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DOE OSTI2021

Fission product distribution in irradiated safety-tested and as-irradiated AGR-2 TRISO particles

In this work, two tristructural isotropic (TRISO)-coated nuclear fuel particles were examined by electron probe microanalysis (EPMA) as part of the Advanced Gas Reactor program. The compacts’ average irradiation temperatures ranged from approximately 1260 to 1290 °C. One particle was examined in the as-irradiated condition, while the other was subject to 1600 °C post-irradiation safety testing. This study was undertaken to test a newly-developed EPMA technique to determine fission product masses in TRISO particles on a layer-by-layer basis, and to compare fission product distributions between an as-irradiated and safety-tested particle. Fission product concentration profiles were collected along two radii in each particle, with measured concentrations used to compute the fission product mass in each TRISO particle layer. These measured masses were then compared to those predicted from ORIGEN modeling calculations. Data collected from these measurements show that for these two particles, masses determined via EPMA were within ± 20% of the calculated masses for the rare-earth elements, Mo, Zr, Cs, I, and Pd. Elements that tend to be less homogeneously distributed include Sr, Te, Eu, Ag, and possibly Ba. Measured Ag masses differed by more than 40% from the calculated mass. Lanthanides other than Eu remain primarily within the fuel kernel in the as-irradiated particle but in the safety-tested particle these element masses were divided approximately equally between the kernel and kernel periphery. In both particles, the majority of Sr and Eu accumulated in the carbon-rich kernel periphery, although in the safety-tested particle, Sr and Eu accumulated farther from the fuel kernel than occurred with irradiation alone. A greater mass fraction of mobile elements, such as Cs and I accumulated in the buffer and IPyC in the safety-tested particle as compared to the as-irradiated particle. When fully developed and tested, this mass balance approach to TRISO particle analysis has the potential to provide insight into fuel behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

A Machine Learning Method for the Forensics Attribution of Separated Plutonium

Plutonium (Pu) source attribution would be a powerful tool to support nuclear nonproliferation efforts. This capability to find the source of a Pu sample would act as a deterrent to smuggling efforts, and also help regulatory agencies verify declared nuclear activities. Work at Texas A&M University yielded a nuclear forensics methodology, which is capable of determining separated Pu’s reactor of origin, fuel burnup, and the time since irradiation (TSI)—three parameters of interest. The methodology used a set of ten intra-element isotopic ratios found in separated Pu, which was compared to a library of isotopic ratio values produced using neutronics simulations for reactors of interest. By calculating the probability that unknown Pu sample’s isotopic ratio set matched a set in the library, the methodology could predict the three parameters of interest of the sample. One shortcoming of this methodology was an inability to correctly attribute spoofed Pu, where Pu sourced from two different reactors or two different fuel burnup levels are mixed. A new methodology to rectify this vulnerability using machine learning (ML) technique is developed, instead of the maximum likelihood calculation previously used and the results are satisfactory. The ML approach leverages the existing simulated data for training the algorithm, but use them efficiently by only using intra-element isotope ratios that contribute to the attribution one of the three parameters at a time. Previously, all isotope ratios were used to attribute all three parameters together. The new methodology attributes the Pu parameters in three steps, one for each parameter, rather than resolving all of the three parameters simultaneously like the previous maximum likelihood approach. First, a support vector machine classifier with a set of seven isotopic ratios finds the reactor of origin and a set of regression models trained using gaussian process predicts the burnup with a different set of seven isotopes. Finally, TSI is calculated analytically using decay equations. Thus far, the new methodology is capable of attributing pure Pu samples and has been validated using experimental data. The next step will to be augment the classifier training data set with spoofed Pu data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Separation of thorium, uranium, and rare earths from a strip solution generated from coarse coal refuse

This work systematically evaluated the separation performance of thorium, uranium, and rare earths when coal and coal byproduct were utilized as a non-traditional feedstock for rare earth production. A synthetic solution was prepared following the elemental composition of a strip solution generated from coarse coal refuse and used throughout the study. Various separation techniques including selective precipitation, solvent extraction, and a modified experimental protocol incorporating the two were applied to extract rare earths while minimizing the non-selective recovery of thorium and uranium into the product stream. Test results indicate that selective precipitation was effective for the removal of thorium at a pH value approaching 5 while solvent extraction preferentially removed uranium from rare earths. Therefore, a modified experimental protocol consisting of both selective precipitation and solvent extraction was subsequently developed and implemented. The findings following the modified experimental protocol indicate that extractant concentration, solvent extraction feed pH, and organic to aqueous (O/A) ratio all played a significant role in the removal of uranium. Among all the tests conducted, the best separation performance was achieved using a one-stage precipitation at a pH value of 4.8, 50 v% TBP, solvent extraction feed pH of 3.5, and an O/A ratio of 3, which corresponded to an overall rare earth, thorium, and uranium recovery of 79.6%, 0%, and 3.1%, respectively.

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2022

Initiating a Roadmap for Solar Fuels R&D: Imagining Beyond Thermochemical Cycles

Sandia National Laboratories in collaboration with the National Renewable Energy Laboratory outline a framework for developing a solar fuels roadmap based on novel concepts for hybridizing gas-splitting thermochemical cycle s with high-temperature electro chemical steps. We call this concept SoHyTEC, a Solar Hybrid Thermochemical-Electrochemical Cycle. The strategy focuses on transforming purely thermochemical cycles that split water (H 2 O) and carbon dioxide (CO 2 ) to produce hydrogen (H 2 ) and carbon monoxide (CO) , respectively, the fundamental chemical building blocks for diverse fuels and chemicals , by substituting thermochemical reactions with high-temperature electrochemical steps. By invoking high-temperature electrochemistry, the energy required to complete the gas-splitting cycle is divided into a thermal component (process temperature) and an electrical component (applied voltage). These components, sourced from solar energy, are independently variable knobs to maximize overall process efficiency. Furthermore, a small applied voltage can reduce cycle process temperature by hundreds of degrees , opening the door to cost-effective solar concentrators and practical receiver/reactor de signs. Using the SoHyTEC concept as a backdrop, we outline a framework that advocates developing methods for automating information gathering, critically evaluating thermochemical cycles for adapting into SoHyTEC, establishing requirements based on thermodynamic analysis, and developing a model-based approach to benchmarking a SoHyTEC system against a baseline concentrating solar thermal integrated electrolysis plant. We feel these framework elements are a necessary precursor to creating a robust and adaptive technology development roadmap for producing solar fuels using SoHyTEC. In one example, we introduce high-temperature electrochemistry as a method to manipulate a fully stoichiometric two-step metal oxide cycle that circumvents costly separation processes and ultra-high cycle temperatures. We also identify and group water-splitting chemistries that are conceptually amenable to hybridization.

14 SOLAR ENERGY↗

DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Modeling of ATR Fuel in DOE Standard Canisters with Helium Backfill

One pathway for road-ready and final disposition packaging configurations for the aluminum-clad spent nuclear fuel (ANSF) fuel is storage within helium backfilled sealed Department of Energy (DOE) standard canisters. The typical packaging configuration for the 15-foot DOE standard canisters places 10 advance test reactor (ATR) elements a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water bound to the surface. A 50-year CFD model of the DOE canister packaged with fuel was developed to provide a temperature profile for coupled chemical modeling of the conditions within the canister. The results of this modeling include results at fully saturated and fully dried fuel cladding conditions. In the associated experimental work, radiolysis experiments tests were completed in a helium environment, and G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. That reaction was coupled with the thermal profiles and gas-phase reactions to develop a 50-year model of the conditions within a sealed DOE canister with ATR fuel. For a nominal scenario of stored ATR fuel, after 50 years the model results give a 1.36 atm total pressure, 7% mole percent hydrogen, for the upper decay heat, 1.51 atm total pressure, 16% mole percent hydrogen, and for upper decay heat with undried fuel 2.6 atm total pressure, 15% mole percent hydrogen. No case modeled yields significant oxygen, and for the lower decay heat case that is modeled, hydrogen concentrations are under the 4% flammability limit after 50 years of storage. The modeled pressures for all cases modeled are below the pressure limit for the DOE standard sealed canister.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Modeling of the AGR 3/4 PIE Experiments

TRISO Fuel Accident Scenarios Transport of Fission Products (FP) through modern nuclear-grade graphites is not well established The AGR 3/4 experimental results need to be modeled to extract FP transport characteristics A finite element model using graphite diffusivities from IAEA TECDOC-978 is compared against measured results

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Recent Progress in U-10Mo Mechanical and Thermophysical Property Characterization

Results are presented on the mechanical and thermophysical property characterization of U-10Mo as part of the MP-1 qualification campaign. The thermal diffusivity, specific heat and thermal conductivity of as fabricated U-10Mo fuel foils and plates were measured using the thermal conductivity microscope (TCM), differential scanning calorimeter (DSC) and laser flash analyzer (LFA) all of which located in a shielded glove box – thermal property cell (TPC). A novel inverse method based on finite element analysis was developed to evaluate the thermal diffusivity of composite/layered materials via the LFA technique such as the 5-lyered U-10Mo monolithic fuel mini-plates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Recent Progress in U-10Mo Mechanical and Thermophysical Property Characterization -- Poster

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

A review of rare earth elements and yttrium in coal ash: Content, modes of occurrences, combustion behavior, and extraction methods

Rare earth elements and yttrium (REY) have attracted considerable attention over the last decade because of their vital roles in clean energy, consumer product, national defense and security applications, among other uses. Due to the retention of REY during coal burning, coal combustion ash is considered as potential alternative sources for REY. Understanding the content, speciation, retention and/or transformation behavior of REY during coal combustion not only expands our knowledge of the combustion behavior of the trace elements in coal, but also provides basis for modeling REY partitioning during coal combustion and for developing economically viable REY recovery technologies. This review makes a critical summary of recent progress in the study of REY in coal ash. The contents and the extraction potentials of REY in coal ash derived from 15 major coal-producing countries worldwide were summarized and evaluated. Various analytical methods for determining REY bulk contents and speciation, together with the solid sample pretreatment, analytical accuracy and precision, advantages and disadvantages were summarized and compared. Modern analytical approaches combined indirect methods (e.g., sequential extraction) shed light on the physical distribution, mineralogy, and the chemical state of REY in coal ash. Three types of REY occurrences in coal ash, including Si-Al glassy association, discrete minerals or compounds, and organic association (bound with unburned carbon) were defined in the review. The glassy association can be further divided into REY minerals closely bound to glass phases and dispersed throughout the glassy structure. REY partitioning in various emission streams, the size distribution, and their enrichment behavior in coal ash were discussed. Additionally, thermal behavior and transformation of various REY forms in coal during combustion process, including organic-associated REY, REY phosphates, REY carbonates, clay-bound REY and among others were summarized. Two possible retention mechanisms of REY by aluminosilicate glass at boiler temperature were proposed: the incorporation of the individual REY phases into the glass as inclusions and the diffusion of REY phases throughout Si-Al glass structures in the melting process. Feed coal mineral types, mineral-mineral associations, boiler conditions, and other factors control the retention process. After coal combustion, the speciation of REY in fly ash may be modified by the reactions of REY phases with flue gas components. Further, an overview of REY transformation mechanisms during coal combustion was deeply discussed. Finally, current extraction techniques for REY recovery from coal combustion ash were introduced. Future outlooks and research problems were also identified.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

MECHANICAL SHOCK AND VIBRATION ANALYSIS OF SPENT NUCLEAR FUEL CARRIED BY THE ATLAS RAILCAR

Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.

Klymyshyn, Nicholas A.↗

DOE OSTI2022

In-situ Laser Induced Breakdown Spectroscopy Measurements in PISCES-A

The goal of the project was to develop a Laser Induced Breakdown Spectroscopy (LIBS) diagnostic that could be employed during the plasma exposure of targets in the PISCES-A linear plasma device to interrogate the surface as the plasma exposure progressed. The objectives of the measurements were twofold; first, to measure in real-time the composition of elements in the near surface ablated zone to enable a tool that could be used to validate material migration models of eroded material, and second, to measure the plasma fuel contained in the ablated surface to determine the static and dynamic inventory of gas in a target during the plasma exposure. The LIBS diagnostic was successfully deployed on PISCES-A and measured both elemental composition and gas accumulation in the near-surface ablated layer during plasma operations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Experimental and Multiscale Modeling Insights for Radiation-Driven Neptunium Redox Processes at Elevated Temperatures

Studies of the radiation-driven reactions of actinide elements at elevated temperatures are extremely limited, but important given that radiation heating of used nuclear fuel can affect the radiolytically promoted actinide redox processes occurring in hydroprocessing environments. Under these conditions, the high concentration of nitric acid present makes the reaction of actinides, such as neptunium, with nitric acid radiolysis products significant. Therefore, here we present derivation of the rate coefficients for the reaction of pentavalent neptunium with the nitrate radical at elevated temperature, resulting in Eyring and Arrhenius parameters for this important reaction. Here, we also revisit our previously published multiscale model predictions for radiation-induced neptunium redox chemistry and present an iteration which successfully operates over a wider range of nitric acid concentrations, 0.1–6.0 M.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Rapid quantitative analysis of trace elements in plutonium alloys using a handheld laser-induced breakdown spectroscopy (LIBS) device coupled with chemometrics and machine learning

Here, we present the first reported quantification of trace elements in plutonium via a portable laser-induced breakdown spectroscopy (LIBS) device and demonstrate the use of chemometric analysis to enhance the handheld device's sensitivity and precision. Quantification of trace elements such as iron and nickel in plutonium metal via LIBS is a challenging problem due to the complex nature of the plutonium optical emission spectra. While rapid analysis of plutonium alloys has been demonstrated using portable LIBS devices, such as the SciAps Z300, their detection limits for trace elements are severely constrained by their achievable pulse power and length, light collection optics, and detectors. In this paper, analytical methods are evaluated as a means to circumvent the detection constraints. Three chemometric methods often used in analytical spectroscopy are evaluated; principal component regression, partial least-squares regression, and artificial neural networks. These models are evaluated based on goodness-of-fit metrics, root mean-squared error, and their achievable limits of detection (LoDs). Partial least squares proved superior for determining content of iron and nickel in plutonium metal, yielding LoDs of 15 and 20 ppm, respectively. These results of identifying the undesirable trace elements in plutonium components are critical for applications such as fabricating radioisotope thermoelectric generators or nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Quantitative uranium elemental reconstruction using spectral x-ray radiography

X-ray radiography offers the ability to noninvasively inspect the internal structure and material composition of objects. Dual-energy radiography allows material discrimination but is limited in quantitative material measurements due to the broad nature of the interrogating x-ray spectrum. Spectral radiography detectors provide an observation of not just the number of x rays that are transmitted through a sample but also the spectrum. Here, experimental validations of a method to use spectral x-ray radiography to accurately quantify the uranium mass in a powder are presented. An accurate system response model was developed, which includes a model of the incident x-ray source spectrum and a pixel-wise detector response that describes how the transmitted x rays were observed. Further, a calibration to account for the effect of partial x-ray attenuation was developed. As a result, the method is demonstrated to be capable of estimating the uranium mass in a variety of uranium oxide powders to a bias of –0.01±0.62%.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Mechanisms of Water-Stimulated Mg 2+ Intercalation in Vanadium Oxide: Toward the Development of Hydrated Vanadium Oxide Cathodes for Mg Batteries

As lithium-ion batteries approach their theoretical limits for energy density, magnesium-ion batteries are emerging as a promising next-generation energy storage technology. However, progress in magnesium-ion battery research has been stymied by a lack of available high capacity cathode materials that can reversibly insert magnesium ions. Vanadium Oxide (V 2 O 5 ) has emerged as one of the more promising candidate cathode materials, owing to its high theoretical capacity, facile synthesis methods, and relatively high operating voltage. This review focuses on the outlook of hydrated V 2 O 5 structures as a high capacity cathode material for magnesium-ion batteries. In general, V 2 O 5 structures exhibit poor experimental capacity for magnesium-ion insertion due to sluggish magnesium-ion insertion kinetics and poor electronic conductivity. However, several decades ago, it was discovered that the addition of water to organic electrolytes significantly improves magnesium-ion insertion into V 2 O 5 . This review clarifies the various mechanisms that have been used to explain this observation, from charge shielding to proton insertion, and offers an alternative explanation that examines the possible role of structural hydroxyl groups on the V 2 O 5 surface. While the mechanism still needs to be further studied, this discovery fueled new research into V 2 O 5 electrodes that incorporate water directly as a structural element. The most promising of these hydrated V 2 O 5 materials, many of which incorporate conductive additives, nanostructured architectures, and thin film morphologies, are discussed. Ultimately, however, these hydrated V 2 O 5 structures still face a significant barrier to potential applications in magnesium-ion batteries. During full cell electrochemical cycling, these hydrated structures tend to leach water into the electrolyte and passivate the surface of the magnesium anode, leading to poor cycle life and low capacity retention. Recently, some promising strides have been made to remedy this problem, including the use of artificial solid electrolyte interphase layers as an anode protection scheme, but a call to action for more anode protection strategies that are compatible with trace water and magnesium metal is required.

25 ENERGY STORAGE↗

DOE OSTI2023

Management of Alkali and Alkaline Earth Fission Products in Used Pyroprocessing Salt

Pyroprocessing of spent nuclear fuel (SNF) involves dissolving metallic fuel into a molten salt electrolyte (typically eutectic LiCl-KCl) and then preferentially depositing actinides onto inert cathodes. Subsequent operations include drawdown of residual actinides and lanthanides from the electrolyte prior to re-using the salt. The recovered actinides are recycled and the recovered lanthanides are disposed as waste. Alkali and alkaline earth metal fission products in the fuel, such as Cs, Sr and Ba, dissolve into the salt during electrorefining. The concentrations of these elements buildup over time in the molten salt electrolyte, which may change the freezing point. The radioactive decay of 137 Cs and 90 Sr (half life 30 and 29 years) generates significant heat and produces strong ionizing radiation fields (β and γ). The increasing heat load and radioactivity as these elements build up in the molten salt requires frequent replacement and disposal of the electrolyte salt. Alternatively, the salt can be treated to remove these and other elements and then recycled to the electrorefiner. An effective strategy to manage these alkali and alkaline earth metal fission products in the molten salt electrolyte would increase the efficiency of pyroprocessing and decrease the volume of salt waste requiring disposal. Alkali and alkaline earth metal fission products are extremely stable in molten salt as chlorides--even more stable than the LiCl-KCl eutectic base salt--making them challenging to remove. They are not removed during drawdown operations to recover residual actinides and lanthanides and a separate operation is required to sufficiently purify the salt for reuse. This work is focused on selecting a method for separating Cs, Sr and Ba from the salt recovered from the lanthanide drawdown operation prior to recycling the cleaned salt back to the electrorefiner. Not addressed in this work is the management of the waste stream produced by the separation. This report summarizes the issues to be addressed when developing removal strategies for cesium, strontium, and barium and reviews existing methods to identify suitable methods and any technological gaps in their application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Delayed onset of discontinuous precipitation-based phase transformation in U10Mo alloys doped with Silicon

A uranium-10 wt% molybdenum (U-10Mo) alloy is one of the primary candidates for metallic fuels that would use low-enriched uranium in place of highly enriched uranium, to support nuclear nonproliferation efforts. Optimal performance of a U-based metallic nuclear fuel can be achieved by retaining the high-temperature, body-centered cubic (bcc) allotrope (γ-U) at room temperature, which can be accomplished in the U-10Mo alloy. However, presence of minor alloying elements can influence the final constitution of room-temperature phases in the U10Mo alloy, specifically, formation of α-U phase which results in anisotropic behavior of the fuel in reactor. Further, through a detailed transmission electron microscopy analysis, the present study reports the constituent phases that are present in a U10Mo alloy containing ~0.1 wt% Si after it is subjected to homogenization heat treatment and thermomechanical processing. For comparison, results from an undoped U10Mo alloy are also included. The experimental results reveal that γ-UMo solid solution is the major phase in a hot-rolled, Si-doped U10Mo alloy metallic fuel foil, along with U 2 MoSi 2 C, UC, and U 2 Mo, after isothermal annealing at 460 °C for 10h. In contrast, after the same heat treatment, the undoped U10Mo alloy metallic fuel had formed a noticeable amount of α-U along prior γ-UMo grain boundaries through discontinuous precipitation (DP, area fraction: ~27.9%) with characteristic lamellar morphology, together with γ-UMo, UC, and U 2 Mo. This result indicates that doping with Si could mitigate the DP reaction in U10Mo alloy and prevent formation of undesirable α-U. This work sheds light on optimizing Si-doping–dominated microstructure in U10Mo fuels and facilitates designing and tuning of microstructures of U10Mo alloys for tailoring the final designed performance of the fuel under irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗