) mission priorities providing underpinning understanding of technically important challenges, including: irradiation induced embrittlement and swelling; stress corrosion cracking and corrosion in extreme environments; ageing of reactor pressure vessels steels; nuclear fuel characterization; and nuclear waste form optimization as well as playing a major role in the development of a materials performance matrix to aid in the design of new materials to meet the needs of advanced reactor concepts. However, the delivery of underpinning understanding is the mission of DOE Office of Science (DOE-SC). While filling such knowledge gaps is important, efforts in this direction should not be confused with the strategic technology focused mission goals of DOE-NE to enable the continued operation of existing U.S. nuclear reactors, to enable the deployment of advanced nuclear reactors, to develop advanced nuclear fuel cycles, and to maintain U.S. leadership in nuclear energy technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Aerosol Measurement from Plasma Arc Torch Cuts on Various Metal for D and D Workers' Internal Dose Calculation - 20253

When dismantling a nuclear power plant, structures must be cut to reduce size. The cutting of radioactive material produces radioactive aerosols, which pose a health risk including radiation risk to site workers and the general public. However, due to a lack of research on the physical and chemical characteristics of aerosols produced in these processes, the calculation of internal dose in the event of an expected or unexpected accident is inaccurate. For the measurement of aerosols generated from cutting for internal dose calculations, we measured the number concentration distribution and nuclide distribution of aerosols from plasma arc cutting in a laboratory experimental setup, considering isokinetic sampling and automatic cutting. The results may suggest internal dose assessment without characteristics of radioactive aerosol has uncertainty and the possibility of being undervalued. (authors)

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

Expanding the chemistry of borates with functional [BO 2 ]- anions

More than 3900 crystalline borates, including borate minerals and synthetic inorganic borates, in addition to a wealth of industrially-important boron-containing glasses, have been discovered and characterized. Of these compounds, 99.9 % contain only the traditional triangular BO 3 and tetrahedral BO 4 units, which polymerize into superstructural motifs. Herein, a mixed metal K 5 Ba 2 (B 10 O 17 ) 2 (BO 2 ) with linear BO 2 structural units was obtained, pushing the boundaries of structural diversity and providing a direct strategy toward the maximum thresholds of birefringence for optical materials design. 11 B solid-state nuclear magnetic resonance (NMR) is a ubiquitous tool in the study of glasses and optical materials; here, density functional theory-based NMR crystallography guided the direct characterization of BO 2 structural units. The full anisotropic shift and quadrupolar tensors of linear BO 2 were extracted from K 5 Ba 2 (B 10 O 17 ) 2 (BO 2 ) containing BO 2 , BO 3 , and BO 4 and serve as guides to the identification of this powerful moiety in future and, potentially, previously-characterized borate minerals, ceramics, and glasses.

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

Fabrication and analysis of hydroxide/ceramic dual-phase (HCDP) membranes

Ceramic materials show potential to serve as support structures for membranes that can be used to uptake carbon dioxide from high temperature environments, like those found in nuclear flue plants. The porosity of ceramic materials allows for the infilling of those pores with molten hydroxides, that show a natural affinity for absorbing CO 2 from its surroundings. The membranes are easy to produce and were made with thickness of 0.5mm, 1.0mm, and 2.0mm. The membranes were exposed to various conditions after the infilling process and their responses were measured. The percent of hydroxides were measured after the infilled membranes had been in a tube furnace at 400°C, followed by exposure to 30 minutes of CO 2 , and lastly 4 hours of steam. Those tests yielded mole percent hydroxide results of 88.4% ± 4.6%, 68.6% ± 5.1%, and 80.7% ± 4.6%, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Technical report Letter: RAFM, ODS steels and MMLC for Nuclear energy application

The lifetime, thermodynamic efficiency, safety and economic viability of new generation fission and fusion reactor concepts can largely be tied to the mechanical performance and stability of structural alloys under extreme environments. In this context, engineered nano materials could have broad-reaching impact on the future of advanced nuclear fuel-cycle and reactors. These systems are characterized by a large number density of interfaces which are efficient sinks for point defects and moderately biased; therefore limiting the deleterious effects of irradiation. Broadly, nuclear nano-technology deals with the use of the latest engineered-nanomaterials for improving the nuclear power performances and safety in all areas of nuclear energy production to bring new generations of nuclear power units. New advanced fuel assembly designs also have implications for securities and safeguards. To support the readiness for potential future license applications, an understanding of the technologies that would enable new reactor designs in the areas of component performance and domestic safeguards is necessary. This technical report letter work explores the technical issues and potential regulatory considerations associated with developing and adopting fuel claddings made of advanced nano- materials. Specifically three classes of nanomaterials are considered: (i) reduced activation ferritic/martensitic (RAFM) steels, (ii)oxide dispersed steels (ODS) and (iii) multi-metallic layered composites (MMLC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Summary of Development for Structural Component Modeling in Fiscal Year 2022

This report summarizes efforts performed during Fiscal Year 2022 to develop capabilities for modeling structural component degradation in support of the U.S. Department of Energy's Nuclear Energy Advanced Modeling and Simulation Program. These efforts were centered around development of capabilities for the Grizzly code. Efforts focused both on foundational engineering-scale analysis capabilities for damage and shell formulations and on material-scale tools for accounting for irradiation effects in Grade 91 steel. A major outcome of this effort was the development of a general-purpose tool for nonlocal averaging of material properties, which enabled nonlocal damage models. In addition, the applicability of the shell elements in MOOSE has been expanded to allow modeling a wider range of component geometries and using a wide variety of material models. For irradiation effects on materials, a previously developed cluster dynamics model for light water reactor pressure vessel steels has been adapted for application to Grade 91 alloy. This work builds on prior efforts to build a flexible, capable code for addressing a variety of aging and component performance issues in nuclear power plant structural components.

36 MATERIALS SCIENCE↗
DOE OSTI2025

In-situ ion irradiation induced nanograin growth in a spent UO 2 fuel

This study investigates the irradiation-driven evolution of nanograins in the early-stage restructured rim region of medium burnup spent uranium dioxide (UO 2 ) fuel. Transmission electron microscopy (TEM) lamellas prepared from Belgium Reactor 3 (BR-3) fuel were subjected to in-situ 300 keV Xe ion irradiations under varying doses, fluxes, and temperatures to evaluate their effect on the evolution of the nanograins. Our results reveal that the nanograins grow during ion irradiation. Additionally, the growth is most pronounced at elevated temperatures (300 °C), moderate at room temperature, and negligible at cryogenic temperature (−223 °C). This behavior indicates that thermal activation, alongside irradiation effects, is essential to overcome grain boundary pinning by fission gas bubbles, metallic precipitates, and porosity. Furthermore, while nanograins (<200 nm) consistently coarsened under irradiation, larger grains did not undergo further restructuring, which can be attributed to the strong defect annihilation at TEM lamella surfaces combined with the limited electronic stopping power of low energy Xe ions used in this work. These findings highlight the roles of thermal spike effects, defect mobility, and impurity pinning in governing grain evolution in the rim region of spent UO 2 fuel during ion irradiation, providing key insights for predictive models of restructuring and performance of the high burnup nuclear fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Effect of charge and anisotropic diffusivity on ion exchange kinetics in nuclear waste form materials

Motifs of radionuclide absorption particles (absorbers or fillers) for the nuclear waste treatment usually have crystal structures with nano-sized tunnels for fast ion diffusion. On the other hand, the accumulation of ions generates an electric field which affects the ion diffusion. Therefore, the strong anisotropic kinetic properties and the electric field affect ion exchange kinetics, especially in an assembly of these particles which may block the fast diffusion tunnels for each other. In this work, we developed a mesoscale model to describe the effect of anisotropic kinetic properties and electric field on ion exchange kinetics. With the model we simulated the effect of anisotropic diffusivity, electrochemical potential, particle morphology, size and aggregation on ion exchange kinetics during batch experiments. Here, tt is found that 1) ion exchange results in charge accumulation inside the particle because different ions have different diffusivities and different diffusion driving forces. The electric neutrality is not preserved; 2) the electric field speeds up the slower diffusion ions (either uptake or release) and slows down the faster diffusion ions to reduce charge accumulation and maintains charge neutrality; 3) the ion exchange becomes very slow at the later stages because a) diffusion driving force (electrochemical gradient, Cs + concentration in the solution and Na + inside the particle) continuously decreases, as observed in batch experiments; and b) the diffusivity decreases due to the Cs + concentration dependence of ion diffusivity. This can explain the observed experimental phenomena that the radionuclide capacity usually cannot be reached; 4) for needle-like particle and their clusters, both surface area and diffusion distance along the direction with the fastest diffusivities have significant impact on the ion exchange kinetics.

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

SISGR: Chemomechanics of Far-From-Equilibrium Interfaces (COFFEI)

Portable, reliable, and deployable devices for energy storage and conversion require fundamental changes in design of solid-state composites comprising ceramics and metals. These materials comprise the electrodes and electrolytes of next-generation solid-oxide fuel cells and all solid-state batteries, forming solid-state functional composites. The advent of solid-state batteries – which replace liquid electrolytes with solid electrolytes capable of lithium ion transport for reliable energy storage in portable batteries – and the increased demand space for all solid-state fuel cells capable of oxygen reduction at intermediate temperatures remain important challenges for improved material stability and decreased system cost. However, little is understood about three fundamental facets of materials that enable such solid-state energy applications. First, how do such materials deform, fracture, or delaminate under operando conditions? Second, how does such mechanical deformation limit or facilitate electronic and ionic transport within and across such material interfaces? Third, how we can predictably design interface-rich composites to engineer both structural and electrochemical stability? This COFFEI Group comprised expertise from Materials Science & Engineering and Nuclear Science & Engineering to integrate unique in situ experiments, simulations, and fabricated interfaces that address these fundamental questions in solid-state interfaces of nanoscale composites that will guide solid-state electrochemistry, transport kinetics, and mechanical deformation for nonstoichiometric materials that enable such applications. In particular, we built on COFFEI’s understanding of chemomechanical coupling among defect concentrations, ionic transport, electron transport, and stored elastic energy that is particularly acute in the far-from-equilibrium conditions typical of energy device applications. By tailoring our focus to solid-state interfaces, we addressed these important issues by (a) developing and applying advanced in situ and ex situ characterization tools to characterize model materials and interfaces synthesized with molecular-level control, under both laboratory-controlled and extreme environments representative of energy device operation; and (b) employing computational modeling and simulation frameworks to predict transport mechanisms, reactivity and stability of these model materials and interfaces under significant chemical strains typical of energy device operation. Recent progress provided insights to additional materials systems and electrochemomechanical fatigue and fracture that were not fully envisioned when the program was initiated. Specifically, in the final three years of COFFEI we pursued two integrated thrusts, with complementary focus. Thrust I focused on failure-resistant electrochemomechanical composites, while Thrust II focused on strain-modulated conductivity and reactivity across interfaces. In contrast to our initial COFFEI focus, these thrusts concentrated wholly on solid-state material interfacial interactions and included greater integration of multiscale visualization including in situ electron microscopy of strained structures/interactions and mesoscale simulations. Successful development of functionally superior and long-lived battery and fuel cell systems and stress adaptable oxides requires a deeper, fundamental understanding of the coupling among the historically important subfields of solid-state electrochemistry, transport kinetics, and mechanical deformation for nonstoichiometric metal oxide electrodes. In this program, the understanding and the application of chemomechanical coupling of defect concentrations, ionic transport, electro-catalytic activity and stored elastic energy, particularly acute in the far-from-equilibrium conditions typical of energy device applications, are being refined and implications for device operation clarified, including for miniaturized solid-state batteries and fuel cells.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Twentieth Exotic Beam Summer School (EBSS2023)

The study of unstable nuclei with unusual ratios of protons to neutrons is one of the frontiers of science. Investigating these rare isotopes is critical for understanding the synthesis of the chemical elements in stellar explosions as well as the fundamental nature of the nuclear forces that bind atomic nuclei together. Scientific progress in this field is driven by the development of exotic beams in present and next-generation rare-isotope beam facilities including the Facility for Rare Isotope Beams (FRIB). Nuclear physics is a broad discipline, influencing our knowledge on subjects as diverse as weakly-bound nuclei, many-body quantum theory, the super heavy elements, and the inner structure of neutron stars. Applications based on nuclear science and technologies include medical diagnostics and therapies, materials science, and national security. The major goal achieved in this project was to hold Exotic Beam Summer School 2023 (EBSS2023), the twentieth installment of EBSS series, July 9-15, 2023 at the Facility for Rare Isotope Beams on the campus of Michigan State University to educate and train the next generation of scientists that will drive research with rare-isotope beams. FRIB became operational in 2022 and is now providing beams of exotic nuclei that will ramp up to unmatched intensities, exceeding what is available today by orders of magnitude. Beams available at FRIB are facilitating a wide variety of studies in nuclear structure, astrophysics, fundamental symmetries and societal applications. There is a large community of scientists interested in working with rare isotope beams; for example, the FRIB User Organization currently has over 1,700 members. In order to maximize the scientific output of FRIB, there must be a workforce continuously trained in both the physics of exotic beams and in the practical techniques of carrying out an experiment. This summer school series is designed to specifically address this need - to ensure that new generations of scientists from a broad range of institutions and backgrounds is trained, motivated, and equipped to push the field forward to new and important breakthroughs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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 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 OSTI2021

Seismic Performance of Deeply Embedded Conceptual Advanced Reactors: Three-Dimensional Nonlinear Soil-Structure Interaction Analyses

The U.S. Nuclear Regulatory Commission’s functional containment concept provides advanced nuclear power plant designers with more flexibility in terms of the civil/structural design if the appropriate set of barriers for prevention of radioactive material release exist. Some of the conceptual advanced reactor structures, without the traditional pressure boundaries of large containment structures, are proposed to be deeply embedded or buried into soil. This approach is expected to provide (1) lesser seismic demands on the structures and safety-critical structures, (2) eased regulatory efforts and overall design against other external hazards such as aircraft impact, and (3) overall cost savings. Further, one of the important aspects of assessing the technical and economic viability of deeply embedding advanced reactor buildings is to assess the seismic performance with the understanding of effects with material and geometric nonlinearities. This study investigates the seismic response of deeply embedded or buried advanced reactors by conducting three-dimensional nonlinear soil-structure interaction analyses. Although the results indicate that there is a general trend of decreased seismic response with increased embedment depths, the change in the dynamic environment with different embedment depths and the nonlinear environment under high-intensity seismic inputs may result in increased peak response at increased embedment depths.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Unfaulting mechanisms of interstitial Frank loops in fluorite-structured ThO 2

Unfaulting of Frank loops in irradiated fluorite-structured oxides is of significance for understanding microstructural evolution. However, the mechanisms have not yet been directly observed. To this end, we utilize molecular dynamics to reveal the atomistic details related to the unfaulting process of interstitial Frank loops in ThO 2 , which involve a single pair or multiple pairs of co-propagating Shockley partials to annihilate the stacking faults. Here, we find that the unfaulting is achieved via a synchronous shear of the partial pairs to remove the extrinsic stacking fault in the cation sublattice and the intrinsic stacking fault in the anion sublattice. The high mobility of oxygen at the dislocation core may reduce the activation barriers of dislocation nucleation and migration. These findings provide a fundamental understanding of the transformation of faulted loops in irradiated ThO 2 , and could be transferable to other fluorite-structured oxides.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Long-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer periods of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) initiated a collaboration with the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program on the Advanced Radiation Resistant Materials (ARRM) program, focused on developing and test degradation resistant alloys for light water reactor (LWR)- relevant environments. Based on a comprehensive microstructure and property screening, the ARRM program selected a total of five alloys (i.e., Grade 92, 310, 690, 718A, and 725), together with 316L and X-750 as references, for further investigations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗