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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Microstructure study of U–35 wt.% Zr alloy after quick annealing at 650 °C

Uranium-35wt.%zirconium (U-35wt.%Zr) alloy was annealed for 1 hour and 24 hours at 650 ºC and characterized to understand the early stage microstructure evolution. Dendritic microstructure with fine (~300 nm in length) α-U precipitates clustered between dendrite branches were observed in the 1-hour annealed sample. After 24-hour annealing 650 ºC, the α-U precipitates coarsened, and the dendritic microstructure disappeared due to microstructure homogenization. Furthermore, microchemical homogenization observed with Energy-Dispersive X-ray Spectroscopy analysis suggests that α-U precipitates are approaching thermodynamic equilibrium in the 24-hour annealed sample. The findings from this study have potential impacts on the manufacturing and computer modeling of metallic nuclear fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling of High-Temperature Corrosion of Zirconium Alloys Using the eXtended Finite Element Method (X-FEM)

Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the a and ß phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized a, and prior ß phase layer growth kinetics under isothermal exposure at high temperature (1000–1500°C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. Furthermore, this implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effect of Heat Treatment on Microstructure and Mechanical Property of 316L Stainless Steel Produced by Laser Powder Bed Fusion

The advanced non-light water reactor designs (Gen IV reactors), including molten salt/ very high temperature/ sodium-cooled and lead-cooled fast reactors, typically operate at higher temperatures and more extreme radiation conditions than light water reactors. An intrinsic part of the deployment and progress of Gen IV reactor designs is selecting the most suitable structural material for a specific application. Additive manufacturing (AM), a fairly new process of making physical, three-dimensional objects from a computer design file, is going to completely change the way of design, build and certify nuclear systems. It offers a range of opportunities to produce complex geometries from existing materials, offers new routes for processing of previously difficult to process materials, allows for design of new high-performance materials, and finally facilitates hybridization of dissimilar materials. This emerging technology has successfully produced cars, wind turbine blade molds and even live cells. It could also open up big opportunities for the nuclear industry to quickly deploy technologies at a fraction of the cost. So far, AM techniques have been preliminarily applied in the field of nuclear reactors, including the classical parts such as the pressure vessel of a small reactor with 508-III steel, the bottom nozzle of a fuel assembly with 304L steel, the fuel cladding with zirconium alloy and the integrated impeller of a pump and the multi-channel valve body with 316L steel [6,7]. The AM applications for operating nuclear reactors started in auxiliary plant components and have slowly migrated to metallic reactors and core components, but many of these are not safety critical components. Although many parts used for nuclear reactors have been fabricated by AM techniques, practical applications in engineering are still a long way off due to the uncertainty factors focused on the processing, material properties, analysis methods and application standards, which feeds the safety and life-cycle of the nuclear reactor. Due to rapid, repeated heating and cooling during production, a high dislocation density was present in the AM material. This microstructure feature is unstable at elevated temperature while high temperature is one of the typical operation environments for nuclear reactors. Thus, it is important to understand the thermal effect on the microstructure of AM material. The objectives of this study are to investigate the effect of heat treatment on the microstructure and mechanical properties of 316L stainless steel produced by laser powder bed fusion additive manufacturing, and to determine an appropriate heat treatment practice that will be applied to the lightweight AM lattice-structured material with the same chemistry. The heat treatment study consisted of annealing the samples at a temperature range of 800 to 1200 oC with a 50 oC increment for different times (1-24 hours), followed by vacuum or air cooling. Microstructural characterization was carried out by Scanning Electron Microscope (SEM). Grain size and crystallographic orientation were investigated by Electron Backscatter Diffraction (EBSD). Vickers hardness tests with a 0.5 kg load were employed to determine the hardness of samples after different heat treatments. After heat treatment, the random crystallographic orientation was preserved, and the volume fraction of high-angle grain boundaries (grain boundary misorientation =15 oC) remained the same. The dislocation density decreased with annealing temperature due to recovery. The fine subgrain structures in the as-printed specimen were quite stable up to 1200 oC. Minimal recrystallization was observed up to 1200 oC. Recrystallization initiated only after 8.5 hours at 1200 oC. The SEM images did not show obvious dependence of microstructure on cooling rate. The hardness of the specimens decreased with increasing annealing temperature as a result of the decrease in dislocation density. It is interesting to note that the AM material showed very similar hardness to the wrought material when annealing at similar temperature, although the microstructures are very different. Annealing at 1050 oC for 1 hour followed by air cooling was selected as the heat treatment procedure for the lattice designed lightweight AM 316L material.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Phosphate‐functionalized Zirconium Metal–Organic Frameworks for Enhancing Lithium–Sulfur Battery Cycling

Abstract Lithium–sulfur batteries are promising candidates for next‐generation energy storage devices due to their outstanding theoretical energy density. However, they suffer from low sulfur utilization and poor cyclability, greatly limiting their practical implementation. Herein, we adopted a phosphate‐functionalized zirconium metal–organic framework (Zr‐MOF) as a sulfur host. With their porous structure, remarkable electrochemical stability, and synthetic versatility, Zr‐MOFs present great potential in preventing soluble polysulfides from leaching. Phosphate groups were introduced to the framework post‐synthetically since they have shown a strong affinity towards lithium polysulfides and an ability to facilitate Li ion transport. The successful incorporation of phosphate in MOF‐808 was demonstrated by a series of techniques including infrared spectroscopy, solid‐state nuclear magnetic resonance spectroscopy, and X‐ray pair distribution function analysis. When employed in batteries, phosphate‐functionalized Zr‐MOF (MOF‐808‐PO4) exhibits significantly enhanced sulfur utilization and ion diffusion compared to the parent framework, leading to higher capacity and rate capability. The improved capacity retention and inhibited self‐discharge rate also demonstrate effective polysulfide encapsulation utilizing MOF‐808‐PO4. Furthermore, we explored their potential towards high‐density batteries by examining the cycling performance at various sulfur loadings. Our approach to correlate structure with function using hybrid inorganic–organic materials offers new chemical design strategies for advancing battery materials.

Liu, Bingqian↗
DOE OSTI2021

Zirconium and hafnium polyhedral oligosilsesquioxane complexes – green homogeneous catalysts in the formation of bio-derived ethers via a MPV/etherification reaction cascade

The polyhedral oligosilsesquioxane complexes, {[(isobutyl) 7 Si 7 O 12 ]ZrOPr i ·(HOPr i )}2 (I), {[(cyclohexyl) 7 Si 7 O 12 ]ZrOPr i ·(HOPr i )} 2 (II), {[(isobutyl) 7 Si 7 O 12 ]HfOPr i ·(HOPr i )} 2 (III) and {[(cyclohexyl) 7 Si 7 O 12 ]HfOPr i ·(HOPr i )} 2 (IV), were synthesized in good yields from the reactions of M(OPr i ) 4 (M = Zr, Hf) with R-POSS(OH) 3 (R = isobutyl, cyclohexyl), resp. I–IV were characterized by 1 H, 13 C and 29 Si NMR spectroscopy and their dimeric solid-state structures were confirmed by X-ray analysis. I–IV catalyze the reductive etherification of 2-hydroxy- and 4-hydroxy and 2-methoxy and 4-methoxybenzaldehyde and vanillin to their respective isopropyl ethers in isopropanol as a “green” solvent and reagent. I–IV are durable and robust homogeneous catalysts operating at temperatures of 100–160 °C for days without significant loss of catalytic activity. Likewise, I–IV selectively catalyze the conversion of 5-hydroxymethylfurfural (HMF) into 2,5-bis(isopropoxymethyl)furane (BPMF), a potentially high-performance fuel additive. Similar results were achieved by using a combination of M(OPr i ) 4 and ligand R-POSS(OH) 3 as a catalyst system demonstrating the potential of this “in situ” approach for applications in biomass transformations. A tentative reaction mechanism for the reductive etherification of aldehydes catalysed by I–IV is proposed.

09 BIOMASS FUELS↗
DOE OSTI2023

Designing Nuclear Fuels with a Multi-Principal Element Alloying Approach

Previous research has shown that multi-principal element alloys (MPEAs) using chromium, molybdenum, niobium, tantalum, titanium, vanadium, and zirconium can form stable body-centered-cubic (BCC) structures across a large temperature region (25°C to 1000°C). This is the same crystal structure as γ-uranium (U), which has shown desirable thermal and irradiation behavior in previous alloy fuel research. It is hypothesized then that the MPEA alloying approach can be used to produce a stable BCC uranium-bearing alloy and to retain its stability throughout anticipated operating regimes of power-producing reactors. Candidate elements were assessed using Monte Carlo N-Particle (MCNP) analysis to determine uranium densities necessary to make the alloy an economically viable fuel compared to conventional fuel forms. Following neutronic considerations, materials property databases and empirical predictors were used to determine the compositions with a high potential to form a BCC solid solution alloy. The final four alloys were MoNbTaU 2 , MoNbTiU 2 , NbTaTiU 2 , and NbTaVU 2 , which were cast using arc melting of raw elemental foils and chunks. Characterization of the fabricated alloys included scanning electron microscopy, X-ray diffraction, energy dispersive X-ray spectroscopy, and transmission electron microscopy. The results showed a two-phase system with a U-rich matrix phase surrounding the refractory precipitates. The U phase was found to contain varying concentrations of the alloying elements and was a BCC γ-U phase. These results warrant further research to identify ideal compositions for use as an advanced alloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Size effects in 3D–printed polymer–derived, zirconium diboride–reinforced ceramic composites

Preceramic polymers are of interest for use in many manufacturing techniques such as injection molding, ceramic fiber infiltration, and additive manufacturing. However, off-gassing of low molecular weight oligomers occurs when these polymers cure, potentially leading to porosity in the cured part. Here to study how porosity and strength are affected by the size of the printed part, and the presence of a high surface area nano-scale filler, polycarbosilane (PCS) microrods of varying diameter were fabricated via direct ink writing (DIW), an additive manufacturing technique, with two ink formulations containing either zirconium diboride (ZrB 2 ) alone, or ZrB 2 and fumed alumina (FA). Sets of microrods were printed in a range of sizes by using print nozzles of 450, 634, 979, 1 346, or 1 702 μm in diameter, which were thermally cured, pyrolyzed to form ceramic composite microrods, and tested in 3-pt flexure. Porosity increased with increasing diameter, while failure strength decreased. For a given nozzle size, the microrods containing FA displayed lower porosity and higher strength (up to ~500 MPa) compared to the microrods containing only ZrB 2 . Weibull strength analysis was performed on each group of microrods and shows that the addition of FA increased Weibull modulus from 4.63 ± 1.56 to 9.35 ± 0.601. In conjunction with optical microscopy, this analysis indicates two distinct flaw populations in the printed materials, porosity which arises during the curing step and cracking which arises during pyrolysis of the larger specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Step-loaded creep testing of Zircaloy-4 cladding at higher temperatures in the α-phase

A refined understanding of zirconium-based cladding thermomechanical performance during rapid transients is essential for enhancing the safety and operation of light-water reactors. Traditional models for zirconium alloys under accident conditions generally assume that creep dominates fuel cladding performance. Here, these historic models have largely remained unchanged and serve as the basis for safety criteria development. As the U.S. nuclear industry pursues higher burnup levels, the increased release of fission gases during transients raises the risk of cladding failure in the low-temperature hcp α-phase, making the fidelity of these models of greater importance. Creep testing was conducted from 550–700°C with 25–120 MPa applied hoop stresses to investigate Zircaloy-4 deformation at accident-relevant temperatures in the α-phase. Step-loading was employed to capture creep behavior across a wide stress range from a single sample. The stress-strain rate data at higher temperatures (650 and 700°C) were well-described by isotropic versions of the Erbacher and Kaddour models, while the lower temperature data (550 and 600°C) were underpredicted by both anisotropic and isotropic model variants. Greater strain rates during the initial loading step at 650 and 700°C were attributed to recrystallization and grain growth of sub-micron crystallites. Yet, texture analysis revealed the basal split texture remained after testing. These observations produced results suggesting Zircaloy-4 claddings experience higher creep rates across the α-phase than previously thought, possibly related to dynamic anisotropy due to temperature dependent activation of deformation mechanisms, effects of biaxial loading, and variation in material condition between the current testing used in previous model development.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Rotational Dynamics of an Amphidynamic Zirconium Metal–Organic Framework Determined by Dielectric Spectroscopy

A zirconium metal–organic framework with a difluorophenylene rotator bearing a permanent electric dipole of ~3.2 D was synthesized, and its rotational motion was analyzed by temperature- and frequency-dependent broadband dielectric spectroscopy. While solid-state NMR confirms fast rotation qualitatively, the dissipation factors measured between 113 and 153 K suggested an activation energy E a = 2.6 kcal/mol, but deviations from a single Debye relaxation suggested a dynamic process that cannot be accounted for by a well-defined potential with a single activation barrier. The dynamic heterogeneity of the dipolar rotor was confirmed by analysis in terms of a Cole–Cole relaxation, which suggested a mean barrier of ~1.9 kcal/mol, with a heterogeneity that decreases as temperature increases. Based on the single-crystal structure, we propose that the kinetic heterogeneity results from a temperature-dependent potential where rotation motion is mediated by the escape of the rotator from an energy well created by a double Ph–H···F–Ph hydrogen bond.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Rapid Polyolefin Plastic Hydrogenolysis Mediated by Single-Site Heterogeneous Electrophilic/Cationic Organo-group IV Catalysts

A homologous series of cationic electrophilic group IV metal hydrides (M = Ti, Zr, Hf) created by chemisorption of the corresponding MNp 4 precursors on highly Brønsted acidic sulfated alumina (AlS) to yield loosely coordinated surface AlS/MNp 2 (Np = neopentyl) species are systematically characterized by ICP, EXAFS/XANES, DRIFTS, and solid-state high-resolution multidimensional NMR spectroscopy (SSNMR), as well as by energy span DFT computation. With effective stirring, these complexes readily undergo reaction with H 2 to yield AlS/M(alkyl)H species, which are highly active for the hydrogenolysis of diverse commercial polyethylenes, α-olefin-ethylene copolymers, isotactic polypropylene, and postconsumer polyolefins including high-density polyethylenes, yielding medium and small linear and branched hydrocarbons at turnover frequencies as high as 36,300 h –1 at 200 °C/17 atm H 2 for M = Zr. For given polyolefin and reaction conditions, turnover frequencies scale approximately as M = Zr > Hf > Ti, whereas catalyst thermal stability scales approximately as M = Hf ≈ Zr > Ti, and these trends are qualitatively understandable from the DFT analysis. These catalytic results reveal that the AlS/Hf(R)H-mediated hydrogenolysis favors wax-like and liquid products, whereas the AlS/Zr(R)H-mediated hydrogenolysis can be tuned between gases and liquids. Finally, DFT analysis identifies β-alkyl elimination as the turnover-limiting C–C scission process, which is particularly facile in these cationic d 0 complexes but not so in the neutrally charged analogues.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide

To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.

Uranium Dioxide, Gadolinium Oxide, Thermal Air Oxi↗
DOE OSTI2025

Development of Innovative Non-Concentrated Alloy (NCA) Fuel Cladding for Advanced Nuclear Reactors

As the nuclear energy sector advances toward next-generation reactors, the need for high-performance fuel cladding materials has become increasingly urgent. Traditional alloys like zirconium and stainless steel are reaching their performance limits under higher temperatures, more corrosive coolants, and extended irradiation. This report presents the development of a new class of fuel cladding materials based on Non-Concentrated Alloys (NCAs) ? multi-element systems designed to deliver enhanced mechanical strength, corrosion resistance, and radiation tolerance. Through a combination of computational modeling (Computer Coupling of Phase Diagrams and Thermochemistry: CALPHAD), simulation-guided alloy selection, and experimental fabrication via arc melting and spark plasma sintering (SPS), three strategic alloy design paths were explored: (1) FeCrAl-based NCAs, (2) refractory-lean neutron-efficient alloys, and (3) equimolar high-entropy compositions. Microstructural analysis confirmed the formation of stable body-centered cubic BCC_A2 phases, while mechanical testing demonstrated hardness values significantly exceeding those of conventional cladding materials. The results highlight the tunability of NCA systems and their potential for balancing strength and ductility ? a critical consideration for in-reactor performance. Looking forward, future work will focus on thermomechanical optimization, CALPHAD refinement, and benchmarking against industry standards to enable scalable deployment. This work not only advances the science of nuclear materials but also supports broader goals in nuclear safety, performance, and nuclear energy innovation.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Antioxidant-induced transformations of a metal-acid hydrocracking catalyst in the deconstruction of polyethylene waste

Plastics are ideal for numerous applications, as driven by the development of complex formulations containing various additives to improve performance and processability. Unfortunately, chemical valorization strategies (e.g., catalytic deconstruction) often can be challenged by the presence of small-molecule additives, and quantification of the impact of these molecular constituents on upcycling processes remains elusive. Here, this dearth of information restricts catalyst design efforts to combine the robustness and performance necessary to improve plastics circularity. In this communication, we describe a systematic study of common plastics additives—phenolic antioxidants, and we quantify the relationship between additive content and deconstruction yields of high-density polyethylene (HDPE) over a platinum on tungstated zirconium (Pt/WO 3 /ZrO 2 ) hydrocracking catalyst. In the simplest case of a base (antioxidant- and slip agent-containing) HDPE resin versus a pure (additive-stripped) HDPE polymer, a two-fold decrease in the yield of gas and liquid products is noted for base HDPE resin. Furthermore, both antioxidant chemistry and concentration strongly impact conversion and individual product selectivities. Using infrared spectroscopy, we determine that antioxidants change the effective ratio of metal to acid sites (i.e., metal-acid balance) through reactions of phenols and/or other functional groups (e.g., acids, esters) with catalyst active sites. Overall, this work demonstrates the impact of one common set of additives on plastics deconstruction, and the analysis herein provides a blueprint for quantitatively assessing the effects of additives on plastics deconstruction processes and for evaluating the development of more robust catalytic strategies or more tolerable additive formulations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

An Investigation of Liquefaction in Irradiated TRIGA Fuel Exposed to Relatively High Temperatures

This report presents the findings of an investigation into high temperature fuel cladding chemical interactions (FCCI) in Training, Research, Isotopes, General Atomics (TRIGA) fuel rods. A TRIGA fuel-rod core or meat is principally composed of uranium (U) particles dispersed in a zirconium hydride (Zr H) matrix. The fuel is clad in sealed 304SS or Incoloy 800 tubes. At high temperatures, the fuel will interact with the cladding, resulting in FCCI. To investigate the FCCI (in this case, liquefaction), irradiated TRIGA fuel rods were exposed to relatively high temperatures in furnace tests. The tests were performed at different temperatures using segments from an irradiated TRIGA fuel rod at the Hot Fuel Examination Facility located at Idaho National Laboratory (INL). The data from this study is important for developing a better understanding of the performance of U-ZrH fuel type during transient events (e.g., a Loss of Coolant Accident (LOCA) event) that may result in high temperatures while operating a TRIGA reactor. Furnace tests were run for 6 hours at 730, 800, 900, 950, and 1,000°C. One test was run for 12 hours at 950°C. Post-test microstructural characterization was performed on the heat-treated samples using optical metallography to look for evidence of liquefaction. Additionally, microstructural characterization was performed using scanning electron microscopy to investigate the microstructure of the as-irradiated fuel before heat treatment and the microstructural changes that occurred after heat treating for the 730 and 950°C heat treatments for 6 hours. Results of the analysis were compared to those produced during diffusion studies performed using unirradiated TRIGA fuel meat and type 304 stainless steel. A gap is typically present in as-irradiated fuel that limits any possibility for chemical interaction between the fuel and cladding. Optical metallography (OM) was used to identify areas in the specimens heat treated at 950 and 1000°C where fuel and cladding contact was satisfactory for the FCCI to transpire. The bulk of the interaction zones that form due to FCCI developed in the fuel meat and not as much in the cladding. Even in the sample tested at 1,000°C, no evidence was found that gross fuel meat or cladding melting had occurred. Only limited porosity was found in the fuel meat that possibly could have been due to melting of a particular precipitate phase or phases that had formed during FCCI. These results agreed with those reported for the unirradiated diffusion couple studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A critical review of the history of fabricating monolithic U-Mo fuel plates

The fabrication of monolithic uranium-molybdenum alloy fuels, specifically those developed for high-performance research and test reactors, began in the early 2000s. The primary fuel form consists of uranium alloyed with 10-wt% molybdenum in a thin foil coated with zirconium and encapsulated in aluminum-6061 cladding. Over the years, the process has evolved with different types of casting, heat treatments, rolling schedules, and cladding applications. This review examines the history of these fabrication processes and their impact on microstructure and fuel-swelling performance. Even though various fabrication methods were used, we found little correlation between fabrication variation and fuel swelling. This insensitive relationship between fabrication variation and fuel swelling is primarily due to inhomogeneous microstructures that formed during casting and grain refinement that occurred during rolling. We conclude that the fabrication processes we examined produced similar microstructures, indicating that the fuel microstructure is somewhat insensitive to the fabrication parameters evaluated. However, the relatively small amount of historic data, such as those for grain sizes, limited this analysis. More recently fabricated materials, such as those from ongoing irradiation experiment, Mini-Plate-1 and Mini-Plate-2, were also excluded from this analysis and are intended to be reviewed separately. The findings, that fuel microstructure is somewhat insensitive to the fabrication parameters, do not imply that any fabrication method is acceptable, given the uncertainties in data and fuel-swelling observations. For example, only arc melting and vacuum induction melting casting processes were previously explored in the historic fabrication efforts. In conclusion, the findings should not be extrapolated to other casting processes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

High Temperature High Vacuum Mechanical Property Assessment of Zirconium Nuclear Fuel Cladding

This report presents the mechanical characterization of a specific Zry-4 cladding batch serving as the foundation for a diverse range of fuel performance research at Oak Ridge National Laboratory (ORNL). This effort supports research needs for the U.S. Department of Energy (DOE), particularly regarding evaluating accident tolerant fuel (ATF) cladding coating concepts, expanding understanding of cladding response to loss-of-coolant accidents (LOCA) transients, refining post-critical heat flux (CHF) limits (t@T), and upcoming irradiation campaigns. The central objective was to define the baseline performance of the substrate Zircaloy-4 (Zry-4) material leveraged across ORNL Advanced Fuel Campaign (AFC) efforts through controlled high-temperature vacuum tensile testing. This work begins to address gaps in existing models where implementation based on nominal heat-treatment labels, such as stress relief annealed (SRA), often fail to capture the interplay of recovery, recrystallization, and grain growth. To quantify this, data was benchmarked against the Pacific Northwest National Laboratory (PNNL) stress strain model to determine where this material falls in comparison to assumed values for materials in the same heat treatment regime. Analysis of the tensile data revealed that this specific SRA batch exhibits a transitional microstructural state best described by an effective cold-work (CW) parameter of 0.09, diverging from the previous estimation of 0.5 for SRA materials. Additionally, comparative testing of Cr coated specimens demonstrated no distinct difference in axial strength relative to the bare substrate. This suggests that the strengthening benefits of Cr coatings observed in burst scenarios are driven by residual stress mechanisms acting solely in the hoop direction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Thermal Activation of Zirconium(IV) Acetylacetonate Catalysts to Enhance Polyurethane Synthesis and Reprocessing

Carbamate formation and exchange catalysts enable efficient polyurethane (PU) manufacturing, as well as emerging recycling and reprocessing methods for PU thermosets. Zirconium β-diketonate complexes, such as Zr acetylacetonate [Zr(acac) 4 ], are effective alternatives to toxic organotin catalysts that have been used for PU reprocessing. Here, we report that Zr(acac) 4 undergoes a thermally activated process in the PU network during reprocessing that transforms it into a more active carbamate exchange catalyst. This process is associated with the irreversible loss of acetylacetonate ligands and is not observed for the more sterically hindered Zr 2,2,6,6-tetramethyl-3,5-heptanedione [Zr(tmhd) 4 ] complex. Crossover experiments between PU thermoplastics indicated enhanced carbamate exchange after the thermal activation of Zr(acac) 4 in the presence of one of the PUs, whereas a sample of Zr(acac) 4 activated in the absence of the PU had no catalytic activity. Thermal gravimetric analysis suggested that this process is associated with the loss of one protonated acac ligand. Stress relaxation analysis of PU thermosets indicated a distinct change in the characteristic relaxation time associated with the thermal activation of Zr(acac) 4 at temperatures above 140 °C; no such change was observed for samples reprocessed using Zr(tmhd) 4 . Density functional theory and molecular experiments suggest that irreversible ligand exchange of acac with alkoxide or carbamate reduces the activation energy for urethane formation and reversion. Furthermore, the Zr(acac) 4 catalyst activated in the presence of a PU’s polyol precursor provided more porous and less dense PU foams compared to those made using the unactivated Zr(acac) 4 catalyst. Furthermore, these findings are important for developing improved PU synthesis and recycling processes. Thermally activating a catalyst during reprocessing may provide more nuanced control of the in-use and reprocessing characteristics of PU thermosets.

Alcohols↗
DOE OSTI2025

Device for steam cladding oxidation testing at TREAT

To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm −2 s −1 can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.

Oxidation↗