Fuel-cladding Chemical Interaction Research Articles

First-principles investigation of lanthanides diffusion in HCP zirconium via vacancy-mediated transport

The diffusion of lanthanide fission products plays an important role in the growth of the fuel-cladding chemical interaction (FCCI) region in metallic fuels. The use of a Zr interdiffusion barrier may mitigate the transport of lanthanides from the fuel to the cladding, but the efficacy of such a liner is not yet known. In this paper, the stability and vacancy-mediated diffusion of La, Ce, Pr, and Nd in hexagonal close-packed (HCP) Zr is investigated via density functional theory (DFT) calculations and self-consistent mean field (SCMF) analysis. DFT is used to calculate the formation, binding, and migration energies of vacancies and vacancy-solute pairs. The DFT energetics are used in the KineCluE code to calculate the Onsager transport coefficients. La is found to be the fastest diffusing species in HCP Zr and experiences an almost isotropic diffusion behavior. The other three species (Ce, Pr, and Nd) demonstrate anisotropic diffusion where the diffusion in the basal planes is significantly faster than that along the c-axis. The calculated lanthanide diffusivities in HCP Zr are fitted to an Arrhenius relation and the activation energies and prefactors are reported for the first time. Furthermore, the vacancy drag and the segregation tendencies were analyzed using the calculated off-diagonal transport coefficients. According to our vacancy-mediated diffusion model, lanthanides are expected to be enriched at vacancy sinks at low temperatures, while at high temperatures, lanthanides are depleted at sinks and will preferably diffuse into the bulk. The enrichment/depletion transition temperature depends on the diffusion direction (basal or axial) and hence will be controlled by the grain texture and orientation.

Comparative study of Cr-coating and Zr-lining method on inner surface of HT9 cladding to mitigate Fuel-Cladding chemical interaction

A comparative study was conducted between electroplated Cr coatings and Zr linings as inner-surface barriers to mitigate chemical interactions in metallic nuclear fuel claddings. The comparison involved microstructural analysis and micro-CT examination to analyze their inner-surface morphology, along with an evaluation of their mechanical properties. Furthermore, diffusion couple tests were conducted to determine the fuel-cladding chemical interaction (FCCI) resistance. Both barrier processes demonstrated superior FCCI resistance compared with untreated HT9; however, several limitations were identified. Although potential improvements were expected for the Cr-coating process through modifications of the applied pulsed current waveforms, the Zr-lining process presented major process-related dilemmas that were challenging to overcome.

Evaluation of Sb-Nd and Te-Nd phases within the U-Zr fuel matrix and their interactions with HT9 alloy

Antimony (Sb) and tellurium (Te) were investigated as potential additives for U-10Zr (wt.%) metallic fuel to limit the fuel-cladding chemical interaction (FCCI) with HT-9 alloy. Neodymium (Nd) was utilized to simulate the formation of lanthanide-based solid fission products which are known to play a detrimental role in FCCI. Fuel alloys of U-Zr-Sb-Nd and U-Zr-Te-Nd were evaluated in their annealed condition and compared against their as-cast conditions. Isothermal diffusion couple experiments were performed between U-Zr-Nd, U-Zr-Sb-Nd, and U-Zr-Te-Nd against the cladding alloy HT9 to evaluate the effectiveness of the additives to stabilize Nd within the fuel alloys, as well as investigate the interaction regions that form between the different fuel alloys and HT9. SbNd and Sb3Nd4, and TeNd are found to be the primary neodymium-based phases formed in the U-Zr-Sb-Nd and U-Zr-Te-Nd alloys, respectively. The zirconium-based phase, Zr2Sb, is also found to form within the former alloy. All phases were found to remain stable through the diffusion experiments and exhibited no interaction with HT9 constituent elements. Preferential interaction between Nd with additivities Te and Sb compared to constituting elements in HT9 was further verified based on density functional theory (DFT) calculated enthalpy of mixing.

A fine pore-preserved deep neural network for porosity analytics of a high burnup U-10Zr metallic fuel

U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model’s performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design.

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

Microstructural characterization of U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln

Antimony is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interactions (FCCI). The most detrimental elements involved in FCCI are fission product lanthanides, leading to brittle intermetallics and low melting eutectic phases. Previous investigations of Sb as an additive focused on U-10Zr, in wt. %, as the fuel. The current investigation expands that to include Pu in the fuel. Two alloys, U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln (wt. %, Ln=53Nd-25Ce-16Pr-6La) have been investigated using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) to characterize the fuel as-cast microstructure and the microstructure after introduction of lanthanides. Sb reacts with Zr initially, forming Zr2Sb and Zr5Sb3, with as much as 20 at. % interstitial Pu present. In the presence of lanthanides, Sb forms Ln4Sb3 with the lanthanides, containing ∼14 at. % Pu. The Pu is substitutional for the lanthanides in the crystal lattice.

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

Effect of additive (Zr) and dopants (Pd & Sb) on the depth of migration of lighter lanthanides (Nd & Ce) from U-bearing metal alloys into T91 steel

The metal alloy fuel, which is considered as the advanced nuclear fuel, has its own limitations; namely, lower liquidus temperature, higher swelling, and is highly prone to breach the cladding due to fuel-cladding chemical interaction (FCCI). The mobile and insoluble fission products (FPs), especially lanthanide FPs are mainly responsible for FCCI. In the present study, the influence of additive (Zr) and dopants (Pd and Sb) on the depth of migration of two lanthanide FPs, namely Nd and Ce, from different uranium-bearing metal alloys into T91 (ferritic-martensitic steel) cladding material is investigated. Seven metal alloys, namely, U-Nd, U-Nd-Pd, U-Ce, U-Ce-Pd, U-Zr-Nd, U-Zr-Nd-Pd, and U-Zr-Nd-Sb, were prepared using arc melting technique. The amount of lanthanide FPs taken in the above alloys corresponds to about 10 at.% burn-up of a typical fast breeder reactor (FBR) fuel. The pulsed laser deposition (PLD) method was used to obtain alloys in the form of thin film on T91 cladding, with simultaneous in-situ annealing. Subsequently, these films were subjected to elemental depth profiling using secondary ion mass spectrometry (SIMS). The depth of migration for Nd and Ce from the seven metal alloy systems into T91 cladding was determined based on SIMS data. The extent of the individual and simultaneous effect of additive and dopant on the depth of migration was determined. Both dopants (Pd and Sb) are effective individually in binding Nd and Ce within the fuel matrix (in the U matrix) by forming stable inter-metallic compounds. The effect of Zr on the migration of lanthanide FPs is negligible in the presence of dopants. Stable inter-metallic phases responsible for binding lanthanides FPs are identified. A comparative study on the efficiency of two dopants (Pd and Sb) is also discussed. For the first time, the depth of migration of Nd and Ce from U metal alloys into T91 cladding was measured in the presence of dopants (Sb/Pd) by using a novel technique of PLD followed by SIMS. These studies are very useful for understanding FCCI.

Zirconium nanoparticle coating development for FCCI diffusion barrier in nuclear cladding

Fuel clad chemical interaction (FCCI) is a pervasive issue for metallic nuclear fuels. FCCI can damage cladding, reduce thermal efficiency, and compromise the structural integrity of the fuel rod. In this study, we evaluated the FCCI mitigation capabilities of electrophoretically-deposited Zr nanoparticle coatings on ferritic-martensitic stainless-steel cladding (HT9). Zr was deposited on the internal surface of 6 mm outer diameter cladding using electrophoretic deposition (EPD). Thickness varied among samples with similar EPD conditions. A uniformly coated sample was selected and sent to the Argonne Tandem Linac Accelerator System (ATLAS) to simulate the coating's behavior under irradiation conditions. The coating prevented FCCI in diffusion couple studies. Results of this study confirm that Zr could potentially make an effective FCCI barrier in metallic fuel rods but the deposition process needs improvement.

Significant difference of lanthanide fission products diffusion in Cr and α-Fe: An atomic-level study

By using first-principles calculations, diffusion of lanthanides (Ln) in bcc Cr and bcc Fe was studied to identify the microscopic mechanisms occurring in barrier materials that mitigate fuel cladding chemical interaction (FCCI) between the U-Pu-Zr and ferritic/martensitic steel systems. The properties of Ln-vacancy complexes were investigated by an oversized solute atoms (OSA) model. Results showed that strong lanthanides-vacancy binding, steep migration barriers and low vacancy concentration could lead to poor vacancy mobility in the vicinity of Ln in Cr. The Ln diffusion coefficients can be 4–8 orders of magnitude lower in Cr than those in Fe, confirming remarkable mitigating effects of Cr on Ln diffusion. Charge transfer and localization in the Cr matrix are analyzed to be responsible for the steep migration barriers. Owning to their high valence bond energies, V, Mo, and W are argued to have potentially similar or better performance in mitigating the diffusion of Ln.

Direct and pulse electroplating effects on the diffusion barrier property of plasma-nitrided Cr coatings on HT9 steel

The introduction of a Cr and chromium nitride multi-coating barrier between the nuclear metallic fuel and the cladding is a potential candidate to mitigate fuel cladding chemical interaction (FCCI). Cr coatings were electroplated on HT9 disks using direct and pulse currents followed by plasma nitriding for synthesizing additional nitride coating. The pulse current induced Cr coatings revealed severe microstructural changes from crack-free to the abundance of cracks during the plasma nitriding due to phase transformation during the plasma nitriding. The phase transformation of hexagonal CrH to cubic Cr occurred at a temperature as low as 110 °C resulting in the liberation of hydrogen pores agglomerated at Cr grain boundaries and internal stress. Finally, the diffusion barrier property of the pulse current induced Cr coatings was significantly reduced, while the plasma-nitrided Cr coating produced by a direct current exhibited a crack-closure effect, and an additional nitride layer successfully hindered the elemental diffusion of Ce and Nd at 650 °C for 25 h. This current study points to the critical effect of current waveforms on the final barrier property of Cr-chromium nitride coatings for nuclear cladding applications.

Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel

Fuel cladding chemical interaction (FCCI) is a key phenomenon needs to be better understood to establish the design basis for U-10Zr metallic fuel performance. Characterizing the microstructure and chemical composition of FCCI at micron and sub-micron scale is critically important toward a more mechanistic understanding of FCCI phenomenon and its potential effects on cladding integrity and metallic fuel performance. This paper, by using transmission electron microscopy, investigated the FCCI region in HT9 cladded U-10Zr fuel irradiated to 5.7% FIMA burnup at a peak inner cladding temperature of 615 °C in Fast Flux Test Facility (FFTF). Four distinct layers are identified in the FCCI region. The migration of Fe into the fuel side leads to the formation of several U-Zr-Fe ternary phases, including χ-Fe0.5Zr0.32U0.18, ε-Fe0.3Zr0.4U0.3, and λ-Fe0.06Zr0.23U0.71, mingled with UFe2, U6Fe, and U phase at various Fe penetration depth up to ∼ 150 µm. On the cladding side, grain coarsening and significant lanthanides infiltration along grain boundaries are observed. Laves phase, (Fe,Cr)2(Mo,W), which typically does not exist in fresh HT9, is identified in a wide radial range in the cladding. The typical HT9 martensitic lath structure and pre-existing M23C6 precipitates disappear, partially or completely, depending on the radial distance from the fuel-cladding interface. Those microstructural and compositional changes could cause mechanical degradation in the HT9 cladding. The present characterization results will improve the understanding of FCCI phenomenon and facilitate the development of microstructure-informed FCCI modeling for metallic fuel.

Advanced Characterization of Fuel Cladding Chemical Interaction between U-10Zr Fuel and HT9 Cladding Tested in Fast Flux Test Facility

Advanced Characterization of Fuel Cladding Chemical Interaction between U-10Zr Fuel and HT9 Cladding Tested in Fast Flux Test Facility

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs, such as the Versatile Test Reactor, researchers need fuel performance calculations to determine the safety criteria of proposed designs. In order to validate the fuel performance results obtained by the fuel performance code BISON for new reactor designs, legacy fuel from EBR-II and Fast Flux Test Facility MFF with post-irradiation examination (PIE) data need to be used as validation cases to benchmark models. In this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for a fuel-cladding chemical interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using a standard error of the estimate to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Possible improvements to current models to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups that is observed for the axial fuel swelling height are discussed. Although two pins in EBR-II, X447/X447A (DP70 and DP75), were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interaction (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the high-Cr tempered martensitic HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature (PICT) reached to 650 ℃. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe2Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ℃).

Chemical interactions between neodymium and advanced stainless steels

During the burnup of U–Zr based metal fuels, lanthanide fission products migrate to fuel periphery and deposit on cladding inner surface, resulting in fuel-cladding chemical interaction (FCCI). Of the lanthanide fission products contained in irradiated U-Zr based fuels, neodymium (Nd) has the highest concentration. In this study, we performed out-of-pile diffusion couple testing to investigate the chemical interactions between neodymium (Nd) and four types of stainless steels (SSs), aiming at elucidating the effects of composition and microstructure on the chemical interactions. The four types of steels are Fe-Cr based oxide dispersion-strengthened (ODS) steel (ODS-1), Fe-Cr-Ni based ODS steel (ODS-2), additively manufactured (AM) 316L SS (A316), and wrought 316L SS (W316). The results show that the interdiffusion between the major constituents in ODS-1 (Fe and Cr) and Nd is suppressed, while nickel was the most active element during diffusion between ODS-2/A316/W316 and Nd, leaving a relatively large depletion area in ODS-2/A316/W316. Compared to W316, the 200-nm pores in the A316 assisted the interdiffusion between the steel constituents and Nd, while the large grain size resulted in slower diffusivity. The kinetics of chemical interaction between Nd and SSs is strongly dependent on the microstructure, including pore structure and grain size.

Post-irradiation characterization of a high burnup mixed oxide fuel rod with minor actinides

The Advanced Fuels Campaign performed a series of irradiation tests of minor actinide-bearing mixed oxide fuel (MA-MOX), the so-called AFC-2C&D experiments, to investigate the transmutation of long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology at burnups exceeding 10 % fission of initial metallic atoms. This manuscript reports the test results derived from one of the five MA-MOX rodlets taken to higher burnup in the AFC-2D irradiation. This includes both non-destructive investigations, such as gamma and neutron spectrometry, and destructive investigations, such as fission gas release, ceramography, and chemical burnup analysis. In addition, the microstructure of the fuel was investigated using advanced electron microscopy techniques including electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). It was observed with EBSD that the pellet had subdivision of the grains and the TEM observed migration of cladding material into the 5 metal precipitates in the fuel which could have been from the higher than desired oxygen/metal ratio. The TEM also showed an enrichment of Cr in fuel clad chemical interaction (FCCI) layer.

Fuel Performance Design Basis for the Versatile Test Reactor

The Fuel Performance Design Basis for the Versatile Test Reactor begins with requirements to maintain safe and efficient reactor operation. For the metal-fueled Versatile Test Reactor, this means a limited number of fuel rod breaches, no fuel melting under steady-state operation and anticipated transients, and continuity of the fuel rod and assembly configuration to avoid impacts to operations of safety systems, maintain expected coolant flow, and allow for efficient fuel handling. Using a large database gathered from previous testing, data were examined to identify and establish preliminary limits on fuel operating conditions. Fuel performance aspects important to fuel operating limits have been identified, including cladding creep, which is addressed with a cladding deformation limit to ensure a limited cladding breach. In addition, fuel-cladding chemical interaction is addressed through limits on cladding temperature and time-at-temperature for steady-state operation, transients, and accidents to mitigate effects leading to cladding breach or fuel melting. Through the implementation of these limits, cladding breach, fuel melting, and deleterious fuel rod and assembly dimensional changes will be prevented.

Investigation of constituent redistribution in U-Pu-Zr fuels and its dependence on varying Zr content

This contribution investigates fuel constituent segregation and fuel-cladding chemical interactions (FCCI) in three U-Pu-Zr fuel pins irradiated to ∼11% burnup in Experimental Breeder Reactor-II as part of the X441 DP-1 experiment. In examined pins, Zr content ranged from 6 to 10 and 14 wt.%, while Pu concentration was constant at 19 wt.% Pu. The primary goal of the investigation was to determine the role Zr content plays in fuel performance and this manuscript provides assessment of both constituent redistribution and FCCI as a function of axial position. FCCI was observed in all pins, though no cladding failure was noted. Optical and scanning electron microscopy (SEM) results show that variation in Zr content alters the number and relative size of constituent redistribution zones in the fuel. For example, larger Zr-rich central regions and smaller U-rich intermediate regions were observed in 14 wt.% Zr fuels. More importantly, all examined fuels have four or more discrete constituent redistribution regions, where each region has dissimilar morphological features. The existence of four to six distinctive regions deviates from traditionally accepted three region redistribution model , highlights complexity of constituent redistribution in this metal fuel , and identifies the need to conduct additional studies using state-of-the-art instrumentation.

Bulk and surface diffusion of neodymium in alpha-uranium: Ab initio calculations and kinetic Monte Carlo simulations

• A Nd interstitial is found to be intrinsically unstable in alpha- U . • The binding between the substitutional Nd and the U vacancy in strong in alpha- U . • Nd atoms migrate in bulk alpha- U via the vacancy mechanism. • Radiation can significantly accelerate the bulk diffusivity of Nd in alpha- U . • Diffusion of Nd adatoms via pore surfaces is much faster than bulk diffusion. A fundamental understanding of lanthanide transport in metallic fuels is critical for high fidelity modeling of the fuel-cladding chemical interaction (FCCI) phenomenon, which can lead to the formation of brittle intermetallic compounds and premature failure of the cladding. Here we report a combined ab initio density functional theory (DFT) and kinetic Monte Carlo (KMC) study of the bulk diffusivity of Nd in α-U , fully taking into account the effect of radiation enhanced diffusion. The vacancy mechanism is considered to be the dominant mechanism for the bulk diffusion of Nd since a Nd interstitial is found to be intrinsically unstable in α-U . The surface diffusivity of a Nd adatom on α-U (001) surface has been further predicted using KMC simulations parameterized by DFT calculations. The present study suggests that Nd transport via the surface diffusion mechanism can be many orders of magnitude faster than bulk diffusion. Furthermore, the results from the present lower length scale study can be used to inform mesoscale phase-field simulations to determine the effective diffusion coefficient of Nd through α-U with a porous microstructure.