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At least 1,243 records · Page 69

Parameter development and characterization of laser powder directed energy deposition of Nb – Alloy C103 for thin wall geometries

This work focuses on the parameter development and microstructural characterization of Nb-based alloy C103 for thin wall structures produced via laser powder – directed energy deposition (LP – DED). Laser power and scanning speeds were varied as part of a design of experiments to identify adequate print parameters. Combinations were evaluated for relative density, porosity, and geometrical accuracy. A combination of a laser power of 1420 W and scanning speed of 14 mm/s resulted in a relative density >99%, exhibited a consistent weld bead profile and was used for further microstructural evaluation. A mix of optical and scanning electron microscopy (SEM) revealed small, slightly elongated grains along the edges and large epitaxial grains in the central region along the Build – Transverse view. Scanning electron microscopy (SEM) and energy dispersive spectroscopy (EDS) analysis revealed Hf rich columnar cells in the center that transition into an evenly spaced cellular structure towards the edges of the built sample. Electron backscatter diffraction (EBSD) scans show a sharp [001] texture when looking at the cross section of the sample along the build direction. The Build – Scan view revealed a zig – zag pattern that follows the back – and – forth deposition strategy that was used. Finally, microhardness measurements were taken in the as – built (AB) and stress relieved (SR) conditions to baseline preliminary mechanical properties. The AB condition exhibited a large amount of scatter in the data and averages up to 11% larger than the SR condition. The reduction in scatter upon applying the SR cycle are indicative of a large concertation of dislocations present in the AB condition.

36 MATERIALS SCIENCE↗

Characterization of nitrogen-containing species produced from nitric acid/water systems

The speciation of gases formed in the vapor phase above HNO 3 -H 2 O mixtures at room temperature is measured and characterized using Fourier-transform infrared (FTIR) spectroscopy. Although studies of speciation of compounds have been performed in the solution phase, in this study we focus solely on the gas phase species formed over HNO 3 -H 2 O mixtures and their concentrations as a function of time. The nitric acid concentration in the experiments range from 14 to 4 M. The NO 2 , NO, and N 2 O (collectively known at NOx) gases are measured by FTIR at 1585, 1850, and 2208 wavenumbers, respectively. The concentrations of these NOx species ranges from 0 to 3 ppm and change over time as a function of acid concentration. Two purities of commercially available HNO 3 are used to prepare the HNO 3 -H 2 O mixtures. An analysis of thermodynamic data shows that the formation of NOx gases without the presence of a catalyst are thermodynamically favorable.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Multi-scale characterization and simulation of impact welding between immiscible Mg/steel alloys

Vaporizing foil actuator spot welding method is used in this paper to join magnesium alloy AZ31 and uncoated high-strength steel DP590, which are typically considered as un-weldable due to their high physical property disparities, low mutual solubility, and the lack of any intermetallic phases. Characterization results from scanning electron microscopy (SEM) and high-resolution transmission electron microscopy (HRTEM) of the weld interface indicate that the impact creates an Mg nanocrystalline interlayer with abundant Fe particles. The interlayer exhibits intact bonding with both DP590 and AZ31 substrates. To investigate the fundamental bond formation mechanisms at the interface, a finite element (FE)-based process simulation is first performed to calculate the local temperature and deformation at the interface under the given macroscopic experimental condition. Finally, taking the FE results at the interface as inputs, molecular dynamics (MD) simulations are conducted to study the interlayer formation at the Mg/Fe interface during the impact and cooling. The results found a high velocity shearing-induced mechanical mixing mechanism that mixes Mg/Fe atoms at the interface and creates the interlayer, leading to the metallurgical bond between Mg/steel alloys.

36 MATERIALS SCIENCE↗

Characterization of the mud displacement in an enlarged wellbore: An integrated rock-fluid model

Cement-mud displacement plays a crucial role in the sealability of cement sheaths. Irregular geometric features of a wellbore due to washout can have a negative impact on mud and cement mobilization. An unstable interface between two fluids always leads to mud channeling, interfluid mixing, and cement contamination, degrading the cement quality. Many factors, such as mechanical and rheological properties of fluids, annulus geometry, flow pattern, and flow rate, significantly influence the displacement efficiency. This study investigates the characterization of the mud displacement in an irregular horizontal well using a 3D computational fluid dynamics (CFD) model. Mud is displaced in an enlarged wellbore by geopolymer and neat class G cement. The wellbore geometry is developed based on the caliper log data from an unconventional shale well in the Tuscaloosa Marine Shale (TMS) lithology. The effects of pump rate, density difference, and mud contamination are evaluated by numerical simulations. The results present those residual muds mainly exist in the upper annulus of the enlarged section. Geopolymer has a better sealing performance and can resist more water-based mud (WBM) contaminations than neat class G cement. The scenario with a low mud-cement density difference and high cement injection rate results in a high cement volume fraction, mitigating the gas migration.

58 GEOSCIENCES↗

Method to account for natural fracture induced elastic anisotropy in geomechanical characterization of shale gas reservoirs

Shale has been usually recognized as a transverse isotropic (TI) medium in conventional geomechanical log interpretation due to its laminated nature. However, when natural fractures exist in the shale rock, additional elastic anisotropy is introduced, converting laminated Shale to an orthorhombic (OB) medium. Previous studies illustrate that neglecting the natural fracture induced anisotropy in shale geomechanical log interpretation could lead to inaccurate evaluations of elastic moduli and in-situ stresses. In this paper, a new method is developed to account for the natural fracture induced anisotropy in geomechanical log interpretation based upon the TI acoustic model developed by the author and a characterization technique of elastic wave anisotropy (Sayers, 1991). The new OB model incorporates the four acoustic log data inputs and five modeling constraints in a nonlinear optimization algorithm to solve for the nine independent stiffness coefficients of an OB rock, and further to solve for the geomechanical properties and in-situ stress profiles in an OB formation. The new method was validated with a Marcellus Gas Shale field case. Both the new OB model and the conventional TI model were applied to interpret the minimum horizontal stress profile for the same formation. By comparing the results, the OB model is more robust than the TI from two aspects. First, the average stress magnitude predicted by the OB model is closer to the one measured by the Diagnostic Fracture Injection Test (DFIT). Second, the OB model predicts a more obvious stress barrier between the lower Marcellus and upper Onondaga Limestone than the TI model does. Finally, the predicted stress barrier is consistent with the observation of the microseismic events of a horizontal well drilled and completed nearby, which reveals that no hydraulic fracture propagates downward through the bottom boundary of Marcellus Shale into the underlying Onondaga Limestone.

03 NATURAL GAS↗

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

On the use of charged particles to characterize precipitation in irradiated reactor pressure vessel steels with a wide range of compositions

Nuclear reactor lifetimes may be limited by nano-scale Cu-Mn-Ni-Si precipitates (CRPs and MNSPs) that form under neutron irradiation (NI) of pressure vessel (RPV) steels, resulting in hardening and ductile to brittle transition temperature increases (embrittlement). Physical models of embrittlement must be based on characterization of precipitation as a function of the combination of metallurgical and irradiation variables. Here we focus on rapid and convenient charged particle irradiations (CPI) to both: a) compare to precipitates formed in NI; and, b) use CPI to efficiently explore precipitation in steels with a very wide range of compositions. Atom probe tomography (APT) comparisons show NI and CPI for similar bulk steel solute contents yield nearly the same precipitate compositions, albeit with some differences in their number density, size and volume fraction (f) dose (dpa) dependence. Yet, the overall precipitate evolutions are very similar. Advanced high Ni (>3 wt%) RPV steels, with superior unirradiated properties, were also investigated at high CPI dpa. For typical Mn contents, MNSPs have Ni 16 Mn 6 Si 7 or Ni 3 Mn 2 Si phase type compositions, with f values that are close to the equilibrium phase separated values. In contrast, in steels with very low Mn and high Ni, Ni 2-3 Si silicide phase type precipitate compositions are observed; and when Ni is low, the precipitate compositions are close to the MnSi phase field. Low Mn significantly reduces, but does not eliminate, precipitation in high Ni steels. A comparison of dispersed barrier model predictions with measured hardening data suggests that the Ni-Si dominated precipitates are weaker dislocation obstacles than the G phase type MNSPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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↗

Atom Probe Tomography Characterization of High-Dose Ion Irradiated MA957

This paper presents a comprehensive characterization study of the microstructural responses to ion irradiation of the oxide dispersion strengthened (ODS) ferritic alloy MA957 using atom probe tomography (APT). The MA957 specimens were irradiated by 1.8 MeV Cr3+ ions to doses and temperatures relevant to advanced fast reactors (100 and 500 dpa at 400-500 °C). The average sizes of the YTiO particles decreased slightly after irradiation at all temperatures, with the most pronounced change at 400 °C. The particle number densities for 400 and 500 °C were slightly below number densities for the unirradiated condition, while the values were slightly higher at 420 and 450 °C. This behavior is consistent with the Nelson-Hudson-Mazey rate theory model that considers the competing effects between recombination and dissolution. An increase in the Y/Ti ratio of the YTiO particles was observed at the two lower temperatures. The formation of Cr-rich precipitates (a') was observed in specimens irradiated at temperatures up to 450 °C. Irradiation-induced segregation on grain boundaries was also investigated. Grain boundary element concentration profiles and ion maps demonstrate differences among the irradiation temperatures, with most datasets differing from the unirradiated condition. A key focus of this ion irradiation study is the high-dose response, and this quantitative assessment revealed relatively small but distinct effects of irradiation on microstructure evolution. Overall, these results showing minor second phase evolution suggest that MA957 should retain mechanical performance generally well up to 500 dpa for the irradiation temperatures studied.

Wang, Jing↗

Atom probe tomography characterization of ion and neutron irradiated Alloy 800H

Alloy 800H is a high temperature and creep resistant alloy, and is considered as a candidate alloy for use in Generation IV nuclear reactor systems. To clarify the alloy's behavior under irradiation, samples were subjected to single ion irradiations up to 20 dpa at 440°C, a dual ion/He irradiation to 17 dpa at 460°C, and a neutron irradiation to 17 dpa at 385°C. The irradiated microstructures were characterized using atom probe tomography to complement previously published transmission electron microscopy studies. Additionally, after single ion irradiation, sparse fine Al and Ti clusters were observed after 1dpa, while high number densities of nanoscale Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops developed after 10 and 20 dpa. The microstructure formed during single ion irradiation exhibited a strong depth dependence. No significant differences were observed between the single and dual ion irradiations. While Ni-Al-Ti clusters, Cr-Ti rich carbides, and Si-decorated dislocation loops were also observed after neutron irradiation, the neutron-irradiated microstructure differed from those found in the ion-irradiated samples.

36 MATERIALS SCIENCE↗

Characterization of PyC/SiC interfaces with FIB-SEM tomography

Tristructural-isotropic (TRISO) fuel is being considered for multiple reactor designs due to its exceptional performance. This work studies pyrocarbon (PyC) and silicon carbide (SiC) substrates that are near-representative of the inner pyrocarbon (IPyC) and SiC in TRISO fuel. Focused ion beam (FIB)-scanning electron microscopy (SEM) tomography was used to understand the influence of different processing routes on the PyC/SiC interface structure. Additionally, this work outlines a methodology for characterizing the IPyC/SiC interface when considering fission product accommodation and retention in TRISO fuel design. The high-fidelity images collected with FIB-SEM tomography can be used to achieve a more in-depth interpretation of microstructural parameters and pore-shape analysis that can inform simulations and processing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Characterization of the molten salt FMgNaK through ab initio molecular dynamics and experimental density measurements

The selection of a salt species for use in a molten salt reactor MSR is a key part of any MSR design. However, many salts have sparse or no thermophysical property data sets, especially those with higher melting points. One such salt is the eutectic mixture of (NaF)0.345(KF)0.59(MgF 2 )0.065, or FMgNaK, and is explored here through ab initio molecular dynamics (AIMD) simulation (1023-1273K) as well as experimental measurement of the liquid density using the Archimedean method (973-1223K). Predicted densities from AIMD are similar to experimentally-measured densities, though with more noise, suggesting that the small scale necessitated by AIMD simulations may be problematic for simulating FMgNaK in particular. Here the coefficient of thermal expansion is predicted from simulation, and salt structure is characterized. Mg-F-Mg chaining is observed in the salt network, though the low concentration of Mg inhibits chaining on the scale that is observed in (LiF) 0.67 (BeF 2 ) 0.33 , or FLiBe.

36 MATERIALS SCIENCE↗

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe 2 Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Microstructural characterization of U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln

Antimony is being investigated as a potential additive to metallic fuel to control fuel-cladding chemical interactions (FCCI). The most detrimental elements involved in FCCI are fission product lanthanides, leading to brittle intermetallics and low melting eutectic phases. Previous investigations of Sb as an additive focused on U-10Zr, in wt. %, as the fuel. The current investigation expands that to include Pu in the fuel. Here, two alloys, U-20Pu-10Zr-4Sb and U-20Pu-10Zr-4Sb-4Ln (wt. %, Ln=53Nd-25Ce-16Pr-6La) have been investigated using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) to characterize the fuel as-cast microstructure and the microstructure after introduction of lanthanides. Sb reacts with Zr initially, forming Zr 2 Sb and Zr 5 Sb 3 , with as much as 20 at. % interstitial Pu present. In the presence of lanthanides, Sb forms Ln 4 Sb 3 with the lanthanides, containing ~14 at. % Pu. The Pu is substitutional for the lanthanides in the crystal lattice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction between Minor Actinides bearing U-Pu-Zr Fuel and AIM1 Cladding

Minor actinides (MA) significantly contribute to the long-term radiotoxicity of spent nuclear fuel (SNF). Separating MA from SNF and incorporating it into metallic fuels for fast reactor transmutation is a potential method to reduce this radiotoxicity. Here, this study focuses on transmission electron microscopy characterization of two samples from the fuel cladding chemical interaction (FCCI) region of an americium (Am) and neptunium (Np)-bearing (MA-bearing) uranium-plutonium-zirconium (U-Pu-Zr) fuel irradiated in the Phenix fast reactor to 9.5 % FIMA burnup at approximately 550 °C cladding temperature. The results show that despite the complex chemical interactions between MA and AIM1 cladding elements, excessive FCCI was not induced, and Am penetration depth in the cladding limited to less than 4 µm. Np remained mostly inside fuel. The Zr-rich compounds layer effectively limited the accumulation of lanthanide on the inner cladding surface. Overall, the FCCI behavior between investigated MA-bearing U-Pu-Zr fuel and AIM1 cladding is benign.

Chemical interaction↗

High-resolution characterization of ceramic-metal interface of TiN coating on ferritic-steels for nuclear application

Advanced fuel cladding is critical for fast reactors, offering sufficient thermal conductivity, mechanical and dimensional stability and radiation tolerance of the cladding base material. Additionally, it must provide corrosion resistance and high temperature coolant compatibility on the cladding outer surface, as well as chemical stability on the cladding inner wall against fuel cladding chemical interaction (FCCI). TiN ceramic coating has been considered an effective diffusion barrier for inner and outer cladding-walls for enhanced performance. The TiN-metal interface microstructure and chemistry play a critical role in coating bond strength and integrity under harsh conditions. High-resolution transmission electron microscopy characterization of ceramic-metal interface at atomic resolution in unirradiated, irradiated and thermal cycled conditions were performed. The interface remained intact after irradiation up to 200 dpa or thermal cycling five times up to 550 °C. In conclusion, this work discusses the potential impact of these results on coating performance and design for advanced claddings.

36 - MATERIALS SCIENCE↗

Synthesis and characterization of the dinuclear cobalt(III) complex: [(C 2 F 5 ) 3 Co(μ-F)] 2 2–

Here, reaction of the versatile tris(perfluoroethyl) cobalt(III) precursor [fac-(MeCN) 3 Co(C 2 F 5 ) 3 ] with [NMe 4 ]F and [PPh 4 ]Cl in THF formed the unexpected cobalt(III) bridging fluoride dimer ([(C 2 F 5 ) 3 Co(μ-F)] 2 2– . The new fluoro-organometallic cobalt(III) complex was characterized by NMR and UV-vis spectroscopies, X-ray crystallography, cyclic voltammetry, and by computational methods. In the strongly coordinating solvent MeCN, ([(C 2 F 5 ) 3 Co(μ-F)] 2 2– exhibits dynamic processes on the NMR timescale which are consistent with solvent coordination. However, in the weakly coordinating solvent CH 2 Cl 2 , a more static structure is observed suggesting that the dimer retains its structure in solution state. An electrochemical analysis of ([(C 2 F 5 ) 3 Co(μ-F)] 2 2– was performed, and the data are compared to previously reported cobalt(III) perfluoroethyl complexes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗