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At least 199 records · Page 11

U 3 Si 2 and UO 2 composites densified by spark plasma sintering for accident-tolerant fuels

This work reports the synthesis and characterization of the U 3 Si 2 and UO 2 composites sintered by spark plasma sintering (SPS) with controlled microstructures for accident-tolerant fuel application. The U 3 Si 2 and UO 2 composites with different silicide and oxide ratios were synthesized by SPS at temperatures from 1000 to 1300 °C for 5 minutes. The microstructure and phase composition of the SPS densified composite fuels were characterized with scanning electron microscopy, X-ray diffraction (XRD), and energy dispersed spectroscopy (EDS). A systematic study of the thermal and mechanical properties was conducted using microhardness testing and laser flash apparatus, along with oxidation resistance measurements using thermogravimetric analysis (TGA). The results show that the synthesis of composite fuels can be achieved with a 90% theoretical density (TD) at 1000 °C and over 95% TD when sintered at 1300 °C. XRD and EDS results confirmed that the dominant phases in the composites are U 3 Si 2 and UO 2 . Improved physical density generally leads to improved hardness, fracture toughness, thermal diffusivity, and onset temperature during the oxidation process. U 3 Si 2 was found to play a dominant role in determining the mechanical and oxidation properties of the composite fuels, whereas UO 2 had a more important impact on controlling the thermal diffusivity of the composites. The composite with 50 wt% UO 2 sintered at 1300 °C displayed the onset oxidation temperature of 500 °C by dynamic oxidation testing using TGA at a ramp degree of 10 °C/min. The composite also achieved a high fracture toughness of ~3.5 MPa m ½ . Finally, these results highlight the potential of composite fuel forms densified by SPS with simultaneously enhanced fissile element density, fracture toughness, thermal transport properties, and oxidation resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Cold Spray Deposition of 304L Stainless Steel to Mitigate Chloride-Induced Stress Corrosion Cracking in Canisters for Used Nuclear Fuel Storage

A feasibility study of the cold spray deposition process of 304L stainless steel on 304L stainless steel substrates as a mitigation method for chloride-induced stress corrosion cracking (CISCC) has been investigated under various substrate conditions. The study is aimed at the application of this technology to mitigate CISCC that may potentially occur in or nearthe fusion welded regions of stainless steel canisters in Dry Cask Storage System (DCSS) for used nuclear fuels. Spherical gas-atomized 304L stainless steel powder, in the size range of 25 µm to 44 µm, was used as a feedstock powder for the cold spray process. The powder was deposited on four types of 304L substrates with various surface conditions: as-polished, oxidized, cold-rolled, plates with prototypical CISCC. The effects of cold spray parameters on quality of cold spray coatings were investigated. Thickness, porosity, and phases in the as-deposited materials were evaluated using scanning electron microscopy (SEM) and X-ray diffraction (XRD), and correlated with microhardness and adhesion strength measured via micro-indentation and ASTM C633 pulling test, respectively. XRD analysis of the coatings was also conducted to examine the effects of cold spray condition on residual stress state in the coating. Detailed cross-sectional examination of coating/substrate interfaces was performed with transmission electron microscopy (TEM) equipped with energy dispersive spectroscopy (EDS). Dense and continuous coatings with good adhesion strength and hardness were produced for the various substrate conditions by adjusting cold spray parameters. The results demonstrate that cold spray stainless steel coating is a viable option to provide a physical barrier against CISCC in fusion weld regions of stainless steel in corrosive chloride salt-bearing environments.

Dry Cask Storage Systems, cold spray, 304L stainle↗

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

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

36 MATERIALS SCIENCE↗

Fabrication of yttrium hydride for high-temperature moderator application

We report given the superior thermal stability and highly attainable hydrogen density, yttrium hydride is an excellent high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. Large-scale, crack-free, bulk yttrium hydride is in high demand; however, fabrication of yttrium hydride is challenging and has not been demonstrated for nearly half century. The associated challenges are discussed herein. In response to these challenges, a hydriding system was designed and constructed at Oak Ridge National Laboratory and was used to successfully fabricate crack-free yttrium hydride in complex geometries at large scales. This was accomplished by precisely controlling the hydrogen’s partial pressure and the retort temperature, which was informed by the well-established thermodynamic properties of the binary H–Y system. Hydrogen content in as-fabricated hydride was determined by weight change, an approach which was considered reliable based on the use of ultra–high-purity yttrium, the absence of oxide phases up to levels detectable using x-ray diffraction (XRD), and the significant weight gain. Hydrogen distribution along one yttrium hydride rod was evaluated with XRD analysis on materials extracted from different locations on the rod. The results indicated a relatively homogeneous hydrogen distribution along the hydride rod, with <3% uncertainty in the fraction of the δ-phase hydride. In addition, significant efforts are being dedicated to establish a complete database summarizing the thermomechanical and physical properties of as-fabricated yttrium hydride and the irradiation response to facilitate its deployment as a high-temperature moderator in advanced nuclear reactors.

36 MATERIALS SCIENCE↗

Tensile behavior and microstructural evolution of a Fe-25Ni-20Cr austenitic stainless steel (alloy 709) from room to elevated temperatures through in-situ synchrotron X-ray diffraction characterization and transmission electron microscopy

Uniaxial tensile tests were done on Fe-25Ni-20Cr Austenitic Stainless Steel (Alloy 709) along with in-situ synchrotron measurement at different temperatures (25 degrees C, 500 degrees C, 700 degrees C and 900 degrees C). The X-Ray diffraction data was collected in-situ in order to follow the phases present as well as to derive the dislocation density as a function of strain through peak broadening analysis based on the modified Williamson-Hall method and see how trends are affected as a function of temperature. The XRD data were complemented by TEM and STEM-EDX characterization done on the post-experiment samples to observe the deformed microstructures. Finally, the paper presents a depiction of the tensile behavior of this advanced steel by interpreting the stress-strain curves and the effect of temperature through the information gained by in-situ XRD and TEM observations as well as fractography done ex-situ on the samples tested to rupture.

36 MATERIALS SCIENCE↗

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YH x , x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YH x specimens were, thus, subjected to neutron irradiations in the range of 0.1–2 displacements per yttrium atom (dpa-Y) in the temperature range of 536–878°C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YH x specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YH x exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Further, nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Microstructurally informed synchrotron x-ray analysis revealing helium defect transitions in ultrafine grained tungsten

The formation of insoluble gaseous defects in materials due to nuclear transmutation or ion implantation involves the diffusion of impurity atoms to form atomic defect clusters that coalesce into bubbles or cavities and ultimately degrade the material properties. Transmission electron microscopy (TEM) is limited in its ability to resolve sub-nanometer gas clusters whereas X-ray diffraction (XRD) provides information pertaining to local atomic changes. Here, in this study, helium (He) implanted ultrafine grained tungsten is explored through a multimodal defect characterization campaign combining TEM-informed Small Angle X-ray Scattering (SAXS) analysis, XRD lattice parameter measurements, and nanoscale He cluster quantification from a region of reciprocal space accessible via Wide Angle X-ray Scattering (WAXS). Moderate elevated temperature implantations are shown to produce high concentrations of sub-nanoscale He clusters and small, homogeneously distributed cavities, which collectively are linked to lattice expansion and further substantiated through complementary atomistic simulations. Increased implantation temperatures encourage the diffusion of these defects to the grain boundaries (GBs), leading to lattice relaxation and the growth of large GB cavities manifesting as bimodal size distributions in the SAXS analysis. Overall, our results demonstrate the utility of multimodal synchrotron X-ray analysis in bridging the gap between microscale He cavity quantification and atomic-scale defect analysis.

36 MATERIALS SCIENCE↗

Physicochemical evolution of uranium nitride kernel microstructure with varying carbon distribution for advanced TRISO fuel forms

Uranium nitride (UN) has emerged as a fuel candidate for advanced nuclear reactor concepts due to its superior uranium density, thermal conductivity, and high melting temperature. However, the fabrication route for converting UO 2 to UN is complex and difficult to standardize. Although the chemistry of this conversion process is well-studied, more insight into the physicochemical dynamics of this conversion using advanced characterization techniques can help further our understanding of this material system. This work leveraged thermogravimetric analysis (TGA), X-ray diffraction (XRD), and nondestructive 3D X-ray computed tomography (XCT) to characterize dynamic microstructural changes in the UO 2 → UCO → UN fabrication pathway for two kernels with a varying carbon distribution in the starting composition. TGA and XRD were used to quantify changes in the mass, density, and chemical composition of the two kernels, while three-dimensional image processing and segmentation of XCT data were used to quantify the volume, surface area, and spatial distribution of features within each kernel for multiple steps along the fabrication pathway. The analysis indicates distinct differences between the two kernels that are correlated to downstream conversion efficiency. In conclusion, this work is among the first to perform 3D quantification of physicochemical evolution during UN conversion, providing quantitative correlation between processing, properties, and expected fuel performance.

Nuclear fuel↗

Hot Hydrogen Exposure of U x Zr 1-x C y Nuclear Fuel: The Influence of Composition and Density

Refractory carbide nuclear fuel has been one of the most promising fuel candidates for space nuclear propulsion due to its high melting point, temperature stability, and compatibility in a hot hydrogen environment. In this study, U x Zr 1-x C y fuel was produced by means of a carbothermic reduction process in different UC compositions including 5,10, 20, and 30 at.% UC in the fuel compound. The powder feedstock was consolidated via direct current sintering with densities up to 97% of the theoretical density. The samples with different U x Zr 1-x C y compositions were exposed to hot hydrogen at 2600 K for a cumulative time of 300 min. The samples were characterized by SEM, XRD, density, and measured for mass losses. The high-density samples displayed improved performance in hot hydrogen by minimizing porous sites and reducing areas of direct contact with hydrogen gas, leading to reduced mass losses. Variations in sample density proved to induce large changes in mass loss rates, increasing them up to 90%. The compositions with higher UC content reported the largest mass losses in the study. The loss of uranium occurred primarily at the surfaces exposed to the hot hydrogen where changes in the lattice constant confirmed losses exceeding 50% of the initial UC content in higher compositions, specifically to U 0.3 Zr 0.7 C y . XRD analyses revealed the presence of UH 3 in U 0.3 Zr 0.7 C y suggesting that metallic uranium formed inside the sample as a product of carbon losses. High-density U x Zr x-1 C y fuel with UC concentrations at or below 20 at.% UC exhibited stability and negligible density changes in a high temperature hydrogen environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

New Synthesis Strategies to Improve Co-Free LiNi0.5Mn0.5O2 Cathodes: Early Transition Metal d0 Dopants and Manganese Pyrophosphate Coating

In this work, we report solution-based doping and coating strategies to improve the electrochemical performance of the Co-free layered oxide cathode LiNi0.5Mn0.5O2 (LNMO). Small amounts of d0 dopants (e.g., Mo6+and Ti4+, 0.5-1 at. %) increase the cathode’s specific capacity, cycling stability, and rate capability. More specifically, a Mo-doped cathode with the nominal composition LiNi0.495Mn0.495Mo0.01O2 achieves a high reversible capacity of 180 mAh/g at 20 mA/g with good retention at higher rates (e.g., 120 mAh/g at 100 mA/g). Effects of 1 at.% Mo dopant on the cathode structure were studied using a suite of characterization tools including X-ray diffraction (XRD), Raman spectroscopy, and transmission electron microscopy (TEM). These measurements demonstrate that Mo6+ reduces Li+/Ni2+cation mixing and mitigates undesirable phase transformations near the cathode surface during cycling. This work also reports the use of an inorganic Mn2P2O7 coating which enhances cycling stability, presumably through formation of a stable cathode electrolyte interphase (CEI) layer. Overall, the synthesis approaches reported herein are quite general and can potentially be expanded to other high voltage LIB cathodes.

Co-free cathodes, d0 cation, pyrophosphate coating↗

Voltage cycling as a dynamic operation mode for high temperature electrolysis solid oxide cells

Solid Oxide Electrolysis Cells (SOECs) have emerged as a promising technology for the efficient production of H2 via high-temperature electrolysis. However, power input from dynamic energy sources remains a significant challenge for their long-term stability. It is important to analyze the tolerance of cells under dynamic operation conditions. This study focuses on evaluating the impact of voltage cycling on the performance and durability of electrode-supported SOECs. We explore the operational limits and degradation mechanisms of SOECs subjected to various voltage conditions and find that the cells have high tolerance for dynamic voltage. Voltage cycling between 1.3 V and 1.5 V for 9000 cycles does not damage the cell. Conversely, cycling to higher voltages (≥1.7 V) results in accelerated degradation. Advanced characterization is used to screen for various degradation modes post operation. Within the oxygen electrode, XRD and STEM EDS find compositional and phase evolution in all voltage cycled samples including increased decomposition of the air electrode resulting in cation migration. Microstructural analysis of the fuel electrode from nano-CT data shows minimal change throughout the sample set and no evidence of Ni migration, indicating the fuel electrode is stable and not impacted by cycling to higher voltages within the timeframe studied.

Zhu, Zhikuan↗

In situ investigation of the deformation behaviors of Fe 20 Co 30 Cr 25 Ni 25 and Fe 20 Co 30 Cr 30 Ni 20 high entropy alloys by high-energy X-ray diffraction

In situ synchrotron-based high-energy X-ray diffraction (HE-XRD) technique was employed to investigate the mechanical behaviors and microstructural evolution of face-centered cubic (FCC) Fe 20 Co 30 Cr 25 Ni 25 and Fe 20 Co 30 Cr 30 Ni 20 high entropy alloys (HEAs) during tensile deformation. Fe 20 Co 30 Cr 30 Ni 20 HEA has a good combination of strength (ultimate tensile strength of 864 ± 35 MPa) and ductility (elongation of 0.627 ± 0.021). The HE-XRD investigation reveals that Fe 20 Co 30 Cr 30 Ni 20 HEA has the transformation-induced plasticity (TRIP) effect, which starts at a critical stress of ~555 MPa. Furthermore, transmission electron microscopy confirmed this deformation-induced new phase is hexagonal-close-packed structured ε-martensite, which follows an orientation relationship of {111} γ //(0001) ε , and (110) γ //[11 2 ¯ 0] ε with the FCC γ matrix. The observation of deformation twins in the deformed samples of the two studied HEAs proves that twinning-induced plasticity (TWIP) effect occurs in both HEAs. The combination of TRIP and TWIP effects lead to the high strength, large ductility and improved strain hardening behavior of Fe 20 Co 30 Cr 30 Ni 20 HEA.

36 MATERIALS SCIENCE↗

The effects of Fe and Al on the phase transformations and mechanical behavior of β-Ti alloy Ti-11at.%Cr

Alloy composition is important for developing desired microstructures in beta-titanium (β-Ti) alloys during thermomechanical processing. In this work, the relatively low-cost alloying elements iron (Fe) and aluminum (Al) were added to a base Ti-11at%chromium (Cr) alloy and in-situ and ex-situ examination of the microstructural evolution and mechanical properties during tension, hardness, elevated temperature X-ray diffraction (XRD) and resonance ultrasound spectroscopy (RUS) were performed. The 400 °C XRD revealed that Ti-11at%Cr underwent β-to-ω and β-to-α transformations. Adding 0.85at% Fe reduced the volume fraction of the ω- and α-phases, and adding 5.3at% Al inhibited the β-to-ω transformation. The 400 °C RUS showed that the alloys containing the ω phase exhibited an increase in shear modulus (G) of ~140%, while the ω-free alloys exhibited an increase of only ~102%. The hardness and strength values of the ω-containing Ti-11Cr and Ti-11Cr-0.85Fe increased with increased ω-phase volume fraction when the ω/α ratio was greater than 0.25. The relationship between ω/α volume ratios and mechanical properties is discussed.

36 MATERIALS SCIENCE↗

Transformation of a ceramic precursor to a biomedical (metallic) alloy: Part I – sinterability of Ta 2 O 5 and TiO 2 mixed oxides

Mixed Ta 2 O 5 – TiO 2 binary system was studied by a combination of differential thermal analysis (DTA), scanning electron microscopy-energy dispersive spectrometry (SEM-EDS), X-ray diffraction (XRD) and in situ high temperature X-ray diffraction (HT-XRD) techniques. Different compositions of the mixed oxide powders were fabricated by ball–milling the powdered compositions, pelletizing the homogenized composite powders, and heating the green pellets in air at different temperatures for fixed time intervals. The sintered pellets were evaluated and characterized with respect to porosity, morphology, and phase distribution. DTA runs of the un-sintered powders indicated the onset temperatures for both exothermic and endothermic changes in the binary system. Significant amount of sintering was observed to take place at temperatures higher than 900 °C. Both room and high temperature X-ray diffraction patterns exhibited consistency in phase formation. A ternary compound (TaTiO 4 ) and a ternary solid solution (Ti 0.33 Ta 0.67 O 2 ) were observed to form in both room and high temperatures in addition to the respective binary phases (Ta 2 O 5 and TiO 2 ). A sintering temperature in the range 900–1000 °C was observed to be adequate to achieve the requisite mechanical strength and optimum internal porosity (40–48%) for subsequent electrochemical polarization experiments.

36 MATERIALS SCIENCE↗

Mesoscale-architecture-based crack evolution dictating cycling stability of advanced lithium ion batteries

The cracking phenomenon of Ni-rich NMC (LiNixMnyCo1-x-yO2, x = 0.6) secondary particles is frequently discovered and believed to be one of critical reasons deteriorating the long-term cycling stability of NMC cathode in lithium ion batteries (LIBs). However, the initiation and evolution of those cracks is still controversial due to the limited quantification especially by in situ monitoring, leading to the challenge of identifying an efficient approach to inhibit the formation of the fractures during repeated cycling. Herein, the irreversible, anisotropic cycling lattice and mesoscale expansion/shrinkage of nano-grain during the first cycle, as revealed by in situ X-ray diffraction (XRD) and in situ atomic force microscopy (AFM), have been quantified and confirmed to be the dominant driving forces of microcracks initiation at the grain boundaries. These microcracks preferentially nucleates at the core region with random oriented nano-grains in early stage. The further growth and aggregation of microcracks into macrocrack eventually results in microfracture propagation radially outward to the periphery region with more uniform nano-grain orientation. This mesoscale nano-grain architecture controlled cracking process highlight the importance of predictive synthesis of cathode materials with controllable multiscale crystalline architecture for high-performance LIBs.

25 ENERGY STORAGE↗

Mechanisms of phosphate removal by Micron-Scale Zero-Valent iron

Extensive studies have been carried out on phosphate removal by zero valent iron (ZVI). Different mechanisms of phosphate removal by ZVI such as phosphate adsorption and precipitation of ferric phosphate and vivianite (Fe 3 (PO 4 ) 2 ·8H 2 O) have been reported. In this study, attenuated total reflectance-Fourier transform infrared spectroscopy (ATR-FTIR), X-ray powder diffraction (XRD), and X-ray absorption near edge spectroscopy (XANES) were used to elucidate the mechanisms of phosphate removal by micron-scale zero-valent (mZVI). Time-series ATR-FTIR analysis revealed that phosphate was removed by ZVI in two steps: a) adsorption of phosphate in the first 3 days, and b) slow conversion of the adsorbed phosphate into the precipitated form in 6 days. XRD and XANES analysis determined that iron oxides, ferric ion, and ZVI with oxide layer inhibited the formation of vivianite, resulting in the formation of phosphate precipitates, such as H 2 Fe(II)P 2 O 7 . The results suggested that phosphate is removed by ZVI mainly through precipitation under aerobic conditions. In conclusion, the knowledge gained in this study improves the fundamental understanding of the phosphate removal processes via ZVI.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Type-II superconductivity at 9K in Pb–Bi alloy

In the present work, we report the synthesis of Pb–Bi alloy with enhanced Tc of up to 9K, which is higher than that of Pb. The alloy is synthesized via a solid-state reaction route in the vacuum-encapsulated quartz tube at 700°C in an automated furnace. The synthesized sample is characterized by X-ray Diffraction(XRD) and Energy dispersive X-ray analysis(EDAX) for its phase purity and elemental composition. Rietveld refinement of XRD reveals that the end product is a majority hexagonal Pb 7 Bi 3 , with minor rhombohedral Bi. The electronic transport measurement shows metallic behavior with the Debye temperature of 108K and a superconductivity transition temperature (T c ) below 9K, which is the maximum to date for any reported Pb–Bi alloy, Pb or Bi at ambient pressure. Partial substitution of Bi at the Pb site may modify the free density of electronic states within the BCS model to attain the optimum T c , which is higher by around 2K from the reported T c of Pb. The superconductor phase diagram derived from magneto-transport measurements reveals that the synthesized alloy is a conventional superconductor with an upper critical field (H c2 ) of 3.9 T, which lies well within the Pauli paramagnetic limit. The magnetization measurements carried out following ZFC(Zero Field Cool) protocols infer that the synthesized alloy is a bulk superconductor below 9K. The isothermal M-H(Magnetization vs. Field) measurements performed below T c establish it as a type-II superconductor. Furthermore, the specific heat capacity measurements show that the Pb–Bi alloy is a strongly coupled bulk superconductor below around 9K with possibly two superconducting gaps.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗