Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe+ and 1 MeV Kr2+ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO2 or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO2 and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.

# Introduction

Uranium carbide (UC) has attracted interest in advanced reactor systems such as gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs) due to its high uranium density and exceptional thermomechanical properties [1]. Historically, UC was studied during early nuclear rocket propulsion programs, including NASA’s Nuclear Engine for Rocket Vehicle Application (NERVA), which leveraged its high thermal conductivity and uranium density to meet rigorous performance criteria in space [2]. Subsequent studies explored the compatibility of UC with liquid metal coolants, prompting investigations into sodium-cooled fast reactors and molten salt systems [3]. Despite these promising attributes, uranium dioxide (UO2) became the commercial standard fuel for light water reactors due to its mature fabrication processes, chemical stability, and well-characterized in-reactor performance. The resurgence of Generation IV reactor initiatives has revitalized research into alternative carbide fuels, particularly UC, to enable higher discharge burnups and improved thermal behavior [4].

The principal advantages of UC relative to UO2 include a uranium atom density of 12.9 g.cm-3 versus 9.7 g.cm-3 for UO2, and thermal conductivity of approximately 22 W·m⁻¹·K⁻¹ at 1200 K compared to about 3 W·m⁻¹·K⁻¹ for UO2 [5–8]. These properties reduce temperature gradients under high power densities, mitigate centerline melting risk, and allow more compact core designs. In addition, UC exhibits chemical compatibility with refractory carbide and nitride materials—such as silicon carbide (SiC) and zirconium carbide (ZrC)—under consideration for structural and cladding components in advanced concepts [9]. While mixed uranium carbide–oxide (UCO) compositions are used for TRISO kernel fuels, single-phase UC remains less explored due to fabrication challenges and concerns over oxidation stability [10]. Additionally, UC pellets enclosed in a SiC composite cladding are proposed for the Energy Multiplier Module (EM2), an advanced reactor being developed by General Atomics [11]

Radiation damage in nuclear fuels manifests as point defect creation, gas bubble nucleation, swelling, and microstructural evolution, critically impacting thermal conductivity, mechanical integrity, and fission gas retention. A quantitative understanding of defect dynamics under reactor-relevant irradiation conditions is essential for assessing fuel performance. In-situ transmission electron microscopy (TEM) under ion-beam irradiation permits real-time observation of bubble formation, dislocation loop evolution, and other microstructural transformations. Although in-situ TEM studies have provided valuable insights into UO2 and surrogate materials like CeO2 and ThO2 [12–14], analogous data for UC are sparse. This study addresses this gap by systematically investigating radiation-induced microstructural changes in UC under Xe and Kr ion irradiation at temperatures up to 900 °C.

The objectives of the present work are to quantify Xe bubble nucleation, growth, and density as functions of irradiation temperature and dose to assess swelling behavior in UC. Additionally, we aim to characterize dislocation loop formation and evolution under Kr irradiation to elucidate damage mechanisms and their temperature dependence. Furthermore, we will compare the in-situ microstructural response of UC with UO2, UN, and other carbide fuels. Lastly, this work will discuss the unexpected emergence of nanograin-like structures in thin regions under irradiation.

# Experimental methods

## Sample synthesis and preparation

Single-phase UC pellets were synthesized at Idaho National Laboratory (INL) via an arc-melting and powder metallurgy technique similar to the method discussed by Wagner et al. and Harp et al [15,16]. A stoichiometric amount of U (50 at%) and graphite (50 at%) was melted using an MRF Tri-Arc melting system. After completion of arc-melting, the ingot was comminuted in an inert atmosphere glovebox (<5 ppm O2) via a hammer mill and planetary ball mill. A sieve was used to determine the approximate particle size of less than 37 µm. UC powder was sieved after ball milling; all material passed through a 400 mesh (37 µm) sieve. Pellets were formed by uniaxially pressing at 550 MPa utilizing zinc stearate as a die lubricant. Pellets were then sintered at 1800 °C and held for 8 h in flowing argon. Sintered pellet density was 95 % TD (13.61 g·cm-3) [17] as determined by the Archimedes technique using ethanol as the immersion fluid. X-ray diffraction (XRD) confirmed single-phase UC with no observable secondary phases, as shown in Fig. 1
                        a. Rietveld refinement was used to determine the lattice constant of UC which confirmed a lattice constant of 4.958 Å [18]. Optical microscopy (Fig. 1b) of polished cross-sections revealed equiaxed grains ranging from 20 μm to 120 μm.

TEM lamellas were prepared and mounted on a Mo grid using the FIB technique at the Irradiated Materials Characterization Laboratory (IMCL), INL through the FEI Quanta 3D FIB. The electron energy loss spectroscopy (EELS) technique was applied to measure the thickness of the samples [19].

## In-situ ion irradiation

In-situ irradiation was performed at Argonne National Laboratory using an Intermediate Voltage Electron Microscope (IVEM), Hitachi-9000 with a double-tilt heating stage. Table 1
                         shows the irradiation conditions and final irradiation dose for the samples. Samples were irradiated with 300 keV Xe⁺ ions or 1 MeV Kr²⁺ ions at fluxes of 6.25 × 1011 and 1.25 × 1012 ions·cm⁻²·s⁻¹, respectively. Xe⁺ ion irradiations were maintained (±5 °C) and monitored by an integrated thermocouple at temperatures of 300 °C, 600 °C, and 900 °C, while Kr²⁺ were maintained at temperatures of 300 °C and 600 °C due to the practical considerations of sample stability, integrity and the timeframe limitations on the experimental setup within the study’s scope.

SRIM-2013 simulations in quick Kinchin–Pease mode (U displacement energy: 40 eV; C displacement energy: 25 eV) estimated ion ranges and damage profiles, as shown in Fig. 2
                         [20–22]. Full-cascade simulations validated dpa estimates, indicating peak damage at ∼50 nm for the 300 keV Xe and ∼120 nm for the 1 MeV Kr.

## Image acquisition and analysis

Bright-field TEM images were acquired under two-beam conditions using a Gatan UltraScan CCD camera at selected fluences. Defect quantification was performed in ImageJ on five separate thin regions per sample [23]. Xe bubble diameters were measured as equivalent circular diameters; number densities were converted to volumetric densities using lamella thickness. Conversely, strain contrast was used to identify dislocation loops, while loop diameters and densities were measured manually using ImageJ as well. Uncertainties represent standard deviations propagated through swelling and dislocation loop calculations, assuming spherical bubble geometry. Post-irradiation selected lamellae underwent HAADF imaging and EDS mapping in a Thermo Scientific Titan TEM (80–200 kV) with ChemiSTEM™. These analyses confirmed bubble composition (Xe) and assessed elemental segregation at defect sites.

# Results

## Xe bubble formation and swelling

At a temperature of 300 °C and an ion fluence of 2.5 × 10¹⁵ ions.cm-2 (corresponding to 9.3 displacements per atom or dpa), the detection of Xe bubbles through TEM was negligible, with sizes less than 1 nm. In contrast, at 600 °C (with the same ion fluence of 2.5 × 10¹⁵ ions.cm-2), the bubbles averaged 1.1 ± 0.3 nm in diameter, exhibiting a volumetric density of 8.3 × 10²² m-3, and resulting in a swelling rate of 0.36 %. When the temperature increased to 900 °C (with an ion fluence of 1.88 × 10¹⁵ ions.cm-2), the average bubble size increased to 1.3 ± 0.4 nm, and the volumetric density was recorded at 7.1 × 10²² m-3, which correlated to a swelling of 0.60 %, as indicated in Table 2
                        . All uncertainties represent the standard deviation on the total number of bubbles counted and measured in each image. The bubbles maintained a spherical shape and exhibited narrow distributions, suggesting limited migration and coalescence phenomena. Fig. 3
                         illustrates the bubbles captured at over and under-focus conditions by 1 micron corresponding to these irradiation scenarios.

## Dislocation loop evolution under Kr irradiation

Fig. 4
                         illustrates the TEM images of the UC samples subjected to irradiation with Kr2+ at 300 °C and 600 °C. At 300 °C, under Kr irradiation, small loops around 4 to 5 nm were observed at the initiation of irradiation; the density increased to approximately 2 × 1021 m-3 at 1 dpa and subsequently plateaued while the mean diameter expanded to approximately 10 ± 1 nm after 2 dpa. Conversely, at 600 °C, the initial density was approximately 1 × 1021 m-3 at the onset of irradiation, with loop sizes around 5 ± 0.8 nm; the density reached its peak at 1 dpa (1.4 × 1021 m-3) before declining as the loops coarsened, reaching diameters of 28 ± 2 nm by 3 dpa. Figs. 5 and 6
                        
                         present the size-distribution analyses, which indicate the presence of loops smaller than 16 nm at 300 °C and a bimodal distribution at 600 °C, suggesting concurrent nucleation and growth phenomena. Unirradiated samples showed some loops forming at high temperatures due to the annealing effect of FIB-induced damage, consistent with previous observations in ThO2 [12].

## Nanograin formation in thin regions

Regions of lamellae with a thickness of less than 50 nm indicated the presence of nanocrystalline grains ranging from 5 to 20 nm following Xe irradiation conducted at 600 °C, as illustrated in Fig. 7
                        . Furthermore, Fig. 8
                         displays selected area electron diffraction (SAED) patterns that reveal rings corresponding to UC alongside minor quantities of UO2; EDS confirmed a mixed UC/oxide composition, as depicted in Figs. 9 and 10
                        
                        , while peaks for Mo and Cu exist as a byproduct of the TEM grids and X-rays from XRD, respectively. The high concentrations of loops and bubbles in these areas suggest a buildup of defects caused by confined geometry, similar to the polygonization effects seen in irradiated ZrC and high-dose UO2 [24,25].

# Discussion

## Xenon bubble formation and swelling behavior

This study elucidates the dynamic formation of Xe bubbles in UC. Notably, while Kr irradiation demonstrated the evolution of dislocation loops concerning irradiation temperature and dose in UC, the impact of irradiation conditions and temperatures influenced the evolution of dislocation loops more significantly than that of Xe bubbles, at least within the specified temperature range. This section will discuss the mechanisms underpinning the operation of dislocation loops and Xe bubbles and compare them with existing literature.

Fig. 11
                         compares with other swelling studies conducted on carbide and nitride fuels [26–28]. In these investigations, Ronchi et al. and Blank et al., employed a combination of early TEM imaging for nano-sized bubbles and optical imaging for larger bubbles in neutron-irradiated samples up to 6 % FIMA. While their measured values are marginally higher than those reported in this study, they gradually increased until reaching a specific temperature. Beyond this threshold, the rate of swelling escalated sharply. In Fig. 11, a line connecting the points where a swelling increase was observed is denoted as the critical temperature, ranging from 900 °C to approximately 1150 °C. It was indicated that these temperatures coincided with the points at which extensive grain growth commenced. It has been documented that carbide fuels exhibit lower bubble concentrations at similar or identical temperatures due to the elevated self-diffusivities associated with carbon [29].

The persistence of Xe bubbles less than 2 nm in diameter up to 900 °C reflects low Xe diffusivity and vacancy mobility in UC, which limits the bubble coalescence. Swelling values (0.36 % at 600 °C, 0.60 % at 900 °C) are significantly lower than UO2 (>1.0 % at similar dpa), consistent with stronger covalent bonding, higher melting points, and lower gas diffusivities in carbide matrices [30,31]. The marginal swelling increase with temperature suggests that UC’s critical swelling onset exceeds 1000 °C, paralleling MX-type (M is U and/or Pu, X is C and/or N) fuel trends.

## Dislocation loop evolution and temperature dependence

Kr ions can displace numerous lattice atoms in cascades; however, not all displacements remain as defects [32]. The direct nucleation of defect clusters and dislocation loops has been theoretically investigated in UO2 [33]. In this context, it has been proposed that particularly large vacancy clusters form directly at the center of these cascades, with loop punching contributing to dislocation loops at their periphery. The observation of dislocation loops in UO2 irradiated at room temperature may support this theory [34].

It is essential to note that the formation and evolution of dislocation loops are contingent on the mobility, survivability, and production of point defects. Various parameters, including irradiation temperature and dose, ion type, sample thickness, and density, influence these factors [35–37].

As indicated in Table 1, the irradiation conditions for samples exposed to Kr ions were uniformly maintained, except for varying irradiation temperatures. The primary focus of this analysis is on the effects of temperature on dislocation loops. Consequently, one can infer that the increased growth rate of dislocation loops observed at 600 °C is attributable to the elevated temperature, which enhances the mobility of defects (vacancies and interstitials). Increased dislocation loop growth with rising temperatures has also been documented in UO2 and its surrogate, CeO2 [13,38].

## Comparative analysis of other nuclear fuels

Fig. 12
                         presents a comparative analysis of dislocation loop sizes among UC, UO2, and UN [38,39]. UC exhibited the largest dislocation loops at higher doses (ranging from 5 × 1014 to 1 × 1015 ions.cm-2), even under lower irradiation temperatures. UO2 demonstrated larger dislocation loop sizes than UN at elevated temperatures and equivalent doses; however, intriguingly, UO2 manifested smaller dislocation loop sizes than UN at reduced irradiation temperatures, as illustrated in Fig. 12.

Kosmidou et al. highlighted the impact of sink and source strengths during irradiation [39]. Uranium defects limit interstitial growth in both UN and UO2. The anti-Schottky defect, consisting of one uranium and two oxygen interstitials, is predominant due to high mobility. Dislocation growth is driven by net interstitial production, expedited in UO2. The effect of irradiation temperature on UO2 is more pronounced than on UN, correlating with their melting temperature disparity; UO2 has a lower melting temperature than UN.

In another study on the phase evolution in an UN-UO2 system, He et al., utilizing the kinetic rate theory to analyze dislocation loop density and size, suggested that loop nucleation and growth are governed by the mobility of the uranium interstitials which were measured to be one order of magnitude higher in UN than in UO2 [40]. Taking that into account, a similar comparison can be made here, suggesting that the mobility of uranium interstitials seems to be higher in UC as well. Multiple studies have been conducted in the past century on the diffusivities of uranium and carbon in UC [33–36] with various diffusion parameters calculated and measured at different temperature ranges that vary from 1500 to 2500 °C [41–44].

UC is anticipated to exhibit behavior comparable to that of UN under these conditions. However, the observed data indicate that UC demonstrates the most significant dislocation growth. A substantial body of literature highlights the critical role of homologous temperature in influencing the overall microstructural evolution and, more specifically, promoting dislocation loop growth, especially in FCC materials [45–48]. Among the materials analyzed, UN has the highest melting temperature, followed by UO2, while UC has the lowest. At reduced temperatures, the lower diffusivity of interstitials in UO2 considerably impedes the growth of dislocation loops.

## Comparative analysis of similar carbides

The crystal structure of UC shares similarities with several other carbides that have been studied for nuclear applications, including ZrC. Fig. 13
                         provides a comparative analysis of UC and ZrC, emphasizing the behavior of dislocation loops under irradiation across various temperatures [49]. Notably, ZrC demonstrates significantly smaller sizes of dislocation loops. Research has shown that ZrC possesses high energy barriers for both zirconium and carbon migration, which impacts its microstructural properties [50]. Additionally, Ulmer et al. have clarified that the observed small dislocation loop sizes stem from the elevated concentration of carbon vacancies present in sub-stoichiometric ZrC1-x [51] Given these considerations, carbides, as documented in the literature and our experimental findings, exhibit an increased propensity for dislocation growth.

## Nanocrystalline structure formation in thin regions

The unprecedented nanograin-like structure observed in UC warrants further investigation into its underlying mechanisms. Notably, similar nano-grained microstructures were previously documented by S. Agarwal et al. in helium-irradiated ZrC, which shares the crystal structure of UC [24]. Their findings showed that these microstructures formed near grain boundaries due to high-temperature irradiation-induced deformation rather than through simple recrystallization.

Furthermore, Matzke et al. have reported instances of grain subdivision in UO2, attributing it to the accumulation of radiation-induced edge dislocations, which may be precursors to high burn-up structures (HBS) [25,52,53]. Their studies employed 300 keV and 500 keV Xe ion beams, methodologies closely aligned with our current investigation. More recently, Miao et al. [46], and Cappia et al. [47] examined this phenomenon at elevated energies, reaching up to 84 MeV, and uncovered the initial stages of HBS formation alongside developing a nanocrystalline structure in ion-irradiated UO2 [54,55]. Their results further support Matzke et al.’s hypothesis that nanocrystalline structures are predominantly formed through polygonization, propelled by radiation-induced dislocations rather than recrystallization. The observation of analogous phenomena in UC is encouraging, prompting plans for additional experiments focused on grain subdivision in single-crystal or micro-grained UC to deepen our comprehension of this intricate process.

# Conclusion

In-situ TEM irradiation of UC at temperatures ranging from 300 °C to 900 °C, conducted under Xe and Kr ion exposure, illustrate its gas retention capabilities and distinctive dislocation dynamics. The observed minimal swelling (<0.6 %) alongside bubble sizes less than 2 nm accentuates UC's fission-gas containment properties at these temperatures. A temperature-dependent loop evolution is noted, exhibiting small, dense loops at 300 °C in contrast to larger, coarsened loops at 600 °C, thereby underscoring the effects of homologous temperature on material behavior. The comparative analysis positions UC favorably when contrasted with UO2, UN, and ZrC for applications necessitating high-temperature and fast-spectrum performance. Additionally, the unanticipated formation of nanograins within thin lamellae indicates phenomena of geometry-driven recrystallization, warranting further investigation. Continued research at elevated temperatures, increased doses, and extended timeframes is imperative to qualify UC for comprehensive deployment in advanced reactor technologies.

# CRediT authorship contribution statement

Rashed Almasri: Writing – original draft, Methodology, Investigation, Formal analysis, Data curation. Adrian R. Wagner: Writing – review & editing, Resources, Investigation, Funding acquisition. Laura Hawkins: Writing – review & editing, Resources, Investigation. Wei-Ying Chen: Writing – review & editing, Resources, Methodology, Investigation. Jennifer K. Watkins: Writing – review & editing, Investigation, Formal analysis. Jian Gan: Writing – review & editing, Resources, Investigation. Lingfeng He: Writing – original draft, Supervision, Project administration, Investigation, Funding acquisition, Formal analysis, Conceptualization.

This work is supported by the faculty startup fund provided by North Carolina State University. The sample preparation and ion irradiation are supported by the U.S. Department of Energy, Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC0705ID14517 as part of Nuclear Science User Facilities. The authors would like to acknowledge the staff of the Electron Microscopy Laboratory (EML) and Irradiated Materials Characterization Laboratory (IMCL) at the Materials and Fuels Complex at INL for their effort in handling, preparing, and transferring of the specimens used in this work. The authors would like to acknowledge Mr. Dzmitry V. Harbaruk for his support of the ion accelerator during the IVEM experiments.