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Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6 ± 1.6 and 5.7 ± 1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6 ± 1.1 kJ / mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Metallic fuels have seen increased interest for future sodium fast reactors due to their material properties: high thermal conductivities and advantageous neutronic properties allow for greater fission densities. One drawback to typical metallic fuels is zirconium redistribution, which impacts this advantageous material and its neutronic properties. Unfortunately, the processes behind zirconium migration behavior are understood using first principles, so before these fuels are implemented in future fast reactors, characterization and fuel qualification regimes must be completed. These activities can be supported through the use of robust modeling using the most accurate empirical models currently available to fuel researchers around the world. The tool that allows researchers to model this complex coupled thermo-mechanical behavior and nuclear properties is BISON. Additionally, BISON model parameters need to be compared against PIE measurements. The current work utilizes two fuel pins from EBR-II experiment X441 to optimize various model parameters, including porosity correction factor, thermal conductivity, phase transition temperature, and diffusion coefficient multipliers, before implementing the final model for seven fuel pins with differing characteristics. To properly evaluate the BISON simulations, the results are compared to PIE metallography data for each fuel pin, to ensure the zirconium redistribution is properly reflected in the simulation results. Six out of seven analyzed fuel pins demonstrate good agreement between the metallography images and BISON results, showing alignment of the Zr-rich, Zr-depleted, and moderately Zr-enriched zones at various axial heights along the fuel pins. Further work is needed to refine the model parameters for general pin use. [ABSTRACT FROM AUTHOR]

• Fuel performance in a Cd shrouded thermal reactor were compared to a fast reactor. • Postirradiation examinations validated the Cd shrouded thermal reactor irradiation. • Fuel microstructure and Zr redistribution is in line with previous literature. • Transmutation fuel showed good fuel performance. • FCCI was affected by Am presence. To optimize nuclear waste repository performance, the destruction of minor actinide elements, particularly Np and Am, in a neutron fast spectrum reactor is possible by incorporating these elements into nuclear fuel. Evaluating the performance of minor actinide containing fuel is of paramount importance to enabling this technology. However, such a task is challenging without an available domestic fast spectrum test reactor. A comparison of fuel performance tested in an available domestic thermal reactor at the Idaho National Laboratory, the Advanced Test Reactor, and in a fast spectrum reactor in France (Phénix) is presented in this study. This study evaluates the capability of using a cadmium shrouded test position to mimic the power profile along the fuel radius present in fast spectrum reactors so that thermally driven phenomenon (e.g., constituent redistribution) can be evaluated in a thermal reactor and determined to be prototypical of a fast reactor. Thus, optical microscopy and scanning electron microscopy has been performed on irradiated 35U-29Pu-4Am-2Np-30Zr fuel samples (where the number preceding the element is the weight percent concentration) from the two mentioned reactors that present similar irradiation temperatures and power conditions. The results indicate that fuel performance phenomena are reproducible in the two irradiation conditions. The redistribution of Zr occurred in the same manner for the two samples. Similar partitioning of U-Pu-Zr phases was observed, and the behavior of Am was similar in the analyzed specimens. Finally, the overall microstructure evolution seems not to be affected by minor actinides addition compared to expected behavior of conventional U-19Pu-10Zr ternary metal fuels for both specimens. Slight differences in fuel cladding chemical interaction were, however, observed. This difference is likely driven by difference in cladding composition rather than irradiation conditions. [ABSTRACT FROM AUTHOR]

● An extended overpower transient (∼32%) was conducted in EBR-II to test the response of 19 binary- and ternary-design metallic alloy fuel elements. ● Post-irradiation examination demonstrated that the overpower transient does not compromise the integrity of fuel elements. ● Post-transient fuel performance should not be significantly altered by the overpower transient. A slow-ramp (0.1%/s), extended overpower (∼32%) transient was conducted in the Experimental Breeder Reactor-II (EBR-II) on 19 metallic alloy fuel elements, including both binary (U-Zr) and ternary (U-Pu-Zr) fuel designs. The elements were clad in either Type 316 SS or HT9. Before the transient, the elements were pre-irradiated under steady-state or steady-state plus duty-cycle (periodic 15% overpower transient) conditions in various EBR-II sub-assemblies and reached burnups ranging from ∼5 to ∼12 at%. The fuel pins were then consolidated into the X512 sub-assembly for the overpower transient. Cladding integrity was maintained for all fuel pins throughout the entirety of normal operation and the overpower transient test. Post-irradiation examination (PIE) demonstrated that mild restructuring was evident within the fuel phase. However, these changes were not significant enough to expect changes in irradiation performance under normal operating conditions if the fuel elements were returned to service. These results empirically validate that limits on sodium fast reactor operational transients (e.g. power ramps) could be applied with adequate safety margins to fuel failure or non-reversible damage. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. Here, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal (U ,Zr) 6 Fe phase (space group I4/mcm) and cubic (U ,Zr)(Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The (U ,Zr)(Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications. [ABSTRACT FROM AUTHOR]

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system. The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This work reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution. Fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems. • The X501 experiment demonstrated potential alloys for minor actinide transmutation. • Performance of the X501-G591 metallic fuel pin is in line with historical experience from EBR-II driver fuel. • Fuel cladding chemical interaction (FCCI) characterization revealed Am infiltration into the cladding. [ABSTRACT FROM AUTHOR]

A fuel performance phenomenon that has previously limited the peak cladding temperature of metallic fast reactor fuel is fuel/cladding chemical interaction (FCCI). This study examined the differences in the phenomenon exhibited by the more recent fuel alloy, U-10Zr and a previously used fuel alloy, U-5Fs (Fs – fissium). Fs is comprised of Mo, Ru, Pd, Zr and Rh simulated fission products representing reprocessing carryover. Studies on the U-Zr alloy had shown the often controlling influence on FCCI of lanthanide (Ln) fission products (Ln: Ce, Nd, Sm, etc.) while a review of the U-Fs alloy behavior indicated a dominance of basic fuel/cladding interdiffusion (Fe, Ni with U). The recent studies on U-Zr fuel showed that migration of lanthanide fission products to the fuel/cladding interface facilitated the Ln dominance of FCCI behavior. This study was focused on understanding why Ln's do not migrate in the U-5Fs alloy to the extent that they do in U-10Zr, or why they do not interact with the cladding like in U-10Zr fuel pins. A small fuel/cladding interaction layer was found containing the cladding and lanthanide components. However, it was also found that the lanthanide fission products are at least partially bound to agglomerates that also contained Pd, Rh and Zr, all minor components of Fs. [ABSTRACT FROM AUTHOR]

• In-situ heating studies by transmission electron microscopy were performed on uranium-zirconium specimens. • The original lamellar structure transformed during the rapid tests. • Changes started at a high temperature in the zr rich phase. • Formation of globular α-Zr nanograins occurred in the original Zr-rich phase. • The u phase retains the α-phase after cooling. The development of U-Zr metallic nuclear fuel for fast spectrum reactors is impacted by a lack of mechanistic understanding of the fuel behavior evolution under thermal irradiation conditions, despite previous works providing substantial fuel performance data. This work uses in-situ transmission electron microscopy heating experiments to study phase and microstructural evolution in several unirradiated U-10Zr specimens during rapid heating ramps (from room temperature to 1000 °C). The starting α- U + bcc-(Zr, U) eutectic microstructure began to decompose above 600 °C. The decomposition initiated from the bcc (U,Zr) phase. Similar results were observed for all specimens even when fabricated by different routes (e.g., cold rolled or annealed). The impact of observed microstructure and phase evolutions at high temperatures on fuel fabrication and in-pile fuel transient test was also discussed. [ABSTRACT FROM AUTHOR]

Chromium (Cr) has been recognized as a promising diffusion barrier candidate to mitigate the Fuel Cladding Chemical Interaction (FCCI) failure in metallic fuels for sodium-cooled fast reactor. This paper, for the first time, conducted an in-depth post-irradiation examination of the microstructure and composition evolution, and micromechanics of the Cr diffusion barrier in U-10Zr fuel/HT9 cladding irradiated in the Advanced Test Reactor (ATR) to 8.7% burnup at an averaged Peak Inner Cladding Temperature of 540–550 °C. Transmission Electron Microscope (TEM) characterization confirmed the preferential intergranular diffusion of Zr and U in the Cr diffusion barrier, suggesting that grain boundaries served as fast path for the diffusion of Zr and U into the Cr diffusion barrier. The interaction zone is dominated by nano crystalline α- ZrCr 2 Laves phase. Despite the interaction, there is no microcracks being observed in the preserved Cr diffusion barrier and HT9 cladding, serving as a good barrier to mitigate FCCI under the studied in-reactor irradiation conditions. High density cavities in uniform distribution are observed inside Cr grains, nano particles contains Cr, Mn, and O are confirmed by Electron Energy Loss Spectroscopy (EELS) analysis. However, it is unclear whether the cavities will become an issue to the barrier integrity at higher burnup. In-situ Scanning Electron Microscopy (SEM) micro-tensile testing uncovered mechanical softening in the HT9 cladding nearing the Cr diffusion barrier, possibly due to the coarsening of lath structure and carbides precipitates. [ABSTRACT FROM AUTHOR]

Metallic fuels, particularly U 10Zr, are promising candidates for next-generation sodium-cooled fast reactors. Irradiation of nuclear fuels in reactors can lead to the formation of solid and gas fission product which subsequently forms microstructural pores, deteriorating fuel performance. Due to the massive amount of pores and complex phases formed, a quantitative description of fission gas pores is not yet available, preventing the development of microstructure-informed fuel performance modeling for fuel qualification. This paper applied a pre-trained deep learning model to ∼10,260 high magnification scanning electron microscopy images. This method increased the accuracy of fission gas pore segmentation and allows statistical features to be extracted which cannot be achieved manually. A pre-trained decision tree model worked on the segemenation resutls and further classified the pores into different categories to produce a correlation between the pores, movement of lanthanides, and temperature gradient during irradiation. This paper emphasizes the potentials of machine learning models to accelerate fuel research, development, and qualification for advanced reactors. [Display omitted] • Deep learning model is used to detect pores in ∼10,260 scanning electron microscopy images. • About 230,000 pores are detected and classified based on their impact on nuclear fuel performance. • Pore statistics and orientations help to understand fuel performance from post-irradiation examination. [ABSTRACT FROM AUTHOR]

• The influence of an overpower transient on fuel performance was evaluated. • Advanced characterization was performed on the top of an irradiated metallic pin. • For the first time the metallic "fluff" structure was examined in this study. • Fission products behavior, FCCI and fuel composition was evaluated in this region. • BISON transient's temperatures simulations were compared to material behavior. The influence of a reactor overpower transient on irradiated metallic fuel performance was investigated in this work. Such transient studies support safety and performance optimization of future sodium fast reactors. A ternary metallic fuel alloy (U-19Pu-10Zr) in HT-9 cladding which was irradiated in the Experimental Breeder Reactor II (EBR-II) to a burnup of 11 at. %, and then subjected to a 30% overpower transient was studied. The sample was selected from the top of a fuel pin where failure is expected to first appear in the EBR-II pin design. This can occur when the pin is subjected to conditions beyond its design envelope. Main elemental distribution, phase characterization, fission product behavior, chemical form and fuel-cladding chemical interaction were analyzed in this study. Simulation of the temperature profile during the transients were performed using BISON and show that the moderate transient did not maintain prolonged high temperature. For the first time, the porous structure commonly found at the top of metallic fuel pin (termed "fluff" structure given its appearance) was examined in detail via postirradiation characterization. The fluff contains the major fuel elements (U, Pu, Zr) and presents a highly porous microstructure (over 40% area fraction). The presence of fission products relevant to source term was investigated. However, semi-volatile and volatile elements (namely Cs, I, Xe, Eu), expected in this region, could not be detected probably due to sample preparation. Europium, also expected in this region, was only detected in precipitates along with the other rare earth elements. Finally, no fluff (fuel) fragments detached from the fuel slug were detected in the plenum region, indicating the fluff stability under the tested moderate ramp. These analyses indicate that the moderate transient experienced by the fuel pin did not significantly influence fuel performance under this transient condition. [ABSTRACT FROM AUTHOR]

The present lack of a domestic fast neutron flux irradiation capability combined with continued development of fast reactor fuels in the U.S. motivated an innovative engineering solution to utilize a unique neutron flux tailoring capability in the Advanced Test Reactor at the Idaho National Laboratory. To achieve the objectives of the fast reactor fuel irradiation tests, the incident neutron flux was hardened substantially by placing fueled irradiation capsules inside specially designed cadmium shrouds. Use of cadmium prevents thermal neutrons from reaching the fuels being tested and alleviates the plutonium self-shielding that would normally arise during irradiations of high density, highly enriched fuels in a thermal neutron spectrum. The present paper illustrates the profound effect this engineered solution has on the efficacy of the experiments. Based on the comparison of post-irradiation measurements of the columnar grain region in fast reactor mixed-oxide fuels with fuel performance calculations, it is demonstrated that thermal conditions achieved in these cadmium-shrouded fuel experiments are substantially prototypic of a sodium fast reactor and are suitable for concept-screening tests supporting development of new fast reactor fuels. It is also shown that if the experiments were conducted in an unmodified ATR neutron spectrum, gross plutonium self-shielding would cause a strong depression of the fission power at the fuel centerline preventing fuel restructuring, a hallmark feature of mixed oxide fuel behavior under fast reactor conditions. Recognizing the need for testing metallic fuels for fast reactors, the impact of neutron flux energy spectrum on the radial temperature distribution in metallic fuel is investigated. It is shown that the use of Cd shrouds allows to attain radial temperature distributions nearly identical to those that exist in an SFR. [ABSTRACT FROM AUTHOR]

To support the development of U-10 wt.% Zr (U-10Zr) metallic fuel for Generation IV sodium-cooled fast reactors, we analyzed a Na-bonded solid U-10Zr fuel cross-section that was irradiated to a burnup of approximately 13.2 at.% at the Fast Flux Test Facility (FFTF). Advanced characterization techniques, including site-specific sample preparation by focused ion beam (FIB) and scanning transmission electron microscopy (STEM), were used to reveal the Zr redistribution and characterize the fuel matrix and secondary phases (such as solid fission products) present at the end of life. Results showed that the fuel pin cross-section is divided into three major concentric zones: a Zr-rich central region, a Zr-lean intermediate region, and a Zr intermediate peripherical region. The phase characterization revealed that the irradiation environment enhanced the development and stabilization of phases not predicted by the standard equilibrium U-Zr phase diagrams. Comparing the current results with the ones from previous studies, it is reaffirmed that the radial temperature profile and the time spent in the reactor, rather than the fuel burnup, are the two factors that most influence the formation of redistribution zones and their extension along the fuel cross-section. Various solid fission products, such as lanthanides, ZrRu, BaTe, CsI, and Ba and Sr oxides, precipitated inside the fission gas pores. This study provides unprecedented nanoscale understandings in the irradiated U-10Zr fuel system that may benefit fuel performance modelling and advanced fuel development. [ABSTRACT FROM AUTHOR]