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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Heat transfer optimization of uo 2 -mo fuel using genetic algorithms

Two genetic algorithm (GA) methods were applied to thermal finite element models to optimize the heat transfer efficacy of a UO 2 -Mo composite fuel pellet with typical pressurized water reactor fuel geometry. Mo additions to UO 2 have been shown to increase the thermal conductivity, thus reducing centerline temperatures and temperature gradients. Previous studies evaluated uniformly dispersed Mo or continuous Mo internal geometries (e.g., fins, plates, discs) that were selected using engineering intuition. The current study uses two different implementations of the same GA to optimize Mo placement and minimize the fuel temperature with the only constraint being a maximum 10% Mo volume fraction. One approach superimposed Mo line elements onto the monolithic UO 2 pellet model, and the other converted entire UO 2 volume elements to Mo. The former method generated 1D heat transfer connections between nodes, whereas the latter method allowed for the formation of 3D structures. Features of the optimal fuel design produced by the GAs included dispersed Mo near the centerline that shifted the peak fuel temperature outward by 0.6 mm, Mo chains in the high-heat-flux region in the mid-to-outer radial zone, and a large continuous structure that spanned the full radius and height of the pellet and accounted for 87.7 % of the total Mo in the pellet. Analysis of this design indicates that the optimal Mo configuration is a balance between creating continuous heat transfer pathways and optimally dispersing Mo to minimize the heat transfer distance through UO 2 . This architecture ultimately produced an effective thermal conductivity of 11.3 W/m·K under the assumed boundary conditions. This result is higher than any previous values from the literature. In conclusion, potential fabrication methods and challenges are discussed in addition to the implications on fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Exploring competitive metal binding and crystallization of UO 2 2+ and Cu 2+ tetrahydrofuran-2,3,4,5-tetracarboxylic acid complexes

Solvent extractions are used to separate actinide elements from fission products in nuclear waste streams and the successful isolation of specific species utilize subtle differences in metal ligand binding. Metal ligand binding is also important in crystallization of metal organic materials and herein we explore the importance of competitive binding in the tetrahydrofuran-2,3,4,5-tetracarboxylic acid (THFTCA) system. This ligand has been previously evaluated for the selective extraction of uranium from other lanthanides and actinides and in the current research, we evaluate the crystallization of uranyl-THFTCA complexes in the presence of Cu 2+ , Sr 2+ , Th 4+ , and Ce 3+ . Three major phases were formed in the crystallization experiments and characterized with single crystal X-ray diffraction, powder X-ray diffraction, thermogravimetric analysis, and Raman spectroscopy. Two of the resulting phases were novel (UTHF1 ((C 4 H 10 N 2 )[UO 2 (C 8 H 8 O 9 ) 2 ]∙2H 2 O) and UTHF2 (Na[(UO 2 )(C 8 H 5 O 9 )(H 2 O)]∙ 3.5 H 2 O)), whereas the third (CuTHF1) was previously reported in the literature. Raman spectroscopy was utilized to evaluate spectral changes in the mother liquor of UTHF1, UTHF2, and CuTHF1 over time to assess the crystallization process. Further, isothermal titration calorimetry was used to determine the binding constants for UO 2 2+ and Cu 2+ to the THFTCA ligand in solution and evaluate the role of competitive metal binding in this system. The thermodynamic parameters and the crystallographic data were used to justify the formation of a weaker UO 2 2+ -THFTCA complex. Formation of a weaker UO 2 2+ -THFTCA complex supports our findings that UTHF1 only forms in homomeric systems, but suggests that the crystallization of UTHF2 and CuTHF1 is reliant on the presence of additional counter ions and ligands to change the amount of available THFTCA ligand.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

A novel symmetric pyrazine (pyz)-bridged uranyl dimer [UO 2 Cl 3 (H 2 O)(Pyz) 0.5 ] 2 2– : synthesis, structure and computational analysis

In this work, we report on the synthesis of (HPyz + ) 2 [UO 2 Cl 3 (H 2 O)(Pyz) 0.5 ] 2 ·2H 2 O which features a novel pyrazine-bridged uranyl dimer, [UO 2 Cl 3 (H 2 O)(Pyz) 0.5 ] 2 2– . A rigorous computational and experimental analysis of this compound was performed to fully explore the influence of coordination on the electronic structure and potential charge-transfer characteristics of this dimer, revealing a delocalized π-system across the bridging pyrazine and the axial components of both uranyl centers. Electrostatic surface potentials, used to rationalize the observed assembly, indicate a decreased basicity of the uranyl oxo versus [UO 2 Cl 4 ] 2– , and signify a lessened capacity for the terminal –yl oxos of the [UO 2 Cl 3 (H 2 O)(Pyz) 0.5 ] 2 2– dimer to participate in supramolecular assembly. A combined density functional theory (DFT) and quantum theory of atoms in molecules (QTAIM) analysis further evidenced an increase in U═O bond strengths within the dimer, which is supported by a blue shift in the characteristic Raman-active uranyl symmetric stretch (ν 1 ) with respect to the more typically observed [UO 2 Cl 4 ] 2– .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Development of a creep model informed by lower-length scale simulations to simulate creep in doped UO 2

Using molecular dynamics, we predict information at the atomistic scale used to develop a mechanistic UO 2 creep model for use in higher length-scale fuel performance codes. The ultimate objective of the model is to better describe the grain size dependence and therefore impact of doping on creep rates in UO 2 . In a previous NEAMS milestone, we found that Nabarro-Herring (bulk diffusional) creep was too low to capture the experimentally observed creep rates in standard UO 2 . Moreover, in that milestone, other mechanisms were explored, such as Coble (grain boundary) creep and dislocation climb, with each mechanism exhibiting different grain size dependencies. Again, these were orders of magnitude too low to describe the experimental creep rates. In this work, we address the previous assumptions made for the Coble creep mechanism by investigating the diffusivity of various defects at grain boundaries in UO 2 and, critically, to determine if enhanced grain boundary diffusivity allows the model to better reproduce experimental results. The diffusivity as a function of temperature for different concentrations of uranium vacancies and interstitials for bulk UO 2 have also been examined using cluster dynamics. Furthermore, using a concentration dependent segregation model, the concentration of defects at the grain boundary were predicted. This atomistic data was then input into the various creep mechanisms and the creep rates compared to the empirical MATPRO correlation (used in BISON) and experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Morphological Characterization of UO 2 F 2 and UF 4 Hydrolysis Products

Bulk samples of UO 2 F 2 and UF 4 were aged for 3 weeks under various relative humidity (RH) conditions to examine morphological changes associated with the hydrolysis of the materials. Aged samples were analyzed using scanning electron microscopy (SEM) to examine morphological changes, and Raman spectroscopy was used to correlate these changes with chemical composition. Little to no effect was evident for either material under lower RH conditions (i.e., 16% and 33%–38% RH). Under higher RH conditions (83% and 99%), multiple morphological changes were evident within the UO 2 F 2 samples because of the formation of hydrolysis products and possible interactions at the solid–vapor interface. Only one major morphological change was observed over the aging period—within the UF 4 system at 99% RH. This change can be attributed to the initial formation of anhydrous UO 2 F 2 and, subsequently, its hydrolysis products. After 37 weeks, UO 2 F 2 persisted as a hydrolysis product of UF4; however, no further hydrolysis products were observed. These results suggest that further research into a possible divergence of UF 4 and UO 2 F 2 degradation pathways is warranted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Neutronic analysis of a PWR-type SMR core using duplex ThO{sub 2}-UO{sub 2} in TRISO fuel particles

Currently, studies on small modular reactors (SMRs) present an important development due to the potential they represent in terms of safety, operational flexibility, economy, and non-energy applications. Furthermore, there is increasing interest in studying the use of thorium as fuel, as an effective way to solve problems such as the shortage of uranium reserves, reduction of nuclear waste and nuclear proliferation. Also, the use of thorium combined with highly enriched uranium TRISO particles has been studied, showing proper performance. In this work, the concept of ThO{sub 2}-UO{sub 2} duplex fuel is used, for the core configuration of a PWR type SMR that uses TRISO fuel, designed to achieve extended fuel cycles. Three distribution cases of ThO{sub 2} and UO{sub 2} in TRISO particles inside the fuel rods are compared. First, particles composed by a homogeneous mixture of ThO{sub 2} and UO{sub 2} are distributed inside of fuel elements. Second, the fuel zone of the fuel elements is divided into two radially, an internal one where the TRISO particles composed of ThO{sub 2} are distributed and the external one where the TRISO particles composed of UO{sub 2} are distributed. The third case is like the previous one, except that the particles that contain UO{sub 2} are distributed in the inner zone and those that contain ThO{sub 2} in the outer zone of the fuel elements. The comparison of the cases is carried out in terms of cycle main isotopes' mass transmutation, power distribution and temperature reactivity coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Anion photoelectron spectroscopic and relativistic coupled-cluster studies of uranyl dichloride anion, UO 2 Cl 2 –

A joint relativistic coupled-cluster and experimental photoelectron (PE) spectroscopic study of the uranyl dichloride anion, UO 2 Cl 2 – , is reported. Sophisticated electronic-structure calculations predict the photodetachment of UO 2 Cl 2 – to involve a U 5f electron and to be followed by significant geometry relaxation. Therefore, the adiabatic electron affinity (EA a ) of the uranyl dichloride neutral molecule, UO 2 Cl 2 , and the vertical detachment energy (VDE) of its anion, UO 2 Cl 2 – , provide valuable information about its uranium 5f orbital energies. The EA a value was computed to be 3.15 eV. The VDE value was calculated to be 3.55 eV by augmenting the computed EA a with a shift derived from a Franck–Condon simulation using coupled-cluster potential energy surfaces. Here, the VDE, which corresponds to the highest intensity peak in the PE spectrum, was measured to be 3.69 ± 0.20 eV, in good agreement with the computed value. The origin transition in the PE spectrum, whose electron binding energy corresponds to the EA a , was assigned to the feature at 3.2 ± 0.20 eV, consistent with the computed EA a .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Role of microstructure on CO corrosion of SiC layer in UO₂-TRISO fuel

The Advanced Gas Reactor Fuel Qualification and Development (AGR) program has focused on qualification of UCO kernel tristructural-isotropic (TRISO) particle fuel relative to UO₂ kernel TRISO particle fuel. However, a UO₂ kernel variant was included in the second AGR irradiation experiment (AGR-2) for comparison and to connect to historic fuel irradiation data. The development of a multiscale, post-irradiation examination (PIE) analysis approach through the AGR Program has allowed for a comprehensive understanding of individual particle failure. This approach has been applied to gain an understanding of SiC layer failure in UO₂ kernel TRISO fuel from AGR-2 after safety testing at 1600–1700 °C. Particle failure by intergranular CO corrosion, facilitated by a compromised inner pyrolytic carbon layer, has been confirmed through the combined application of modern x-ray computed tomography and electron microscopy techniques. In addition, a relationship between grain boundary character and CO corrosion has been identified. This finding provides an opportunity to develop mitigating strategies to improve the resilience of the SiC layer to CO corrosion in UO₂ TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural evolution of Mo-UO 2 cermets under high temperature hydrogen environments

Ceramic-metallic (cermet) materials show promise for use in nuclear thermal propulsion applications due to attractive thermophysical properties including high temperature stability and high thermal conductivity. In this work, molybdenum-uranium dioxide (Mo-UO 2 ) cermet fuel elements were fabricated by means of spark plasma sintering (SPS) and were subsequently exposed to hydrogen at high temperatures (2500 K). Mo-UO 2 samples pre- and post-exposure were characterized by means of optical microscopy, scanning electron microscopy, and X-ray diffraction (XRD). Microscopy analyses of the as-produced material displayed microscopic cracking on the interior of the spherical UO 2 fuel particles but confirmed that the fuel particles were fully encapsulated in the Mo matrix. The results further showed mass loss, macroscopic swelling, and cracking in the cermet samples which occurred during high temperature hydrogen testing. Nanoscale swelling was evidenced by XRD in the Mo matrix and UO 2 fuel structure due to the incorporation of defects and accompanied microstrain.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Mechanistic verification of empirical UO 2 fuel fracture models

Standard UO 2 fuel pellets used in light-water reactors fracture during irradiation due to the large thermal gradient in the radial direction. Over the decades, numerous researchers have explored fuel cracking from experimental and modeling points of view. To date, there have been both empirical and mechanistic approaches to predict the number of fragments that form in UO 2 . The empirical models only consider maximum power and burnup as inputs. Existing mechanistic approaches for normal operation have not accounted for irradiation effects. Here, this work employs a mechanistic fuel cracking model using the extended finite element method to explore radial crack formation while including a sensitivity analysis that accounts for the randomization of tensile strength within the fuel, the strength randomization criteria (uniform or volume-weighted Weibull), power ramping rates, computational mesh density, maximum power level, and irradiation (burnup) effects. The results indicate that the uncertainty in this mechanistic modeling approach envelopes the predicted values from three different empirical correlations in almost all cases. This means that, for computationally intensive analyses involving UO 2 fragmentation, the empirical correlations can be used. However, since the mechanistic calculations bound those of the empirical correlations, there is confidence in the applicability of the mechanistic approach developed in this work to generate a correlation for fuel types where limited data exists (e.g., doped-UO 2 , U 3 Si 2 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Enhancing thermal conductivity of UO 2 with the addition of UB 2 via conventional sintering techniques

We report that uranium dioxide has been the primary fuel type used in light water reactors for more than 40 years and proven to be reliable and robust. However, the Fukushima-Daiichi nuclear accident has motivated new work evaluating fuels with characteristics promoting accident tolerance, including enhanced thermal conductivity. Recently, additives have been investigated to increase thermal conductivity, but research has been largely focused on non-fissile additions. This study investigated the use of fissile additives to not only increase the thermal conductivity but also increase the uranium loading. Uranium diboride was chosen as the additive for this study due to its promising corrosion behavior as well as its significantly higher thermal conductivity at 573 K (25 Wm -1 K -1 ) when compared to UO 2 (7 Wm -1 K -1 ). Uranium diboride powder was fabricated via the arc melting technique and a ball milling process prior to mixing with UO 2 in a 90/10 wt% UO 2 /UB 2 ratio. Green bodies were made using a uniaxial die and subjected to a traditional pressureless sintering technique at 2073 K in argon. Sintered samples were analyzed via laser flash analysis for thermal diffusivity and differential scanning calorimetry for specific heat capacity in order to calculate thermal conductivity. The samples displayed an increase of 36-55% in thermal conductivity between 323 K and 1273 K when compared to the benchmark samples (pure UO 2 ) as reported in open literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

On the product phases and the reaction kinetics of carbothermic reduction of UO 2 +C at relatively low temperatures

The synthesis of UC using carbothermic reduction of UO 2 and C mixtures has been well studied at high temperatures. However, the product phase behavior of carbothermic reduction at low temperatures (≤1773 K) is not well studied. Such a study is important as low temperatures permit single phase UC synthesis without forming secondary higher carbides, and it further supports the knowledge base of the process that needs to be used for transuranic elements such as plutonium that have high vapor pressures at elevated temperatures. Therefore, a low temperature carbothermic reduction of two different C/UO 2 molar ratios under inert and reducing environments have been studied here. Two different sample holding crucibles, alumina (Al 2 O 3 ) and graphite, were also used here to differentiate the hypostoichiometric (UC 1-a ) and oxygen dissolved (UC 1-x O x ) uranium monocarbide phases adding more details on the two systems. Also, the reaction kinetics involved in the formation of UC via the carbothermic reduction of UO 2 +C using product phases instead of evolved gases such as carbon monoxide is reported here. Under inert atmospheres but with significant oxygen partial pressures, the low temperature carbothermic reduction of UO 2 +C produced up to 90 wt.% UC 1-x O x type oxycarbides as was confirmed by Xray powder diffraction. Reducing Ar-4%H 2 environments at these temperatures were not successful in synthesizing UC as it reduces the amount of C required for the carbothermic reduction, leaving UC phase at a non-equilibrium state. Inert atmospheres with low or negligible oxygen partial pressures on the other hand produced near stoichiometric UC at high phase purity, especially at 1673 – 1773 K temperature range. An activation energy of 377±75 kJmol -1 was also calculated using product phase concentrations of the carbothermic reduction of UO 2 +C under these inert Ar (g) atmospheres.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Core design and performance of the Westinghouse lead fast reactor with UO 2 and MOX configurations

For this work, Westinghouse partnered with Argonne National Laboratory to design, model and optimize UO 2 - and MOX-fueled core designs for a medium size (950 MWt) Lead Fast Reactor that was pursued by Westinghouse. Using Argonne’s suite of reactor analysis codes together with Westinghouse fuel cost economic models, thousands of candidate cores were considered to achieve the economics-optimized cores presented in this paper. This optimization process considered detailed reactor physics, fuel performance, transient performance, and fuel economics models. The reactor performance of the resulting optimized UO 2 - and MOX-fueled core designs are described and compared in this paper. Both cores show fuel performance and transient behavior that is considered acceptable for the optimization presented herein, while further testing campaigns on material performance in high-temperature liquid lead will be required to confirm acceptability at the operating conditions chosen. A multi-batch strategy was selected for the UO 2 core for best fuel utilization with minimum fuel inventory costs. A single-batch fuel management was instead selected for the MOX core to maximize cycle length and minimize the impact of the longer refueling outage resulting from the higher decay heat of the discharged MOX fuel relative to the discharged UO 2 fuel, requiring a longer cooling time before dry-lift of discharged fuel could take place.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Reactions of Studtite UO 4 ·4H 2 O in Alkali Hydroxides: Isolation of Single-Crystal Uranate Phases under Mild Hydrothermal Conditions

Uranyl peroxide complexes, particularly studtite (UO 4 ·4H 2 O), are important phases within the nuclear fuel cycle, forming through radiolysis-induced reactions on the surfaces of spent fuel and in waste environments. Studtite has been identified in Hanford’s irradiated fuel storage basins, Chornobyl’s corium lavas, and is anticipated on Fukushima’s fuel debris. While its formation and stability have been extensively studied, the reactivity of UO 4 ·4H 2 O in highly alkaline environments such as those encountered in high-level nuclear waste remains underexplored. These environments contain molar concentrations of [OH - ] and present a chemically dynamic and reactive environment where actinide behavior is not well understood. Reported here are investigations of uranium reactions in concentrated alkali hydroxides under mild hydrothermal conditions (<200 °C) that have resulted in the isolation of the alkali metal uranates Li 2 UO 4 , α-Na 2 UO 4 , γ-Na 2 U 2 O 7 , and K 2 U 2 O 7 . In contrast to conventional solid-state methods (>800 °C) that typically yield polycrystalline powders, our approach enables the isolation of these uranates as single crystals, allowing us to provide single-crystal structure solutions of certain uranates for the first time. Our findings match the results of high-level waste tank sampling, confirming that sodium diuranate (Na 2 U 2 O 7 ) is the most persistent uranium phase, potentially forming through reactions of uranyl peroxide intermediates with NaOH under radiolytic and highly alkaline conditions.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

An Atomic-Scale Understanding of UO 2 Surface Evolution during Anoxic Dissolution

Our present understanding of surface dissolution of nuclear fuels such as uranium dioxide (UO 2 ) is limited by the use of non-local characterization techniques. In this paper, we discuss the use of state-of-the-art scanning transmission electron microscopy (STEM) to reveal atomic–scale changes occurring to UO 2 thin films subjected to anoxic dissolution in deionised water. No amorphisation of the UO 2 film surface during dissolution is observed, and dissolution occurs preferentially at surface reactive sites that present as surface pits which increase in size as the dissolution proceeds. Using a combination of STEM imaging modes, energy-dispersive X-ray spectroscopy (STEM-EDS), and electron energy loss spectroscopy (STEM-EELS), we investigate structural defects and oxygen passivation of the surface that originates from the filling of the octahedral interstitial site in the centre of the unit cells and its associated lattice contraction. Taken together, our results reveal complex pathways for both the dissolution and infiltration of solutions into UO 2 surfaces.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Impact of Controlled Storage Conditions on the Hydrolysis and Surface Morphology of Amorphous-UO 3

The hydration and morphological effects of amorphous (A)-UO 3 following storage under varying temperature and relative humidity have been investigated. This study provides valuable insight into U-oxide speciation following aging, the U-oxide quantitative morphological data set, and, overall, the characterization of nuclear material provenance. A-UO 3 was synthesized via the washed uranyl peroxide synthetic route and aged based on a 3-factor circumscribed central composite design of experiment. Target aging times include 2.57, 7.00, 14.0, 21.0, and 25.4 days, temperatures of 5.51, 15.0, 30.0, 45.0, and 54.5 °C, and relative humidities of 14.2, 30.0, 55.0, 80.0, and 95.8% were examined. Following aging, crystallographic changes were quantified via powder X-ray diffraction and an internal standard Rietveld refinement method was used to confirm the hydration of A-UO 3 to crystalline schoepite phases. The particle morphology from scanning electron microscopy images was quantified using both the Morphological Analysis of MAterials software and machine learning. Results from the machine learning were processed via agglomerative hierarchical clustering analysis to distinguish trends in morphological attributes from the aging study. Significantly hydrated samples were found to have a much larger, plate-like morphology in comparison to the unaged controls. Predictive modeling via a response surface methodology determined that while aging time, temperature, and relative humidity all have a quantifiable effect on A-UO 3 crystallographic and morphological changes, relative humidity has the most significant impact.

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

Polymorphism and phase transitions in layered uranium(VI) hydroxides: Ab initio lattice dynamics simulations of UO 2 (OH) 2

The phase transitions and thermodynamics of stoichiometric α-, β-, and γ-UO 2 (OH) 2 polymorphs are investigated using density functional perturbation theory. The pressure-induced β(Pbca) → α(Cmca) phase transition is reproduced by calculations, with a volume reduction of ΔV/V = -14.7% similar to experiment. Consistent with observation, a temperature-driven γ(P2 1 /c) → β(Pbca) phase transition is predicted near 533 K. At 298.15K, the computed standard molar heat capacity of α-UO 2 (OH) 2 is C p 0 = 112.1 J mol -1 K -1 , only 1.6% smaller than the value of C p 0 = 113.96 ± 0.12 J mol -1 K -1 measured by calorimetry. C p 0 = 112.4 and 104.8 J mol -1 K -1 are predicted for the β- and γ-UO 2 (OH) 2 polymorphs, respectively. The calculated molar enthalpy and Gibbs energy functions of the α-, β-, and γ-UO 2 (OH) 2 polymorphs are also reported.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2022

Advanced Microscopy for Characterization of High Burnup Commercial UO 2 Fuel Before and After LOCA Testing

This report presents the results from the microstructural examination of high burnup UO 2 under pre- and post- loss of coolant accident (LOCA) conditions. The work in this report builds and expands upon work previously conducted on high burnup UO 2 fuel at Oak Ridge National Laboratory (ORNL). This work investigates the microstructural evolution that occurs during steady state irradiation of UO 2 , particularly in high burnup fuel. Several high burnup fuel samples were available from historic fuel shipments to ORNL. Microstructural characterization will aid in understanding why the fuel is susceptible to fuel fragmentation and support future work to investigate the formation mechanisms of certain features that form under irradiation. High burnup UO 2 microstructural data is also needed to improve constitutive models intended to predict high burnup fuel fragmentation (HBFF). This report summarizes the microstructural evaluation of these fuel samples to interpret and analyze high burnup fuel microstructures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗