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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

High-Entropy Borides under Extreme Environment of Pressures and Temperatures

The high-entropy transition metal borides containing a random distribution of five or more constituent metallic elements offer novel opportunities in designing materials that show crystalline phase stability, high strength, and thermal oxidation resistance under extreme conditions. We present a comprehensive theoretical and experimental investigation of prototypical high-entropy boride (HEB) materials such as (Hf, Mo, Nb, Ta, Ti)B2 and (Hf, Mo, Nb, Ta, Zr)B2 under extreme environments of pressures and temperatures. The theoretical tools include modeling elastic properties by special quasi-random structures that predict a bulk modulus of 288 GPa and a shear modulus of 215 GPa at ambient conditions. HEB samples were synthesized under high pressures and high temperatures and studied to 9.5 GPa and 2273 K in a large-volume pressure cell. The thermal equation of state measurement yielded a bulk modulus of 276 GPa, in excellent agreement with theory. The measured compressive yield strength by radial X-ray diffraction technique in a diamond anvil cell was 28 GPa at a pressure of 65 GPa, which is a significant fraction of the shear modulus at high pressures. The high compressive strength and phase stability of this material under high pressures and high temperatures make it an ideal candidate for application as a structural material in nuclear and aerospace fields.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Nanocluster Evolution in D9 Austenitic Steel under Neutron and Proton Irradiation

Austenitic stainless steel D9 is a candidate for Generation IV nuclear reactor structural materials due to its enhanced irradiation tolerance and high-temperature creep strength compared to conventional 300-series stainless steels. But, like other austenitic steels, D9 is susceptible to irradiation-induced clustering of Ni and Si, the mechanism for which is not well understood. This study utilizes atom probe tomography (APT) to characterize the chemistry and morphology of Ni–Si nanoclusters in D9 following neutron or proton irradiation to doses ranging from 5–9 displacements per atom (dpa) and temperatures ranging from 430–683 °C. Nanoclusters form only after neutron irradiation and exhibit classical coarsening with increasing dose and temperature. The nanoclusters have Ni 3 Si stoichiometry in a Ni core–Si shell structure. This core–shell structure provides insight into a potentially unique nucleation and growth mechanism—nanocluster cores may nucleate through local, spinodal-like compositional fluctuations in Ni, with subsequent growth driven by rapid Si diffusion. This study underscores how APT can shed light on an unusual irradiation-induced nanocluster nucleation mechanism active in the ubiquitous class of austenitic stainless steels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

In situ EXAFS study of Sr adsorption on TiO 2 (110) under high ionic strength wastewater conditions

In order to provide important details concerning the adsorption reactions of Sr, batch reactions and a set of both ex situ and in situ Grazing Incidence X-ray Absorption Fine Structure (GIXAFS) adsorption experiments were completed on powdered TiO 2 and on rutile(110), both reacted with either SrCl 2 or SrCO 3 solutions. TiO 2 sorption capacity for strontium (Sr) ranges from 550 ppm (SrCl 2 solutions, second order kinetics) to 1400 ppm (SrCO 3 solutions, first order kinetics), respectively, and is rapid. Sr adsorption decreased as a function of chloride concentration but significantly increased as carbonate concentrations increased. In the presence of carbonate, the ability of TiO 2 to remove Sr from the solution increases by a factor of ~4 due to rapid epitaxial surface precipitation of an SrCO 3 thin film, which registers itself on the rutile(110) surface as a strontianite-like phase (d-spacing 2.8 Å). Extended X-ray Absorption Fine Structure (EXAFS) results suggest the initial attachment is via tetradental inner-sphere Sr adsorption. Moreover, adsorbates from concentrated SrCl 2 solutions contain carbonate and hydroxyl species, which results in both inner- and outer-sphere adsorbates and explains the reduced Sr adsorption in these systems. These results not only provide new insights into Sr kinetics and adsorption on TiO 2 but also provide valuable information concerning potential improvements in effluent water treatment models and are pertinent in developing treatment methods for rutile-coated structural materials within nuclear power plants.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

High Entropy Alloys: Irradiation

High entropy alloys (HEAs) have been considered as structural materials for nuclear applications due to their promising mechanical properties and radiation resistance. For this purpose, irradiation-induced defect evolution and microstructure change have been characterized to evaluate the irradiation performance of different HEAs. This article reviews recent advances in understanding the irradiation response of HEAs, including the effect of disordered states on the defect energy landscape and defect evolution, irradiation-induced microstructure changes, void swelling, phase stability, mechanical properties under irradiation, as well as the He irradiation effects.

Zhao, Shijun↗
DOE OSTI2023

Multi-scale modeling of the evolution of structure and properties in materials for nuclear energy applications

Nuclear energy is an important component of an overall strategy to address climate change. Idaho National Laboratory (INL) is the U.S. Department of Energy’s primary facility for research and development in nuclear science and technology for energy generation, supporting the improvement and life extension of the existing reactor fleet and the development and licensing of new reactor designs. Computational modeling is an important component of these activities, particularly in the area of materials for nuclear applications, where experimental data can be very challenging and expensive to acquire, and where data is especially scarce for new reactor designs. INL has used multi-scale modeling – linking atomistic, mesoscale, and engineering scales – to improve the ability to predict the performance of materials for nuclear energy applications. These modeling efforts make extensive of MOOSE (Multiphysics Object-Oriented Simulation Environment), a general-purpose open source finite element framework developed at INL. In this talk, I will give an overview of the approach and tools used, and several examples of application, including performance of nuclear fuels, understanding radiation-driven formation of nanoscale void and gas bubble superlattices, and powder densification through electric field assisted sintering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Multi-scale modeling of the evolution of structure and properties in materials for nuclear energy applications

Nuclear energy is an important component of an overall strategy to address climate change. Idaho National Laboratory (INL) is the U.S. Department of Energy’s primary facility for research and development in nuclear science and technology for energy generation, supporting the improvement and life extension of the existing reactor fleet and the development and licensing of new reactor designs. Computational modeling is an important component of these activities, particularly in the area of materials for nuclear applications, where experimental data can be very challenging and expensive to acquire, and where data is especially scarce for new reactor designs. INL has used multi-scale modeling – linking atomistic, mesoscale, and engineering scales – to improve the ability to predict the performance of materials for nuclear energy applications. These modeling efforts make extensive of MOOSE (Multiphysics Object-Oriented Simulation Environment), a general-purpose open source finite element framework developed at INL. In this talk, I will give an overview of the approach and tools used, and several examples of application, including performance of nuclear fuels, understanding radiation-driven formation of nanoscale void and gas bubble superlattices, and powder densification through electric field assisted sintering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Status report on HFIR irradiation of optimized alumina forming alloys

Properties of FeCrAl alloys under neutron irradiation are of interest because of these materials’ potential application as accident-tolerant fuel cladding in nuclear systems. In parallel, alumina-forming austenitic (AFA) alloys are of interest for use as structural materials in advanced nuclear systems for their potential higher resistance to embrittlement and high-temperature steam oxidation resistance. An irradiation campaign for fiscal year 2024 has been developed under the Advanced Fuels Campaign to perform irradiation testing of various FeCrAl and AFA alloys in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR). The goals of this irradiation campaign are to (1) study the impact of minor alloying elements on the neutron-irradiated mechanical properties of FeCrAl alloys and (2) collect neutron-irradiated mechanical properties on AFA alloys for comparison with those of FeCrAl alloys. This campaign will include both tensile and fracture toughness specimens tested following HFIR irradiation at temperatures representative of normal operating conditions in light-water reactors. The pre-irradiation characterization to date, the irradiation plan for the FeCrAl and AFA specimens, and the subsequent post-irradiation experimental test plan are presented in this report, along with the status of HFIR builds and scheduled insertion dates.

36 MATERIALS SCIENCE↗
DOE OSTI2024

A spatially-resolved model of neutron-irradiated tungsten coupling stochastic cluster dynamics and finite deformation plasticity

Structural materials used in nuclear reactors face severe degradation in mechanical properties, such as hardening and embrittlement. At the microscopic scale, this occurs due to creation and accumulation of irradiation-induced defects and their interaction with system dislocations. Although techniques exist which can model evolution of irradiation defects, for instance kinetic transport theory-based models, their interaction with mechanical deformation of the bulk material has not been investigated extensively. In this work, we demonstrate a novel spatially-resolved multiscale coupling between microscopic irradiation defect evolution, modeled using Stochastic Cluster Dynamics (SCD) and macroscopic mechanical deformation modeled using a finite-deformation plasticity model. SCD is used to determine the statistically averaged defect cluster spacing, dependent on operating conditions such as irradiation dose and temperature. This acts as an initial condition that governs the critical resolved shear stress of dislocation glide in the macroscopic plasticity model. This framework is used to predict mechanical behavior in post-mortem test of irradiated Tungsten samples, which has found its importance as structural material used in nuclear reactors. The results obtained using the coupled approach are in good agreement with experimental data of uniaxial tension tests. The model is able to capture the effect of temperature and irradiation dose on the material hardening. Two methods are proposed to estimate hardness – using Tabor's Law relating uniaxial yield stress to hardness and from flat-punch simulations. The results are in reasonable agreement with hardness data from micro-indentation experiments of irradiated Tungsten samples. Finally, the model is also able to reveal microstructural details such as spatial variation in defect density and local stress.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Neutron Total Scattering Analysis of Materials for the Nuclear Fuel Cycle

Here, this paper reviews recent experimental efforts at the University of Tennessee and Oak Ridge National Laboratory to comprehensively characterize the structural details of materials relevant for the nuclear fuel cycle by employing advanced neutron scattering techniques. For the study of nuclear ceramics, neutron scattering offers distinct advantages over traditional laboratory or synchrotron X-ray diffraction, including enhanced sensitivity to elements with a low atomic mass, such as oxygen, nitrogen, and carbon. The key to these efforts is the recent advancement in the neutron scattering infrastructure at the high-flux diffractometers at the Spallation Neutron Source. The high neutron flux at these instruments enables neutron total scattering, a nondestructive bulk technique that simultaneously captures both short-range structural effects through pair distribution function analysis and long-range order through diffraction pattern analysis. This approach is particularly important for a comprehensive description of defective, disordered, or amorphous nuclear materials. The case studies presented here include analyses of the local defect structure in hyperstoichiometric uranium oxides and short-range order of ion-irradiated ceramics. This advanced analytical methodology will improve our understanding of the behavior of materials in extreme environments and contribute to the development of more resilient nuclear materials.

Neutron scattering↗
DOE OSTI2020

Nitrogen effects on radiation response in 12Cr ferritic/martensitic alloys

Ferritic/martensitic steels are one of the best candidates for structural materials for high dose applications in next generation nuclear reactors. The composition of structural materials must be optimized for reliable service during irradiation. This study reports the effect of interstitial nitrogen on radiation response in 12Cr ferritic/martensitic HT9 steels having a controlled nitrogen concentration. Results show that a high amount of ‘free’ nitrogen in the matrix stabilizes the interstitial clusters which leads to (i) larger loop sizes (ii) lower loop density and (iii) slightly reduced radiation induced hardening. It also affects diffusion mechanism of Ni and formation of Ni/Si-rich precipitates.

36 MATERIALS SCIENCE↗
NASA NTRS2008

Structural Materials and Fuels for Space Power Plants

A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

Bowman, Cheryl↗
DOE OSTI2022

Capability Needs for Irradiated and Radioactive Materials Research (Ad hoc Committee Summary Report)

It is essential for the U.S. nuclear energy community to have access to world-leading equipment suitable for conducting research on both nuclear fuels and structural materials, including neutron irradiated–and therefore activated–materials. To address future research infrastructure requirements to support the DOE-NE mission, the Office of Reactor Fleet and Advanced Reactor Deployment established an ad hoc committee to gather information on potential capability gaps for radioactive materials and radiation effects research. This document summarizes the discussions of the committee addressing the high-level challenges in irradiated and radioactive materials research and the capabilities needed to address these challenges. After considering the various needs, the committee agreed on the four top-level targets and priority capability gaps.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Plutonium oxide melt structure and covalency

Advances in nuclear power reactors include the use of mixed oxide fuel, containing uranium and plutonium oxides. The high-temperature behaviour and structure of PuO 2-x above 1,800 K remain largely unexplored, and these conditions must be considered for reactor design and planning for the mitigation of severe accidents. Here, in this study, we measure the atomic structure of PuO 2-x through the melting transition up to 3,000 +/- 50 K using X-ray scattering of aerodynamically levitated and laser-beam-heated samples, with O/Pu ranging from 1.57 to 1.76. Liquid structural models consistent with the X-ray data are developed using machine-learned interatomic potentials and density functional theory. Molten PuO 1.76 contains some degree of covalent Pu-O bonding, signalled by the degeneracy of Pu 5f and O 2p orbitals. The liquid is isomorphous with molten CeO 1.75 , demonstrating the latter as a non-radioactive, non-toxic, structural surrogate when differences in the oxidation potentials of Pu and Ce are accounted for. These characterizations provide essential constraints for modelling pertinent to reactor safety design. The molten structure of plutonium oxide-a component of mixed oxide nuclear fuels-is measured, showing some degree of covalent bonding. Its atomic structure is similar to that of cerium oxide, which could be a non-radioactive structural surrogate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A new method for quantifying 64 Cu in nuclear debris samples

Quantifying 64 Cu in post-detonation nuclear debris samples can provide important diagnostic information regarding the structural materials used within a nuclear device. However, this task is challenging due to the weak gamma emissions associated with the decay of 64 Cu, its short half-life (12.701 h), and the presence of interfering fission product radioisotopes. Large quantities of debris sample are generally needed to accurately quantify 64 Cu, which can be problematic in sample-limited scenarios where other radiometric analyses are required. Herein, we present a new method for the separation of 64 Cu from solutions of mixed fission products and demonstrate the quantification of its activity through use of gas-flow proportional beta counting. Here, the new method was validated through a series of rigorous tests and was shown to improve the detection limit of 64 Cu by over two orders of magnitude, from 2.5 × 10 6 to 1.3 × 10 4 atoms/sample for 100 min measurements.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

A robust alloy design (RAD) strategy for next-generation (IV) nuclear fission reactors

Next-generation nuclear reactors demand structural materials capable of withstanding extreme conditions, including high temperatures, intense neutron flux, and corrosive environments. Multi-Principal Element Alloys (MPEAs) have emerged as promising candidates due to their exceptional radiation tolerance, thermal stability, and compositional flexibility. This study introduces a versatile and customizable Robust Alloy Design (RAD) strategy for systematically designing MPEAs for GEN-IV reactor fuel cladding. The RAD framework integrates nuclear-relevant selection criteria, empirical parameter assessments, and high-throughput CALPHAD simulations to efficiently narrow compositional space and identify stable alloys. A unified RAD score developed for the first time, combines key performance metrics, including fuel-clad chemical interaction (FCCI), neutron absorption cross-section (NAC), valence electron configuration (VEC), and melting point factor (MPF), into a flexible ranking system adaptable to reactor-specific priorities. Among 724 candidates, V555(5Al–5Cr–5Fe–85V) emerged as the top alloy, validated experimentally with a homogeneous single-phase BCC microstructure and superior mechanical properties (nano-indentation: 3.389 ± 0.258 GPa; Vickers hardness: 240 ± 6.7 HV), significantly outperforming Zircaloy-4 and V-4Cr-4Ti. Importantly, the RAD strategy is not limited to nuclear applications; its customizable weighting system enables scalability to other extreme environments. This adaptability positions RAD strategy as a versatile tool for advanced materials design across multiple industries.

Alloy design↗
DOE OSTI2024

Elucidating the role of Cr migration in Ni-Cr exposed to molten FLiNaK via multiscale characterization

Structural materials used in nuclear reactor environments are subjected to coupled extremes such as high temperature, irradiation, and corrosion which act in concert to degrade their functional performance. Connecting alloy microstructure such as grain boundaries, and accumulating point defects with corrosion attack and pore morphology is critical to understanding underlying mechanisms. We uncover the compositional variations and morphology at multiple length scales in corrosion-damaged Ni-Cr alloy after exposure to oxidants in molten fluoride salts. A complex network of dense corrosion pores is detected by surface-level SEM observations. The corrosion pores take on a 1-dimensional morphology and are enriched with Ni and depleted of Cr 1–5 µm from the pore surface. STEM-EDS and 4D-STEM strain maps acquired simultaneously highlight the local variations in composition and structure of a ≤ 200 nm Cr-rich layer identified from a cross-section taken at the bottom of an isolated corrosion pore between the Ni-Cr alloy matrix and the embedded salt. However, the absence of an observed interface between the Ni-Cr alloy matrix and the FCC-structured Cr-rich layer suggests that Cr plating from the salt did not transpire. These findings support a proposed Cr lattice diffusion mechanism rather than Cr-precipitation from the salt to accommodate temperature transient conditions during sample cooling. Through the development of a 1D phase field model, these results are rationalized by formation energies for the Ni- and Cr-oxidation into the molten salt. This study reveals the locally altered microstructure caused by high temperature corrosion in non-steady-state molten salt nuclear reactor environments.

36 MATERIALS SCIENCE↗