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Laser ultrasonic assessment of the effects of porosity and microcracking on the elastic moduli of nuclear graphites
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In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening
Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.
Ab initio study of interstitial cluster interaction with Re, Os, and Ta in W
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The role of grain size in He bubble formation: Implications for swelling resistance
Here, nanocrystalline metals are postulated as radiation resistant materials due to their high defect and particle (e.g. Helium) sink density. Here, the performance of nanocrystalline iron films is investigated in-situ in a transmission electron microscope (TEM) using He irradiation at 700 K. Automated crystal orientation mapping is used in concert with in-situ TEM to explore the role of grain orientation and grain boundary character on bubble density trends. Bubble density as a function of three key grain size regimes is demonstrated. While the overall trend revealed an increase in bubble density up to a saturation value, grains with areas ranging from 3000 to 7500 nm 2 show a scattered distribution. An extrapolated swelling resistance based on bubble size and areal density indicated that grains with sizes less than 2000 nm 2 possess the greatest apparent resistance. Moreover, denuded zones are found to be independent of grain size, grain orientation, and grain boundary misorientation angle.
Microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln
Palladium is being investigated as a potential additive to metallic fuel to bind fission product lanthanides, with the goal of reducing or preventing fuel-cladding chemical interactions (FCCI). A primary cause of FCCI is the lanthanide fission products moving to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. The current study is the microstructural characterization of annealed U-20Pu-10Zr-3.86Pd and U-20Pu-10Zr-3.86Pd-4.3Ln in wt. %, where Ln = 53Nd-25Ce-16Pr-6La, using scanning electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). In both alloys, the matrix is comprised of ζ-U 0.4 Pu 0.6 and δ-(U,Pu)Zr 2 . Based on the matrix compositions, modifications to the room temperature extrapolated U-Pu-Zr ternary phase diagram are suggested. In U-20Pu-10Zr-3.86Pd, there is very little δ phase, due to formation of PdZr 2 . In U-20Pu-10Zr-3.86Pd-4.3Ln, a lanthanide-rich phase is present, although it does not have the crystal structure for 53Nd-25Ce-16Pr-6La. Furthermore this phase cannot be identified based on known compounds. (Ln,Pu)Pd is the primary lanthanide phase, with Pu substituting into the crystal structure in place of Nd.
Comparative analysis of microstructure and reactive sites for nuclear graphite IG-110 and graphite matrix A3
Nuclear graphite and graphite matrix are both high-purity, fine-grain, graphitized materials designed for use in Fluoride-Salt Cooled High Temperature Reactors (FHRs), High Temperature Gas Cooled Reactors (HTGRs), and Molten Salt Reactors (MSRs). Their ability to chemisorb radioisotopes (such as tritium) and their chemical stability is dependent upon the density of reactive carbon sites (RCS), which varies among grades of graphite and changes with neutron irradiation. While nuclear graphites have been extensively characterized, much less data is available for graphite matrix. The current study performs a comparative analysis of IG-110 nuclear graphite and A3 graphite matrix, by X-ray Diffraction (XRD), Raman, and H 2 uptake at 700 °C with H 2 partial pressure range of 300 Pa and 22 kPa. At 10 kPa, the RCS occupied by H 2 uptake per number of carbon atoms at the edge of a crystallite, H/C edge , is 600(300) appm for IG-110 and 2300(200) appm for A3. Prior empirical studies for nuclear graphite establish a linear relationship between H/C and C edge /C and degree of graphitization. The present study shows that these correlations no longer hold when extrapolating to graphite matrix; the type of defects and the defect density at the crystallite surface are different between IG-110 and A3 due to differences in raw materials and manufacturing method.
Kinetics and transport of hydrogen in graphite at high temperature and the effects of oxidation, irradiation and isotopics
The kinetics of uptake and desorption impact the performance of graphite as a vector for tritium in high-temperature fission reactors and in the blanket of fusion reactors. Graphite components in these reactors are exposed to temperatures > 500 °C and H 2 partial pressures of few Pa and desorption temperatures are limited to < 1600 °C; limited data is available at these conditions. Here we review the mechanisms for uptake in, transport and desorption of hydrogen from graphite at high temperature, compiling data on uptake rates, diffusion coefficients and activation energies and providing a discussion of the impact of irradiation, pre-oxidation and isotope. At FHR conditions, trapping impacts uptake rates, leading to a reduction in apparent diffusivity by 35 to 80% compared to higher partial-pressure uptake. Timelines for desorption are not clearly defined; extrapolating from available data, at 1150 °C desorbing 80% of tritium uptaken at FHR conditions may take from 100 to 10,000 h.
On the thermal oxidation of nuclear graphite relevant to high-temperature gas cooled reactors
Thermal oxidation of nuclear graphite components is highly undesirable because it can cause structural and property degradation that negatively affect a reactor's intended operation. In this work, the body of knowledge of nuclear graphite oxidation is highlighted, including when O 2 , H 2 O, and/or CO 2 are the oxidant. Oxidation conditions relevant to high-temperature gas-cooled reactors (i.e., when oxidation could occur either as an acute or chronic phenomenon) are emphasized. Here, the objective is to summarize graphite oxidation data in a practical and accessible way to inform future research and regulatory requirements. Although each grade of nuclear graphite is different, the oxidation mechanism has underlying commonalities. Although oxidation behavior is grade dependent, the general temperature dependence is well described by a sequence of elementary steps which become rate limiting. Because the regime transition temperature depends on sample microstructure, size, and oxidant supply rate, extrapolating results beyond the experimental range should be done cautiously. Gravimetric oxidation rate measurements generally replicate well. However, caution must be exercised when rates are estimated by other methods or for samples that deviate significantly in size. Air oxidation data for IG-110, NBG-18, and PCEA graphite is critically reviewed to emphasize this point. Despite the amount of experimental data, gaps remain. Sample size and shape effects are not fully explained. Data on the oxidant penetration depth are insufficient. Analytical assessments demonstrate that lower temperature oxidation does not necessarily imply that oxidation is uniform throughout the bulk. Oxidation occurs faster at higher temperature but is more localized to the exposed surface. Paradoxically, at equal mass loss percentage, low temperature oxidation leads to greater property degradation than at high temperature. The isolated effect of oxidation is important; however, a gap remains in the systematic understanding of any potential effect of neutron irradiation on graphite structure and reactivity.
Porosity, swelling, and composition evolution in high-burnup monolithic U-Mo fuel
The microstructural progression of very high-burnup (>8 × 10 21 fissions/cm 3 ) monolithic uranium-molybdenum (U-10wt.%Mo) was analyzed, providing crucial insights into the behavior of post-recrystallized nuclear fuel, where scant data exists. Three focused ion beam cuboids sourced from a fuel plate with varying local burnups of 8.86 × 10 21 , 9.05 × 10 21 , and 9.36 × 10 21 fissions/cm 3 were characterized. The porosity and composition of the samples were evaluated to characterize the evolution of the microstructure as a function of fission density and different locations on the fuel plate, while simultaneously isolating plate-specific parameters such as Zr diffusion barrier thickness, hot-isostatic-pressing conditions, enrichment, and reactor conditions. The porosity was segmented, and the three-dimensional distribution of the porosity was extrapolated from the two-dimensional segmentation. The composition was assessed and quantified using energy-dispersive X-ray spectroscopy areal mapping. The porosity fraction increased as a function of the burnup from 27.77±0.51, 35.12±1.54, and 37.71±0.44 % for 8.86 × 10 21 , 9.05 × 10 21 , and 9.36 × 10 21 fissions/cm 3 , respectively. When compared to literature, the porosity volume fraction plateaus at burnups greater than 6 × 10 21 fissions/cm 3 , while the pore size grows linearly as a function of fission density. The number of large pores increased in number density as a function of burnup, while the smallest pores (<0.3 µm) increased up to 9.05 × 10 21 fissions/cm 3 , followed by a decrease at 9.36 × 10 21 fissions/cm 3 . The delamination and cracking in the fuel plate propagated through an interconnected porosity sublayer identified ∼5 µm from the diffusion barrier. The local swelling of the specimens was within or near the prediction bounds of the Robinson-Williams model for local swelling. The fission products, strontium, barium, cerium, and cesium, precipitated into the pores, while neodymium accumulated adjacent to the pores. Furthermore, these findings have direct implications for the development of fuel performance codes and the accurate documentation of the microstructure evolution in high burnup U-Mo, thus enhancing the safety and efficiency of nuclear fuel usage.
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.
Interpreting experimental measurements of helium bubbles using stochastic cluster dynamics models of heterogeneous nucleation and growth in irradiated ferritic alloys
Among a number of other advantageous features, ferritic/martensitic steels are being considered as fusion reactor structural materials due to their low intrinsic swelling under irradiation. However, under high-energy neutron irradiation, He produced through (n, α) reactions stabilizes vacancy clusters, which then act as seeds for bubble formation and growth, negating the intrinsic swelling resistance of these alloys. Standard models of irradiation damage accumulation and microstructural evolution consider homogeneous nucleation as the basis for bubble formation and growth. However, this generally does not explain the large bubble densities and sizes observed experimentally under a number of different conditions. Here, we propose a more realistic physical model of bubble nucleation, formation, and growth designed to capture recent experimental measurements of He-bubble formation and evolution during co-implantation of He and Fe ions in model ferritic alloys. We find that experimental results are explained only when the following three features are considered simultaneously: (i) heterogeneous nucleation of He-vacancy bubbles at defect sinks (e.g., dislocations, grain boundaries, and second-phase precipitates), (ii) nucleation and growth of bubbles via the ‘trap mutation’ mechanism (i.e., spontaneous production of Frenkel pairs due to absorption of He atoms), and (iii) transition from a growth-limited, He-stabilized bubble regime to a ‘runaway’ void-type growth scenario. The model is implemented into a stochastic cluster dynamics framework that takes advantage of cluster size grouping methods to accelerate the simulations, allowing us to reach 10 dpa of simulated irradiated dose, and to capture cluster sizes in excess of 20 nm. Further, a careful extrapolation exercise conducted assuming classical nucleation theory leads to excellent agreement with the experimental measurements at 50 dpa of irradiation.
Nanochannel electrodes facilitating interfacial transport for PEM water electrolysis
Proton-exchange membrane water electrolyzers (PEMWEs) are a promising technology for green hydrogen production; however, interfacial transport behaviors are poorly understood, hindering device performance and longevity. Here, we first utilized finite-gap electrolyzer to demonstrate the possibility of proton transfer through water in PEMWEs. The measured high-frequency resistances (HFRs) exhibit a linear trend with increasing gap distance, where extrapolation shows a lower value compared with HFRs in regular zero-gap electrolyzers, indicating that ohmic resistance could be further reduced. We introduce nanochannels to facilitate mass transport, as evidenced by both liquid-fed and vapor-fed electrolysis. Nanochannel electrodes achieve a voltage reduction of 190 mV at 9 A·cm –2 compared with the Ir-PTEs without nanochannels. Furthermore, nanochannel electrodes show negligible degradation through 100,000 accelerated-stress tests and over 2,000 h of operation at 1.8 A·cm –2 with a decay rate of 11.66 μV·h –1 . These results provide new insights into localized transport dynamics for PEMWEs and highlight the significance of interfacial engineering for electrochemical devices.
Elongated Fullerene Unlocks Oxygen-Resistant Perovskite Solar Minimodules
State-of-the-art encapsulation cannot prevent the permeation of oxygen into perovskite solar cells (PSCs). Here, we report the finding that p-i-n structured PSCs degrade quickly under oxygen exposure. Oxidation of C60, rather than perovskites, dominates the degradation. Chemical absorption of oxygen by C60 is identified, which impairs its electron transport property. We modify the electron transport layers to address the instability under oxygen exposure. The elongated fullerene C70 is found to react with oxygen orders of magnitude slower than C60. In combination with a compact SnO2 buffer fabricated by atomic layer deposition, which can slow down the oxygen diffusion, the resulting unencapsulated PSCs with C70 retained 90% of their initial efficiency after 1-sun illumination in pure oxygen for 1,200 h at 70 degrees C, improving stability by hundreds of times. Testing of unencapsulated perovskite minimodules of different perovskite compositions with C70 gives extrapolated lifetimes of 17-41 years at 50 degrees C.
Fast charge implications: Pack and cell analysis and comparison
This study investigates the effect of 50-kW (about 2C) direct current fast charging on a full-size battery electric vehicle's battery pack in comparison to a pack exclusively charged at 3.3 kW, which is the common alternating current Level 2 charging power level. Comparable scaled charging protocols are also independently applied to individual cells at three different temperatures, 20 °C, 30 °C, and 40 °C, to perform a comparative analysis with the packs. Dominant cell-level aging modes were identified through incremental capacity analysis and compared with full packs to gain a clear understanding of additional key factors that affect pack aging. While the cell-level study showed a minor impact on performance due to direct current fast charging, the packs showed a significantly higher rate of capacity fade under similar charging protocols. This indicates that pack-level aging cannot be directly extrapolated from cell evaluation. With this being said, delayed fast charging, completing shortly before discharge, was found to have less of an impact on battery degradation than conventional alternating current Level 2 charging
Low-lying states and total internal partition sums of CH
The electronic structure and spin-orbit states of the CH radical have been systematically investigated using multi-reference configuration interaction (MRCI) and single-reference coupled-cluster (CC) methods. These calculations were performed in conjunction with large correlation-consistent basis sets of quadruple-, quintuple-, and sextuple-ζ quality. To achieve high accuracy, electronic energies for all states were extrapolated to the complete basis set (CBS) limit, enabling the detailed construction of potential energy curves and determination of reliable spectroscopic constants. Spin-orbit coupling effects were explicitly incorporated, and vibrational energy levels were computed via Numerov analysis. Furthermore, the resulting values exhibit good to excellent agreement with available experimental data. Dipole moment and transition dipole moment curves were evaluated to assess the opacity characteristics of CH, revealing that transitions such as Χ 2 Π (u′′ = 0) → Α 2 Δ (u′ = 0), Χ 2 Π (u′′ = 0) → Β 2 Σ − (u′ = 0), Χ 2 Π (u′′ = 0) → C 2 Σ + (u′ = 0), and Χ 2 Π (u′′ = 0) → D 2 Σ + (u′ = 3) are particularly probable. Finally, the total internal partition function sum (TIPS) of CH was computed over a broad temperature range (10–30,000 K) based on our high-accuracy ab initio results.
Non-linear thermal evolution of the crystal structure and phase transitions of LaFeO{sub 3} investigated by high temperature X-ray diffraction
The crystal structure, anisotropic thermal expansion and structural phase transition of the perovskite LaFeO{sub 3} has been studied by high-temperature X-ray diffraction from room temperature to 1533 K. The structural evolution of the orthorhombic phase with space group Pbnm and the rhombohedral phase with R3{sup Macron }c structure of LaFeO{sub 3} is reported in terms of lattice parameters, thermal expansion coefficients, atomic positions, octahedral rotations and polyhedral volumes. Non-linear lattice expansion across the antiferromagnetic to paramagnetic transition of LaFeO{sub 3} at T{sub N}=735 K was compared to the corresponding behavior of the ferroelectric antiferromagnet BiFeO{sub 3} to gain insight to the magnetoelectric coupling in BiFeO{sub 3}, which is also multiferroic. The first order phase transition of LaFeO{sub 3} from Pbnm to R3{sup Macron }c was observed at 1228{+-}9 K, and a subsequent transition to Pm3{sup Macron }m was extrapolated to occur at 2140{+-}30 K. The stability of the Pbnm and R3{sup Macron }c polymorphs of LaFeO{sub 3} is discussed in terms of the competing enthalpy and entropy of the two crystal polymorphs and the thermal evolution of the polyhedral volume ratio V{sub A}/V{sub B}. - Graphical abstract: Aniostropic thermal evolution of the lattice parameters and phase transition of LaFeO{sub 3}. Highlights: Black-Right-Pointing-Pointer The crystal structure of LaFeO{sub 3} is studied by HTXRD from RT to 1533 K. Black-Right-Pointing-Pointer A non-linear expansion across the Neel temperature is observed for LaFeO{sub 3}. Black-Right-Pointing-Pointer The ratio V{sub A}/V{sub B} is used to rationalize the thermal evolution of the structure.
Electronic structure and optical properties of the nonlinear optical crystal Pb{sub 4}O(BO{sub 3}){sub 2} by first-principles calculations
Pb{sub 4}O(BO{sub 3}){sub 2} has a layered-type arrangement with optimally aligned BO{sub 3} triangles. The optical band gap is 3.317 eV obtained via the extrapolation method from the UV-vis-IR optical diffuse reflectance spectrum, consequently the absorption edge is about 374 nm. Density functional calculations using a generalized gradient approximation were utilized to investigate the electronic structures and optical properties of Pb{sub 4}O(BO{sub 3}){sub 2}. The calculated band structures show a direct gap of 2.608 eV, which is in agreement with the experimental optical band gap. A delocalized {pi} bonding of BO{sub 3} triangles and the stereo-effect of the lone pair 6s{sup 2} of lead cations are studied in electron densities. The birefringence is about 0.039-0.061 with the wavelength larger than about 375 nm. The calculated second-order susceptibility d{sub 24}=3.5 d{sub 36} (KDP) which is well consistent with the powder SHG intensity. - Graphical abstract: The density of state (DOS) show that the bottom of the valence bands is mainly derived from of the lone pair 6s{sup 2} of Pb{sup 2+}, and the top of the valence band is attributed to the hybridization orbitals from B-O groups. Calculated electronic structures indicate that the BO{sub 3} group with typical delocalization {pi} orbitals and strongly distorted lead oxygen polyhedra with highly asymmetric lobes on lead cations make a large SHG effect in Pb{sub 4}O(BO{sub 3}){sub 2}. Highlights: Black-Right-Pointing-Pointer Lone pair effect on Pb{sup 2+} and delocalization {pi} orbital in BO{sub 3} group is studied. Black-Right-Pointing-Pointer The combination of PbO{sub n} (n=3,4,5) and BO{sub 3} group makes Pb{sub 4}O(BO{sub 3}){sub 2} a large SHG effect. Black-Right-Pointing-Pointer Pb{sub 4}O(BO{sub 3}){sub 2} is a direct gap material with the gap 2.608 eV by the ab initio method. Black-Right-Pointing-Pointer The calculated birefringence is about 0.039-0.061 with the wavelength of about 375 nm. Black-Right-Pointing-Pointer The second-order susceptibility d{sub 24}=3.5d{sub 36} (KDP) in accordance with the experimental result.