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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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Site specific porosity-thermal performance correlations in neutron irradiated U-10Zr fuel

In this work, we present a site-specific three-dimensional analysis of porosity evolution in neutron-irradiated U–10Zr metallic fuel using high-resolution synchrotron X-ray tomography. Focused ion beam cubic lift-outs from four radial positions—fuel center, middle, edge, and fuel-cladding interaction (FCCI) zones—were imaged, reconstructed, and segmented to quantify pore volume, density, morphology, and pore connectivity. Porosity increased modestly from 5.5% at the center to 10.5% at the edge, yet pore number density increased by over two orders of magnitude in the fuel portion near the FCCI interface (from 4.8 x 10 4 to 6.0 x 10 6 pores/mm 3 ). Morphological classification revealed a progression from small spherical pores at the center to equiaxed and tortuous networks at the periphery, with enhanced orientation along the radial direction. Connectivity and permeability analysis reveal that the FCCI region maintains dense, highly interconnected pores despite a moderate volume fraction, enabling rapid fission gas transport and lanthanide migration. Effective thermal conductivity models incorporating sodium logging confirm that these site-specific pore features critically influenced heat transport during reactor irradiation. These findings demonstrate that pore topology—not just porosity fraction—controls thermal performance and thermal pathways in U–10Zr fuels, impacting fuel thermal performance in a reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Understanding Fission Gas Bubble Distribution and Zirconium Redistribution in Neutron-irradiated U-Zr Metallic Fuel Using Machine Learning

U-10wt.% Zr (U-10Zr) based metallic fuel is the leading candidate for next-generation sodium cooled fast reactor in United States. Currently, Idaho National Laboratory (INL) has been the leading national laboratory for research, development, and demonstration (RD&D) on metallic fuel. Advanced post-irradiation characterization will help to understand fuel microstructure and property change during irradiation, benefiting fuel qualification for commercial application. Characterization capabilities ranging from sub-nanometer to micrometer, such as scanning electron microscopy (SEM), focused ion beam (FIB) sampling, transmission electron microscopy (TEM) characterization, and local thermal conductivity microscopy (TCM), have been utilized recently on irradiated U-10Zr fuel samples to gain a better understanding of nuclear fuel microstructure and property evolution inside a reactor. The FIB/SEM coupled with energy dispersive X-ray spectroscopy (EDS) can capture the essential information to achieve better understanding of fuel behaviors. Inside a nuclear reactor, the phase and microstructure of U-10Zr is constantly changing under neutron bombardment. For example, the gaseous fission product atoms have a limited solubility inside fuel matrix and tend to precipitate out in bubble form, which not only contribute to fuel thermal conductivity degradation but also provide a shortcut for movement of fission products, i.e. lanthanides. The resultant deposition of lanthanides at the cladding inner surface will potentially trigger a chemical reaction/interaction between nuclear fuel and cladding at reactor operational conditions, threatening fuel integrity and safety. FIB/SEM coupled with EDS can provide the fission bubble information as well as probe into phase separation or Zr redistribution, which is fundamental to predict the fuel performance. With high velocity image data generating method, such as FIB/SEM, an automatic way to extract the microstructural information quantitively can better serve the needs from post irradiation characterization. A trained machine learning model, named Decision Tree, is employed to generate a bubble classifier and to categorize bubbles into three categories: isolated bubble, connected without lanthanides, and connected with lanthanides bubbles[3]. This work presents a showcase of this approach on six regions of a fuel cross-section along the radial temperature gradient. We obtained distributions of bubble categories and porosity rates along the six regions. Moreover, a secondary phase U-Zr2 was determined and found on regions 5 and 6. The secondary phase fraction was increasing from 15.61% in region 5 to 34.79% in region 6 based on this approach . This quantitative data offers insights into the lanthanide migration and potentially thermal conductivity degradation. This information from machine learning will be fed into fuel design code for better prediction of fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

The impact of chemistry on anion migration in bixbyite-structured lanthanide oxides

This manuscript describes atomistic calculations of oxygen vacancy and interstitial migration in bixbyite structured lanthanide oxides. We examine two types of compounds, one in which only one type of lanthanide cation is present and a second class in which two lanthanides are present in a 3:1 ratio as dictated by the symmetry of the bixbyite crystal structure. Using temperature accelerated dynamics and the nudged elastic band method, we quantify the role of chemistry on the energy barriers for the most important pathways for both vacancy and interstitial migration. We then analyze the impact of these variations on the overall diffusivity of each defect, quantifying the contribution of each pathway using the theory of kinosons. We find that vacancy mobility can vary by as much as three orders of magnitude through changes in chemistry at 500 K. Changes in interstitial mobility are more modest but can still vary by an order of magnitude. This points to the ability to tune the mass transport characteristics of these compounds through appropriate choices in chemistry. We have also included supplementary information containing the atomic structures of the relevant pathways.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Interactions and immobilization of lanthanides with dopants in uranium-based metallic fuels

In uranium-based fuels, production and migration of lanthanide fission products and subsequent chemical interactions with cladding constituents lead to fuel–cladding chemical interaction (FCCI), limiting fuel performance. In general, dopant addition to the fuel matrix to arrest lanthanides within the fuel by forming intermetallics is found to be effective in mitigating FCCI. Recently, we proposed ab-initio based alloy-design principles, which can be useful in identifying dopant(s) that can bind a lanthanide (Nd) inside the fuel matrix. Here, we demonstrate the robustness of such principles by performing a systematic study to choose dopants to form compounds with a range of experimentally-observed lanthanide fission products. Results reveal that for a given dopant, lanthanide Ce and La exhibit similar behavior in their compound-forming tendencies compared to Nd and Pr, which is linked to lanthanide intrinsic characteristics such as the electronic configurations. Our predicted lanthanide-dopant intermetallic formation is verified experimentally in selective cases. Finally, we showed alloy design principles that accurately identify previously known dopants like Pd, and also predict new dopants As and Se, that can be effective in binding all lanthanides within the uranium matrix. Altogether, this research helps to develop generic alloy-design principles for complex multi-component systems based on lanthanide and dopant intrinsic characteristics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

First Examination of Irradiated Fuel with Pulsed Neutrons at LANSCE (Preliminary Results)

We present preliminary results on the characterization of an irradiated U-lOZr-lPd fuel sample that was prepared from the irradiated AFC-3A-R5A sample. U-lOZr metallic fuels are researched as host materials for potential transmutation fuels and the addition of palladium strives to bind lanthanides, thus preventing fuel-cladding chemical interactions (FCCI). These interactions limit the lifetime of metallic fuels and are caused by migration of lanthanide fission products to the periphery of the fuel slug, where they start to interact with the D9 or HT9 steel cladding, ultimately leading to failure of the mechanical integrity of the cladding. Neutrons offer bulk characterization of irradiated materials for which X-ray tomography methods are not suitable due to the immense gamma background emitted from the samples. In particular pulsed neutrons provide information from the ability to resolve the neutron energy using their time-of-flight and thus the potential to utilize neutron absorption resonance to characterize the spatial distribution of isotopes. This, in turn, may allow to characterize the distribution of fission and neutron capture products non-destructively and may ultimately be applied to the bulk of an irradiation capsule prior to destructive post-irradiation examination to identify regions of interest. To allow the characterization of entire irradiation capsules, a cask is under development in the advanced post-irradiation work package at LANL and progress on this development was reported elsewhere. In parallel, an irradiated U-lOZr-lPd sample cut from the AFC-3AR5A irradiation was shipped to LANL and will be fully characterized with an NSUF funded rapid turnaround experiment (RTE) in the 2020 LANSCE run cycle. The sample emits at a dose rate of ~3R/hr on contact and is therefore manageable with remote handling, without requiring a cask. The disk-shaped material is larger than samples prepared for analysis using electron or X-ray methods and is therefore an intermediate step towards characterization of bulk samples at LANSCE. However, since it covers the full diameter of the irradiated fuel slug, some insight on redistribution of elements, spatially resolved information on microstructure, e.g. phase composition and texture, will be possible using the pulsed neutron-based methods developed for fuel characterization at LANSCE. This report describes the development of procedures to handle the sample at LANSCE as well as preliminary data and results from tests conducted in December 2019 on the energy-resolved neutron imaging (ERNI) beam line at flight path 5 and the high pressure-preferred orientation diffractometer (HIPPO) at LANSCE. This effort is a collaboration between LANL, INL, and ORNL. To compare our capabilities with prior work, we present an overview of previously reported bulk characterization of irradiated or spent fuels. The overview addresses neutron diffraction and neutron absorption resonance spectroscopy, both of which have only few reported applications on irradiated or spent nuclear fuel, as well as neutron radiography. This literature review was already described in a previous report but is repeated here to put our current efforts in context of previous work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Understanding fission gas bubble distribution, lanthanide transportation, and thermal conductivity degradation in neutron-irradiated α-U using machine learning

U—10Zr based metallic nuclear fuel is the leading candidate for next-generation sodium-cooled fast reactors in the United States. US research reactors have used and tested this fuel type since the 1960s and accumulated considerable experience and knowledge about the fuel performance. Most of the knowledge, however, remains empirical. The lack of mechanistic understanding of fuel performance puts a large burden on proof through experimental verification for the qualification of U—10Zr fuel for commercial use. Further, this paper proposes an image data-driven machine learning approach, coupled with domain knowledge provided by advanced post irradiation examination, to provide unprecedented quantified insights into the morphology, size, density and the connectivity of fission gas bubbles and their effects on the fission product transportation and thermal conductivity. Specifically, we developed a method to automatically detect, extract statistics, and classify ~19,000 fission gas bubbles into different categories, and quantitatively link the data to lanthanide transportation through connected bubbles and degradation of thermal conductivity along the radial temperature gradient in a neutron irradiated U—10Zr annular fuel. Results indicate the approach can be modified to study other irradiation effects, such as secondary phase redistribution and gaseous fuel swelling in other irradiated nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

An efficient instance segmentation approach for studying fission gas bubbles in irradiated metallic nuclear fuel

Abstract Gaseous fission products from nuclear fission reactions tend to form fission gas bubbles of various shapes and sizes inside nuclear fuel. The behavior of fission gas bubbles dictates nuclear fuel performances, such as fission gas release, grain growth, swelling, and fuel cladding mechanical interaction. Although mechanical understanding of the overall evolution behavior of fission gas bubbles is well known, lacking the quantitative data and high-level correlation between burnup/temperature and microstructure evolution blocks the development of predictive models and reduces the possibility of accelerating the qualification for new fuel forms. Historical characterization of fission gas bubbles in irradiated nuclear fuel relied on a simple threshold method working on low-resolution optical microscopy images. Advanced characterization of fission gas bubbles using scanning electron microscopic images reveals unprecedented details and extensive morphological data, which strains the effectiveness of conventional methods. This paper proposes a hybrid framework, based on digital image processing and deep learning models, to efficiently detect and classify fission gas bubbles from scanning electron microscopic images. The developed bubble annotation tool used a multitask deep learning network that integrates U-Net and ResNet to accomplish instance-level bubble segmentation. With limited annotated data, the model achieves a recall ratio of more than 90%, a leap forward compared to the threshold method. The model has the capability to identify fission gas bubbles with and without lanthanides to better understand the movement of lanthanide fission products and fuel cladding chemical interaction. Lastly, the deep learning model is versatile and applicable to the micro-structure segmentation of similar materials.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Experimental Investigation of FCCI Using Diffusion Couple Test Between UZr Fuel with Sb Additive and Cladding

Alloying additions are introduced into U-Zr fuel in order to bind lanthanides (e.g., cerium) and prevent their migration to the fuel-cladding interface. Antimony (Sb) is being investigated as a candidate additive. The present study focuses on the diffusion couple behavior of U-10Zr (wt%) alloy with Sb against cladding (iron or HT9) at 640°C. The diffusion cross sections were analyzed using a scanning electron microscope and X-ray diffraction. Zr-rind was found at the interface of the fuel alloy, Sb was found to be bound in Sb-Zr precipitates or Sb-Ce precipitates, and no reaction was found between Sb precipitates and the cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Atomistic modeling of lanthanide diffusion in refractory body-centered cubic molybdenum

Lanthanide fission products can strongly interact with candidate cladding alloys, but their transport properties in refractory metals remain poorly understood. Here, in this work, we investigate the atomic-scale diffusion behavior of La, Ce, Pr, and Nd in body-centered cubic (bcc) molybdenum, a potential candidate for advanced nuclear cladding. Self-consistent mean-field transport modeling is performed to evaluate the fission product transport and vacancy mobility, informed by first-principles and nudged elastic band calculations of vacancy formation energies, migration barriers, and solute–vacancy binding characteristics. Compared with bcc Fe, lanthanide solutes in bcc Mo exhibit slower tracer diffusion due to higher vacancy formation and migration energies. Furthermore, the calculations reveal that the influence of fission products on migration barriers in bcc Mo are not as extensive in range compared to bcc Fe. Among the studied lanthanides, La exhibits the strongest vacancy binding while also being the fastest diffuser in Mo. These findings highlight how refractory bcc alloys can reduce fission product infiltration, offering valuable insight into the development of durable cladding systems for advanced reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Cobalt-60, Barium-133, Cesium-137, and Europium-152 migration from cementitious sources through sediment under field conditions

Safe and effective storage of radioactive waste is essential to protect human and environmental health. Due to the potential for accidental releases and the severity of the associated risks, it is imperative to further understand radionuclide transport should an accident occur. This study was the second set of measurements conducted in 2022 of an ongoing experiment that has analyzed the vadose zone migration of radionuclides from cementitious wasteforms at the Savannah River Site over the last ten years. The radionuclides introduced within the sources are prominent constituents of radioactive waste or analogs for other groups or series of radionuclides. Lysimeters were first analyzed in 2016 using a collimated high-purity germanium gamma-ray spectrometer to non-destructively measure the concentration of each radionuclide in the sediment column as a function of depth. Following these measurements, the lysimeters were redeployed for another 4 years. All radionuclides in all lysimeters were observed to transport further during the redeployment period; however, the extent of migration varied with the material used for introduction. Except for 137Cs, migration through the sediment control system increased with decreasing ionic potential (ionic charge/radius); migration order: 152Eu<137Cs<60Co<133Ba. Overall, the cementitious wasteforms were observed to decrease radionuclide migration extent relative to natural vadose zone conditions. In both cementitious wasteforms, the migration extent increased in the order 152Eu<133Ba<60Co<137Cs. However, less migration was measured when the radionuclides were incorporated into a reducing grout wasteform. The novelty of this paper is the demonstration of a technique capable of creating non-destructive measurements over decade time scales. Ultimately, this work provides insight into the long-term migration of alkali, alkali earth, divalent transition metal, and trivalent (e.g., lanthanide and actinide element) isotopes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Parametric Study of Used Nuclear Oxide Fuel Constituent Dissolution in Molten LiCl-KCl-UCl 3

Prior work identified dissolution of used nuclear oxide fuel constituents from a uranium oxide matrix into molten LiCl-KCl-UCl 3 at 500°C, prompting a subsequent series of three progressive studies (including an initial scoping study, an electrolytic dissolution study, and a chemical-seeded dissolution study) to further investigate associated parameters and mechanisms. Thermodynamic calculations were performed to identify possible reaction mechanisms and their propensities in used oxide fuel constituent dissolution. Used nuclear oxide fuels with varying preconditions from fast and thermal test reactors were separately immersed in the subject salt system to assess fuel constituent migration from the bulk fuel matrix to the salt phase in an initial scoping study. Dissolution of expected fuel constituents, including alkali, alkaline earth, lanthanide, and transuranium oxides, into the chloride salt phase varied widely, ranging from 12% to 99% in the initial study. Uranium isotope blending between the salt phase and bulk fuel matrix was also observed, which was attributed to reducing conditions in the fuel matrix. Electrolytic and chemical-seeded dissolution studies were subsequently performed to effect reducing conditions in the fuel. Other parameters, including temperature (at 500°C, 650°C, 725°C, and 800°C) and uranium trichloride concentrations (at 6, 9, and 19 wt% uranium), were investigated in the latter two studies, resulting in fuel constituent dissolution above 90%. Extents of dissolution were based on initial and final fuel constituent concentrations in the oxide fuels following operations in the salt and subsequent removal of the salt via distillation. Finally, in this series of progressive studies, oxide fuel preconditioning and in situ reducing conditions, along with elevated temperature and uranium trichloride concentrations, were the primary parameters promoting used nuclear oxide fuel constituent dissolution in accordance with identified reaction mechanisms.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Flow Injection Electrochemical Quartz Crystal Microbalance with ICP-OES Detection: Electroprecipitation and Stripping of Lanthanum and Neodymium in a Flow Cell

We used an improved version of our hyphenated analysis technique, flow injection electrochemical quartz crystal microbalance (EQCM) with inductively coupled optical emission spectroscopy, to investigate the electroprecipitation of lanthanum and neodymium from flowing solutions. The improved version has two independent flow injection circuits, allowing different conditions for electrochemical precipitation (pH=4.3 or 2.6) and quick stripping (2% HNO3) with EQCM mass detection and ICP-OES elemental analysis. Lanthanum or neodymium ions in a 500 uL sample were injected into a carrier stream and deposited onto an EQCM housed in a micro flow cell, owing to a localized high-pH layer. The deposits are subsequently stripped from the electrode using HNO3 and analyzed downstream using an ICP-OES. We found that using acidic solutions without supporting electrolyte leads to an increase of the amount of lanthanum detected by 4-fold. The enhanced deposition can be attributed to enhanced mass transfer by migration. Furthermore, we showed that by applying chronopotentiometry, we can detect a change in the hydrogen evolution reaction mechanism that enables the precipitation of lanthanides on the surface of the electrode. Understanding and enhancing the deposition of lanthanides is relevant for elemental or isotopic detection in nuclear forensics.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

UV–Vis–NIR Reflectance Spectroscopy and Chemometrics for Monitoring Pu Directly on an Ion Exchange Column

Here, we present a fiber-optic UV–vis–NIR reflectance spectroscopy method for direct, noninvasive monitoring of Pu(IV) in a glass ion exchange column during dynamic loading and elution in a glovebox. A movable probe enables spatially resolved spectral acquisition along the column axis, capturing distinct features associated with Pu(IV) nitrate complexes during loading and free ions during elution. Principal component analysis was applied to extract the dominant spectral variance and resolve relative concentration profiles without requiring precise knowledge of optical penetration depth or species identity. This in situ approach reveals spatial gradients and speciation dynamics in real time, which provides actionable insight into Pu(IV) ion migration, resin saturation, and breakthrough behavior under evolving flow conditions. The method offers a practical, fiber-compatible strategy to monitor glass column–based separations for Pu and other lanthanides or actinides and to characterize metal–resin interactions in flow-through systems.

actinide↗
DOE OSTI2021

Accessing Lanthanide-to-Lanthanide Energy Transfer in a Family of Site-Resolved [Ln III Ln III '] Heterodimetallic Complexes

The ligand H 3 L (6-[3-oxo-3-(2-hydroxyphenyl)propionyl]pyridine-2-carboxylic acid), which exhibits two different coordination pockets, has been exploited to engender and study energy transfer (ET) in two dinuclear [Ln III Ln III '] analogues of interest, [EuYb] and [NdYb]. Their structural and physical properties have been compared with newly synthesised analogues featuring no possible ET ([EuLu], [NdLu], and [GdYb]) and with the corresponding homometallic [EuEu] and [NdNd] analogues, which have been previously reported. Photophysical data suggest that ET between Eu III and Yb III does not occur to a significant extent, whereas emission from Yb III originates from sensitisation of the ligand. In contrast, energy migration seems to be occurring between the two Nd III centres in [NdNd], as well as in [NdYb], in which Yb III luminescence is thus, in part, sensitised by ET from Nd. This study shows the versatility of this molecular platform to further the investigation of lanthanide-to-lanthanide ET phenomena in defined molecular systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

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. Further, 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Plutonium migration and phase evolution in irradiated U-Pu-Zr metallic fuels: An integrated EPMA-SEM-TEM study

Constituent redistribution is a defining feature of irradiated U-Pu-Zr metallic fuels, yet its mechanisms and effects on fuel performance are not sufficiently resolved to guide model development. Although decades of irradiation testing have established broad trends, a true mechanistic understanding of constituent redistribution has not been achieved. Here, in this study, we use electron probe microanalysis (EPMA), scanning electron microscopy (SEM), and transmission electron microscopy-based (TEM) selective area electron diffraction (SAED) on a EBR-II irradiated U-19 wt.% Pu-6 wt.% Zr fuel pin cross-section to correlate the composition, porosity, and crystallographic phases formed after irradiation. Constituent redistribution is thought to consist of three distinct zones, in which uranium and zirconium migrate while plutonium remains relatively unchanged. Our EPMA results resolve eight distinct compositional regions, and more importantly, show that plutonium redistributes alongside zirconium, contrary to historical assumptions. The distribution of fission products was highly asymmetric with a few large lanthanide precipitates observed at isolated sites on the pin periphery instead of a uniform distribution of smaller precipitates around the periphery. Using thermodynamic data from TAF-ID and the measured EPMA compositions, matrix phase fractions were predicted across the fuel radius. Phase predictions based on composition did not match TEM/SAED results, which revealed a much higher fraction of α−U phase than would be expected if phases were retained from reactor temperatures. These findings highlight the need for expanded SAED phase identification to capture post-irradiation and storage effects, as well as rigorous uncertainty quantification in fuel performance and phase diagram modeling to better constrain predictions from compositional data.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗