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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Implications of point defect accumulation on UO 2 thermal conductivity and fission gas release under accelerated fuel irradiation

Evaluation of thermal properties is a crucial factor for nuclear fuel performance. During reactor operation, the accumulation of fission products and irradiation-induced lattice defects are responsible for degradation in thermal conductivity. Consequently, it affects fuel temperature and fission gas release (FGR) among other Multiphysics processes important for economics and safety analysis. We analyze the implications of point defects (PD) accumulation described using a rate theory (RT) Model on lattice thermal conductivity of UO 2 . Here, we demonstrate that fission rate-dependent point defect concentrations have the largest impact on in-pile thermal conductivity in the periphery of light water reactor fuels below a temperature threshold governed by the migration barrier of defects. Our analysis provides a mechanistic description of this phenomena which current fuel performance codes treat empirically. The reduction of thermal conductivity in the low -temperature rim region acts as additional thermal resistance and leads to a temperature notably larger than suggested by Lucuta thermal conductivity correlation. These effects are anticipated to have notable impacts when fuels are exposed to accelerated radiation. The impact of such point defect-informed treatment of thermal conductivity on fuel performance is evaluated by a detailed analysis of fission gas behavior and its release. We consider several models capturing different stages of fission gas bubble evolution and fission gas release (FGR). Finally, a new fission rate-dependent correction to the Lucuta correlation is proposed. The results show a significant reduction in thermal conductivity at the fuels’ periphery and an increase in fuel centerline temperature specifically at low burnups. Ultimately a modified LC shows a higher FGR compared to the original LC, while the acceleration process results in a reduction in overall FGR.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

In-situ ETEM study of plasma-facing tungsten nanofuzz oxidation at atmospheric pressure: Microstructure evolution and substrate-free oxidation kinetics

To enable sustainable carbon-free fusion energy, managing reactor structural material degradation during normal operation as well as accident scenarios is vital. Tungsten (W) plasma-facing materials (PFMs) are susceptible to aggressive high-temperature oxidation during air-ingress fusion reactor accidents, yet there's a lack of oxidation kinetic data for irradiated tungsten. Here, in this study, we utilize atmospheric environmental transmission electron microscopy (ETEM) to present the first kinetic data for substrate-free W nanofuzz oxidation at 400 ºC and 500 ºC in 1 bar dry air. Comparison with pristine bulk W during the early parabolic stage suggests an irradiationdecelerated oxidation for W nanofuzz. Our time-resolved in-situ characterization reveals a durable amorphous surface oxide, likely promoted by high-flux He + irradiation-induced surface defects, serving as an effective passivating layer that impedes nanofuzz oxidation onset. This surface oxide layer also interfaces well with newly formed orthorhombic WO 3 , facilitated by stress relief through He bubble shrinkage, providing lasting passivating protection throughout the nanofuzz parabolic oxidation. This new finding challenges conventional notions of irradiation's negative impact on metal oxidation, and calls for advanced characterization to enhance our understanding of fusion energy materials degradation, informed by further accident modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Radiation-induced vacancy injection in heterogeneous multiphase materials

Understanding the synergy between corrosion and defect dynamics is a key consideration in the development of advanced materials for extreme environments. Here, we reveal a surprising phenomenon for the transport of point defects induced by irradiation in a heterogeneous multiphase structure of a metal and an oxide, similar to that formed under metal corrosion that takes place in most environments. Despite the confinement of the produced damage within the oxide, vacancies were injected into the unirradiated metal and coarsened with dose. Furthermore, the results show that the nature of the oxide layer dictates the defect evolution in the metal layer. This work reveals an interesting mechanism for point defect interactions in multiphase materials, with broad implications in many fields, while also emphasizing the complex coupling between corrosion and irradiation. Corrosion leads to the formation of multiphase materials, while irradiation enhances diffusion within the heterogeneous phases, which can impact corrosion rates.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Thermal conductivity measurement of U-Mo and U-Mo/Al interaction layers generated from in-pile irradiation using the suspended-bridge method

Here, this study presents the first measurement of the individual thermal conductivity of U-7wt.%Mo fuel particles and U-Mo/Al interaction layers (ILs) from in-pile irradiated dispersion fuel plates, using the suspended-bridge method. Nanorods of U-7wt.%Mo fuel and U-Mo/Al ILs were extracted by focused ion beam (FIB), and their microstructures were characterized with transmission electron microscopy (TEM). TEM revealed finely distributed nanobubbles in the U-7wt.%Mo matrix, along with an amorphous structure in the ILs. The thermal conductivity of in-pile irradiated U-7wt.%Mo was approximately 30% lower than that of the unirradiated material, ranging from 6.7 W/m·K at 300 K to 8.5 W/m·K at 380 K. The ILs exhibited even lower thermal conductivity, from 2.1 W/m·K at 300 K to 2.7 W/m·K at 380 K. These reductions, attributed to nanobubbles, fission products, and irradiation-induced point defects, were analyzed through a combination of microstructural characterization and literature-based transport models, which successfully reproduced the observed degradation trends.

42 - ENGINEERING↗
DOE OSTI2021

Summary of Previous Mechanical Test Data on ODS Alloys 14YWT and OFRAC up to 1000ºC

The Nanostructured Ferritic Alloys (NFA) 14YWT and OFRAC were developed for future fission and fusion nuclear energy reactors requiring high-temperature mechanical properties that are tolerant to extreme neutron irradiation environments. The NFA contain a high concentration of Ti-, Y- and O-enriched nanoclusters (NC) and ultra-fine grains to achieve high temperature strength and creep properties and high sink strength for trapping irradiation induced point defects to minimize hardening and swelling and transmutated He atoms to form intragranular nano-size bubbles that prevent formation of coarse bubbles on grain boundaries that cause embrittlement. The mechanical properties of 14YWT and OFRAC have been acquired from tensile and creep tests conducted in the past at Oak Ridge National Laboratory. The development of 14YWT started in 2000, resulting in the production of numerous heats. Tensile properties were obtained from eleven heats of 14YWT from room temperature to 800ºC. Tensile data for the SM10 heat of 14YWT was extended to 1,000ºC. Initial development of OFRAC occurred in 2016. Tensile tests conducted from room temperature to 800ºC revealed similar properties of OFRAC with those obtained from the newer generation of 14YWT heats. The creep properties of 14YWT-SM10 evaluated using constant stress tensile and load time-to-failure tests at 800ºC showed low minimum creep rates with stresses of 200 and 100 MPa. The single time-to-failure test of 14YWT-SM10 at 800ºC and 100 MPa was terminated after 20,357 hours with no specimen failure and a low creep strain of ~0.24 %. The creep properties of OFRAC determined from strain-rate jump tests at were similar those of 14YWT-SM10. The stress exponent for 14YWT and OFRAC at 800ºC are similar and are consistent with threshold stress behavior. Since both 14YWT and OFRAC are candidates for fuel cladding in future fast reactors, several fabrication studies were recently conducted and have successfully demonstrated that thin wall tubes can be fabricated from 14YWT and OFRAC by cold pilger rolling and high precision tube rolling. The high temperature mechanical properties and feasibility of fabricating thin wall tubes make 14YWT and OFRAC candidates for application as heat pipes in advanced micro-reactors. The purpose of this report is to summarize the previously obtained tensile and creep data at temperatures up to 1000ºC for NFA 14YWT and OFRAC that have been acquired over the past 20 years at ORNL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Nanostructured Alumina Forming Austenitic Alloy (NAFA) Production Using Mechanical Alloying and High-Temperature Consolidation

Alumina-forming austenitic (AFA) alloys are a promising class of nuclear materials because of their high-temperature oxidation/corrosion resistance and mechanical properties. Unfortunately, these alloys are limited in use for core material applications, and they are specifically limited for use as nuclear fuel cladding because of their high Ni-transmutation and helium generation rate in-service. Improving these alloys through a fine dispersion of oxide precipitates and thus increasing the effective irradiation sink strength of the alloy system may mitigate many of the degradation phenomena expected during alloy deployment. These phenomena include high-temperature helium embrittlement and cavity swelling for lead-cooled fast reactor applications. This work effort uses combination of conventional and advanced manufacturing approaches are being used to fabricate nanostructured AFA (NAFA) materials. As the first objective of this initiative, a conventional AFA was modified using mechanical alloying and extrusion to alter the precipitation characteristics to include a fine dispersion of nanoscale oxides intended to serve as traps for irradiation-induced point defects and transmuted He within the lattice. This analysis compared the efficacy of the conventional mechanical alloying and extrusion approach with the unalloyed AFA consolidated approach using hot isostatic pressing (HIP). It was found that, although the mechanical alloying approach is successful in producing a fine distribution of oxides within the first nanostructured AFA (NAFA-1), the distribution is heterogeneous because of the mild milling parameters used to prevent cold welding of powder to the spherical milling media. The additional dispersion of oxides, coupled with a higher volume fraction of other secondary phases in the NAFA-1, produces higher alloy strengths that range up to 600°C in comparison to the unalloyed HIP AFA, thus exceeding the operating temperature of lead-cooled fast reactors. However, the strength of the NAFA-1 is lower than that of the HIP AFA at 800°C, which is presumed to be a function of increased secondary phases from the nonoptimized AFA chemistry. Future work is planned on a newly procured NAFA-2 chemistry that is more suitable for advanced reactor applications exploring new manufacturing routes.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Irradiation-Induced Structural Disorder and Its Influence on the Mechanical Response of Polycrystalline MoS2

Molybdenum disulfide (MoS2) thin films are widely used as dry-film lubricants and protective coatings in aerospace and other radiation-exposed environments. Conventional synthesis routes produce polycrystalline films whose grain boundaries and other native defects cause their mechanical and tribological behavior to differ substantially from that of ideal single crystals. Under irradiation, these films progressively evolve from polycrystalline structures, composed of layered MoS2 grains, into highly disordered and eventually amorphous structures, altering both their tribological performance and mechanical integrity. Here, we employ reactive atomistic simulations to investigate irradiation-driven structural evolution in bulk polycrystalline MoS2. Using controlled primary knock-on atom (PKA) events, we characterize the progressive transition from a polycrystalline microstructure to an amorphous network by tracking defect accumulation and structural disorder. We then establish how this transition modifies the dominant deformation mechanisms and the temperature-dependent tensile response. Specifically, irradiation suppresses interlayer sliding and delamination, mechanisms which facilitate the deformation of the pristine polycrystal, resulting in defect-induced hardening. Broadly, our results establish direct process–structure–property relationships linking irradiation-induced defect accumulation, microstructural evolution, deformation mechanisms, and mechanical behavior, providing an atomistic framework for understanding the structural integrity and long-term reliability of irradiated MoS2 coatings.

Moore, Daniel [Sandia National Laboratories (SNL)]↗
DOE OSTI2021

Enhanced defect annihilation capability of the graphene/copper interface: An in situ study

It is well known that energetic heavy ion irradiations can often induce defects and ultimately lead to material degradations. Interfaces, such as high-angle grain boundaries (HAGBs), are generally used as defect sinks for alleviating the irradiation damage. However, HAGBs are often unstable during radiation. Here we investigate the interfacial irradiation responses of the graphene (Gr)/Cu composites by using in situ Kr ++ ions irradiation under transmission electron microscopy. The results revealed that the Gr/Cu interface exhibits higher defect annihilation capability compared to the HAGBs in Cu. Moreover, the atomistic simulations suggested a slightly higher and larger range of stress field for the Gr/Cu interface, which contributes to the enhanced defects absorption capability. The present findings are essential to understand and design a new class of carbon/metal composites with superior irradiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Determination of Thermal Expansion, Defect Formation Energy, and Defect-Induced Strain of α-U Via ab Initio Molecular Dynamics

Uranium (U) is often alloyed with molybdenum (Mo) or zirconium (Zr) in order to stabilize its high-temperature body-centered cubic phase for use in nuclear reactors. However, in all metallic fuel forms, the α phase of U remains in some fraction. This phase decomposition due to temperature or compositional variance can play an outsized role on fuel performance and microstructural evolution. Relatively little is known about fundamental point defect properties in α -U at non-zero temperatures, from either computational or experimental studies. This work performs the first thorough evaluation of the α phase of U via ab initio molecular dynamics (AIMD). A number of thermophysical properties are calculated as a function of temperature, including equilibrium lattice parameters, thermal expansion, and heat capacity. These results indicate a two-region behavior, with the transition at 400 K. The thermal expansion/contraction in the a / b direction occurs rapidly from 100 up to 400 K, after which a more linear and gradual change in the lattice constant takes place. The volumetric expansion matches experiments quantitatively, but the individual lattice constant expansion only matches experiments qualitatively. Point defect formation energies and induced lattice strains are also determined as a function of temperature, providing insight on defect populations and the fundamentals of irradiation growth in α -U. Interstitials induce significantly more strain than vacancies, and the nature of that strain is highly dependent on the individual lattice directions. The direction of point defect-induced lattice strain is contrary to the irradiation growth behavior of α -U. This work shows that isolated point defects cannot be the primary driving force responsible for the significant irradiation-induced growth of α -U observed experimentally.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Simultaneous exposure to neutron radiation and hydrogen environment: Effect on hydrogen retention

Here, this study demonstrates a novel methodology that simultaneously exposes materials to both neutron irradiation and a hydrogen environment, enabling study on in pile interaction between radiation-induced defects and hydrogen in tungsten. The hydrogen environment was established by releasing hydrogen from vanadium hydride within an irradiation capsule. The hydrogen pressure during irradiation was estimated as 14.2 Torr. Post-irradiation thermal desorption spectroscopy analysis showed increased hydrogen retention in irradiated tungsten compared to unirradiated samples. Two dominant desorption peaks at ∼470 °C and ∼700 °C were observed in irradiated and unirradiated tungsten samples, suggesting similar trapping mechanisms between the two samples. However, an additional desorption peak at ∼800 °C in irradiated tungsten from the hydride capsule suggests an interplay between irradiation induced defects and hydrogen exposure, leading to the formation of additional hydrogen trapping sites. These findings provide critical insights into hydrogen retention in tungsten under fusion-relevant conditions and stimulate future studies to investigate the synergistic effects of neutron irradiation and hydrogen exposure.

08 HYDROGEN↗
DOE OSTI2024

Influence of electrical field on the susceptibility of gallium nitride transistors to proton irradiation

Radiation susceptibility of electronic devices is commonly studied as a function of radiation energetics and device physics. Often overlooked is the presence or magnitude of the electrical field, which we hypothesize to play an influential role in low energy radiation. Accordingly, we present a comprehensive study of low-energy proton irradiation on gallium nitride high electron mobility transistors (HEMTs), turning the transistor ON or OFF during irradiation. Commercially available GaN HEMTs were exposed to 300 keV proton irradiation at fluences varying from 3.76 × 10 12 to 3.76 × 10 14 cm 2 , and the electrical performance was evaluated in terms of forward saturation current, transconductance, and threshold voltage. The results demonstrate that the presence of an electrical field makes it more susceptible to proton irradiation. The decrease of 12.4% in forward saturation and 19% in transconductance at the lowest fluence in ON mode suggests that both carrier density and mobility are reduced after irradiation. Additionally, a positive shift in threshold voltage (0.32 V and 0.09 V in ON and OFF mode, respectively) indicates the generation of acceptor-like traps due to proton bombardment. high-resolution transmission electron microscopy and energy dispersive x-ray spectroscopy analysis reveal significant defects introduction and atom intermixing near AlGaN/GaN interfaces and within the GaN layer after the highest irradiation dose employed in this study. According to in-situ Raman spectroscopy, defects caused by irradiation can lead to a rise in self-heating and a considerable increase in (∼750 times) thermoelastic stress in the GaN layer during device operation. The findings indicate device engineering or electrical biasing protocol must be employed to compensate for radiation-induced defects formed during proton irradiation to improve device durability and reliability.

gallium nitride↗
DOE OSTI2023

In situ measurements of non-equilibrium positron state defects during He irradiation in Si

Radiation-induced property changes in materials originate from the energy transfer from an incoming particle to the existing lattice, displacing atoms. The displaced atoms can cause the formation of extended defects including dislocation loops, voids, or precipitates. The non-equilibrium defects created during damage events determine the extent of these larger defects and are a function of dose rate, material, and temperature. However, these defects are transient and can only be probed indirectly. Herein, we present direct experimental measurements and evidence of irradiated non-equilibrium vacancy formation, where in situ positron annihilation spectroscopy was used to prove the generation of non-equilibrium defects in silicon.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Radiation effects in Mn+1AXn phases

M n+1 AX n phases exhibit unique laminated atomic structures that result in properties typical of both ceramics and metals. Due to their unusual characteristics, these materials have been proposed for use in a wide variety of industrial applications, including in nuclear reactors—both fission and fusion systems—where they will be exposed to extreme irradiation conditions and high temperatures. Recently, numerous studies have investigated radiation effects on the M n+1 AX n phases, revealing complex behavior—phase transformations, surface modification, and mechanical property changes—induced by ion or neutron irradiation over a range of temperatures. This review summarizes recent experimental and theoretical work on the response of the M n+1 AX n phases to irradiation and discusses the intrinsic controls on the radiation tolerance of these materials. Based on the review of the present body of work, a comprehensive understanding of the mechanisms of irradiation-induced structural modification and defect evolution in M n+1 AX n phases is developed, as well as proposed strategies for designing novel M n+1 AX n phases with enhanced performance under extreme irradiation conditions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Flash electropolishing for TEM: Reducing FIB‐induced defects in tungsten with protocols for new materials

Focused ion beam (FIB) milling has become the dominant approach for site-specific transmission electron microscopy (TEM) specimen preparation; however, FIB damage remains a critical limitation for reliable microstructural characterisation, particularly in radiation effects studies. Tungsten is especially susceptible to FIB damage due to its high nuclear stopping power, which promotes the formation and strong diffraction contrast of FIB-induced ‘black spot’ defects that are indistinguishable from very fine irradiation-induced loops/defects resulting from low to intermediate temperature neutron irradiation. In this work, flash electropolishing is systematically evaluated as a post-FIB treatment for minimising preparation-induced artefacts for TEM analysis of tungsten-based alloys. Using a range of non-, ion-, and neutron-irradiated tungsten materials, the effectiveness of flash electropolishing has been assessed through direct comparison with conventional FIB and plasma-FIB preparation including low-energy Ga, Ar, Xe ion cleaning. The results demonstrate that flash electropolishing effectively removes FIB-damaged layers and ‘black spot’ defects, thereby enabling reliable observation of irradiation-induced dislocation structures. Key processing parameters governing flash electropolishing quality – including lamella thickness, applied voltage, polishing duration, electrolyte chemistry, and cathode geometry – have been systematically evaluated, and clear criteria were established for determining when flash electropolishing is required to ensure reliable microstructural analysis. This work also provides practical guidance for implementing flash electropolishing as an artefact-controlled specimen-preparation approach for TEM characterisation of FIB-produced specimens. The systematic protocol can be extended to other, non-tungsten materials.

TEM sample preparation↗
DOE OSTI2020

Irradiation-enhanced torsional buckling capacity of carbon nanotube bundles

Molecular dynamics simulations are used to understand the torsional buckling of pristine and irradiated carbon nanotube (CNT) bundles. Irradiation-induced inter-tube defects are shown to significantly increase the critical buckling torque and critical buckling angle, while slightly increasing the torsional stiffness. In contrast, intra-tube defects are found to degrade the torsional properties. Such competing interactions cause irradiation enhancement to occur in large bundles where significant inter-tube bonding can occur. However, the irradiation enhancement effect becomes weak for very large bundles in which enhanced inter-tube interactions already exist in unirradiated bundles. In pristine CNT bundles of all sizes under torsional loading, CNTs can slip via the weakly interacting van der Waals force, whereas in the irradiated bundles, the inter-tube defects prevent slipping. The study further shows that the formation of one-dimensional carbon chain defects contributes to enhanced friction under slipping.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2025

Defect formation and microstructure tuning via proton irradiation to control electrochemical and phase reversibility in layered battery materials

The reversibility of phase transformation influences the functionality of electrode materials in batteries. In many battery materials, nanosized grains favor phase reversibility but at the cost of cyclability due to aggravated side reactions with the electrolyte. Here, in this study, we present a novel approach to enhance the phase transformation reversibility of layered oxide cathodes, exemplified by Na 2/3 Fe 1/2 Mn 1/2 O 2 through proton irradiation. In addition to forming defects, proton irradiation at sufficiently high doses can subdivide single grains into multiple nanodomains without physically rupturing them. Hence, the single grains of the material assume a pseudo-secondary particle nature without reducing the overall grain size. Preserving the grain size is advantageous, as it reduces side reactions, which is not possible with conventional grain size reduction methods. While chemical transformations and defect formation induced through proton irradiation can influence the stability of battery materials, it is expected that structural reorganization due to cycling-induced phase transformation will be contained within these nanodomains. Such confinement of phase transformation is potentially responsible for enhancing the reversibility of layered oxide materials in our study. Thus, our study suggests that grain subdivision could become an effective microstructure tuning strategy for managing electrochemical cycling-induced phase changes in battery electrodes.

25 ENERGY STORAGE↗
DOE OSTI2025

Production of Precursor Materials for the Development of Iron-Based Superconductor (IBS) Wire

The goal of this agreement will be devoted to the understanding of the radiation-induced defects landscape produced on irradiated bulk samples and of the intrinsic properties of the stoichometric phases with special regards to the difference interaction with multiple chemical additions and doping elements. Samples will be produced by ENEA Frascati per Fermilab specifications and then shipped to Fermilab for the characterization testing and the execution of irradiation experiments.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2025

Probability Density Function for the spatial and intensity distribution of neutron-induced defects in Silicon

The ability to model semiconductor device degradation under neutron irradiation depends upon having a robust modeling capability for the neutron-induced collision cascades as well as a means to analytically fit the resulting probability distributions of defect production and ionizing energy deposition for purposes of extrapolation to low-probability, high-consequence scenarios. In this paper, the widely-utilized binary collision approximation codes MARLOWE and SRIM are deployed in conjunction with a critical examination of their parameterizations as benchmarked against higher-fidelity molecular dynamics simulations. A simple 3-parameter form described by the Generalized Logistic Distribution is shown to be a good fit to Frenkel pair and ionization intensity distributions in bulk silicon. The BCA codes are then applied to simulate cascades in 5 nm layers of a representative gate-all-around nanosheet transistor, where joint probability distributions of threshold levels of damage to multiple layers are evaluated.

36 MATERIALS SCIENCE↗