Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “sequential”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Numerical study of the effect of propped surface area and fracture conductivity on shale gas production: Application for multi-size proppant pumping schedule design

Hydraulic fracturing is a technique extensively used in the oil and gas industry, where water, proppant (sand) and additives are injected into unconventional reservoirs to enhance the recovery of shale hydrocarbon. Although some previous studies have developed pumping schedules that maximize gas production for a single-size proppant, there are very few studies that consider the effect of varying proppant diameters across pumping stages on shale gas production. Motivated by this, we carried out an extensive sensitivity analysis to determine the effect of different proppant diameters on the average fracture conductivity (FC), average propped surface area (PSA) and cumulative shale gas production volume. We found out that the cumulative shale gas production volume depends on both the average PSA and average FC. We also found out that small-diameter proppant resulted in higher average PSA and lower average FC, whereas large-diameter proppant resulted in lower average PSA and higher average FC. Hence, we designed a multi-size proppant pumping schedule considering both of these parameters into account for simultaneously propagating multiple fractures to maximize shale gas production from unconventional reservoirs. Since the size of injected proppant particles determines the average PSA and average FC for the propped hydraulic fractures, we developed a novel framework called Sequentially Interlinked Modeling Structure (SIMS) to predict the average PSA, average FC and cumulative shale gas production volume at the end of 10 years for a given pumping schedule. Then, we used this SIMS framework to obtain a multi-size proppant pumping schedule that maximizes shale gas production. In closing, we demonstrated that the obtained pumping schedule gives a cumulative shale gas production volume greater than the values obtained from the existing pumping schedules.

03 NATURAL GAS↗
DOE OSTI2020

On residual stress analysis and microstructural evolution for stainless steel type 304 spent nuclear fuel canisters weld joint: Numerical and experimental studies

Chloride-induced stress corrosion cracking (CISCC) in the weldments of austenitic stainless steel canisters is one of the primary safety concerns during the dry storage of used nuclear fuel at Independent Spent Fuel Storage Installations in coastal areas. In order to evaluate the CISCC behavior in the canister, a 3D sequentially coupled thermo-mechanical finite element model was built to simulate the residual stresses induced by two intersecting longitudinal and circumferential multi-pass welds in austenitic stainless steel mockup canister. Weld-induced residual stresses from simulation were in good agreement with the experimental measurement results by deep-hole drilling and contour methods. Through-wall high tensile axial and hoop stresses were observed in both the fusion zone and heat affected zone (HAZ) in the longitudinal and circumferential weld, respectively. On the other hand, the microstructure within the weldments of 304 stainless steel mockup was characterized using optical microscopy, scanning electron microscopy, electron back-scattered diffraction and energy dispersive X-ray spectroscopy methods. A few amount of MnS inclusions in both the base metal and HAZ, and a lot of (Mn,Si) x O y inclusions in the fusion zone were found. Meanwhile, there is some α'-martensite in the base metal, while the martensite was reversed in the HAZ. In the fusion zone, there is a lot of δ-ferrite. The martensite reversion phenomenon was also verified by the combination of Feritscope test and thermal simulation, which could predict a martensite reversion temperature as 400C°. The welding residual stresses and microstructure analysis results could offer the instructive information for the prediction of CISCC behavior of the spent nuclear fuel canisters.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A quantitative study of retention and release of deuterium and tritium during irradiation of y-LiAlO2 pellets

This study reports on the quantification of deuterium in ion-irradiated gamma-LiAlO2 pellets as a function of dose and temperature. The gamma-LiAlO2 pellets were sequentially irradiated with He+ and D2+ ions to the same fluences of 5E16, 1E17 and 2E17 He++D+/cm2 at 188 K. Additional irradiation was performed to 1E17, 2E17 and 4E17 He++D+/cm2 at 573 K. A set of the pellets irradiated at 188 K was shipped and stored at low temperatures from 80 to 132 K and characterized using time-of-flight secondary ion mass spectrometry at ~173 K. The deuterium depth profiles show a Gaussian-like distribution in the low-temperature pellets. The total deuterium retention is found to be directly proportional to the ion fluence. About 27 at.% of the implanted deuterium atoms were released from the pellet irradiated to 2E17 He++D+/cm2 at 188 K during storage at room temperature for ~1 month. Retention of the trapped or bound deuterium during ion irradiation at 573 K increases initially with ion fluence and tends to saturate at a high fluence. The amount of the released deuterium is observed to be quantitatively consistent with that of the released tritium from similar standard pellets during neutron irradiation at 573 K.

Deuterium retention and release, ion irradiation, ↗
DOE OSTI2022

Behavior of helium cavities in ion-irradiated W-Ni-Fe ductile-phase toughened tungsten

This study reports on the distribution of helium (He) cavities in a hot-rolled W-Ni-Fe ductile-phase toughened tungsten (DPT W) composite irradiated to a dose and a helium concentration that are comparable to those in the material after 5-year irradiation in a conceptual fusion power plant. Here, the DPT W sample consists of W particles embedded in a ductile-phase NiFeW matrix with a nominal composition of 90W-7Ni-3Fe by weight. It was hot-rolled to a thickness reduction by 87% (87R DPT W). Sequential irradiation of the material with 1.2 MeV Ni + ions to a fluence of 2.15×10 16 Ni + /cm 2 and 90 keV He + ions to 6.5×10 15 He + /cm 2 was performed at 973 K. Larger He cavities with a lower number density are observed in NiFeW than W. Helium cavities are aggregated preferentially along the NiFeW/W interphase boundary. This behavior is not observed along the W/W grain boundary under the same irradiation conditions. A similar distribution of He cavities with comparable sizes is observed in 87R DPT W irradiated with He + ions only, suggesting that the He cavity distribution is not significantly affected by pre-irradiation of Ni + ions at 973 K. The data from this study does not provide any evidence for formation of visible voids or Ni precipitates in W phase irradiated with Ni + ions only. Diffusion and clustering of He atoms and mobile vacancies during ion irradiation at 973 K are responsible for the formation and growth of the He cavities.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Mechanical and Thermophysical properties of low enriched uranium-10wt% molybdenum rolled foils

In this study, low-enriched (~19.75% 235 U) U-10Mo(wt%) foils were evaluated for both mechanical and thermophysical properties to establish baseline characteristics prior to irradiation testing. Cast U-10Mo ingots were sequentially hot and cold rolled to two different final foil thicknesses, resulting in different levels of cold work being present in the finished foils. The mechanical response of the U-10Mo foils was evaluated in the temperature range of 293–623 K. Increasing test temperatures resulted in decreases of both yield and ultimate tensile strength. Comparison of results to earlier work performed on depleted U-10Mo rolled foils indicates that the material in this study has lower strength and increased ductility. However, a comparison of the two different foil thicknesses in this study found that they were nearly identical in yield and ultimate tensile strength, differing by only 10–20 MPa (<2%). Thermal conductivity was calculated from the measurements of specific heat, linear thermal expansion, and thermal diffusivity from 323 to 1273 K on both thicknesses of the foils. These data were approximately 5–10% lower than previously published data above 573 K and were similar to data for temperature ranges below 573 K. Differences are likely related to differences in grain size and/or impurity content and variation in fabrication history.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Helium interaction with solutes and impurities in neutron-irradiated nanostructured ferritic alloys: A first principles study

Density functional theory calculations are performed to explore the binding between He and alloying solutes, impurities, and transmutation products expected in neutron irradiated nanostructured ferritic alloys (NFAs), here 14YWT is taken as an example. Elements that exhibit significant binding (attraction) with an interstitial He are Y (binding energy = 0.46 eV), Mg (0.32), O (0.33), Ti (0.16), and C (0.15). Those that provide significant binding to a substitutional He are O (1.44), Y (1.24), N (0.73), H (0.56), Mg (0.52), Ti (0.34), Si (0.34), C (0.33), Al (0.32), Ni (0.26), Ta (0.23), and Mn (0.16). The presence of these elements in Fe matrix could reduce the transport of He towards oxide particles, dislocations, and internal boundaries, and could promote He bubble nucleation in the matrix. For convenience, we compile existing binding energy data of He with He n and He n V (He-vacancy) clusters. Dissociation pathway analysis reveals that, in general, the most likely dissociation of a He n V cluster is by a sequential emission of individual He atoms. Furthermore, larger bubbles are more prone to dissociation than smaller ones. In addition, we estimate the binding energy (segregation energy) of He in bulk Y 2 Ti 2 O 7 (YTO) single crystal, YTO/Fe interface, and YTO particle embedded in Fe, with respect to interstitial He in Fe, from existing formation energies of He in these structures. We also compile available data of He binding with Fe self-interstitial atom (SIA), SIA clusters, and edge and screw dislocations. Note that given the absence of DFT data, the binding with SIA clusters and dislocations are gathered from simulations with empirical potentials. Finally, the data presented in this paper is important to inform multiscale simulations of He bubble accumulation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Ion irradiation study of lithium silicates for fusion blanket applications

Here, this study reports on the microstructural and compositional evolutions, deuterium release and lithium loss behavior in lithium silicates. Pellets of biphasic Li 4 SiO 4 and Li 2 SiO 3 were fabricated using hot pressing of powders. Sequential irradiation with Si + , He + and D + ions was performed up to 773 K to emulate one-year 6 Li burnup in 6 Li 4 SiO 4 inside the SlimCS DEMO. Crystalline Li 4 SiO 4 phase was not observed in the irradiated depth region of all the samples in this study. In spite of full amorphization in the near-surface region, Li 2 SiO 3 irradiated to a high dose at 773 K remained crystalline in the damage peak region, suggesting that Li 2 SiO 3 is more irradiation resistant to amorphization than Li 4 SiO 4 . Radiolysis may be primarily responsible for silicate decomposition and amorphization. Data from this study also show that D release from the pellet is very efficient during ion irradiation at 773 K, which is accompanied with a significant Li loss. Thermal annealing for 10 min at 773 K for the room-temperature irradiated pellet leads to a nearly complete D release without an observable Li loss. A concept of Ni coating on tritium breeder materials is discussed with supporting data to minimize Li loss without a significant impact on tritium diffusion and release.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Microstructural and compositional evolutions in $γ$-LiAlO 2 pellets during ion irradiation at an elevated temperature

This study reports on the microstructural, phasic and compositional evolutions of γ-LiAlO 2 during ion irradiation. Polycrystalline γ-LiAlO 2 pellets were irradiated sequentially with He + and D 2 + ions to the same combined fluences of up to 3 × 10 17 (He + +D + )/cm 2 at 773 K. The irradiated pellets were characterized using scanning transmission electron microscopy and atom probe tomography. Surface amorphization likely due to radiolysis and planar defects as a possible precursor for formation of precipitates are created at 5 × 10 16 (He + +D + )/cm 2 , followed by the formation of nano-sized precipitates and fractures at higher doses. Spinel-like precipitates of non-stoichiometric LiAl 5 O 8 and gas-filled cavities are observed to grow with increasing dose. Faceted precipitates and rounded cavities appear at 2 × 10 17 (He + +D + )/cm 2 . With further increasing ion fluence to 3 × 10 17 (He + +D + )/cm 2 , amorphization of the precipitates takes place and micron-sized fractures appear. Surface exfoliation could occur at an extremely high ion fluence. There are compositional changes in the γ-LiAlO 2 pellets during the microstructural evolution. In the precipitate and amorphized regions, Li concentrations decrease to ~7 and 3.7 at.% from 25 at.% in γ-LiAlO 2 , respectively. This study reveals a full-cycle microstructural evolution with corresponding compositional changes in γ-LiAlO 2 pellets during ion irradiation at 773 K. In conclusion, the data could help model, assess, and predict the material performance during neutron irradiation in reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Understanding the effect of minor alloying elements on helium bubble formation in ferritic-martensitic steels

Ferritic-martensitic steels are promising structural materials for advanced nuclear reactors. To minimize long-term radioactivity, reduced-activation ferritic-martensitic steels have been developed by substituting high-activation elements like Ni and Mo with low-activation elements such as W. However, the impact of these alloying modifications on helium bubble formation, which plays a key role in material swelling, remains unclear. Here, in this study, we compared helium bubble formation in ferritic-martensitic steel T91 and reduced-activation ferritic-martensitic steel F82H. Both materials were irradiated with sequential 100 keV, 150 keV, and 200 keV helium ions to a dose of 0.5 dpa and a helium concentration of 9,000 appm at 500°C. The helium bubbles in F82H exhibited a larger average size and a lower density than those in T91, suggesting differences in minor alloying elements may influence the bubble growth. Here, to investigate the effects of these alloying elements, we characterized radiation-induced segregation near bubbles and grain boundaries. Prominent Ni-Mn-Si enriched clusters were found near bubbles in T91, while only Mn-Si enriched clusters were found near bubbles in F82H. In addition, the obvious Cr enrichment near grain boundaries was absent around bubbles in both steels. The different segregation trends among elements revealed the variations in element diffusion mechanisms and the different sink biases between bubbles and grain boundaries. Cr enrichment near grain boundaries is mostly driven by interstitial-mediated diffusion. However, since bubble growth relies on net vacancy flux, vacancy-mediated diffusion plays a dominant role in controlling element segregation near bubbles. Therefore, Cr enrichment was not found near bubbles. Because of preferential vacancy-drag diffusion for Ni, Si and Mn, these elements were enriched near bubbles. Due to the strong binding energies of vacancies with these solute atoms, the vacancy diffusivity can be reduced near these solutes. Therefore, the more prominent Ni-Si-Mn clustered near helium bubbles in T91 lead to stronger suppression of helium bubble growth compared to F82H.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Machine learning with knowledge constraints for process optimization of open-air perovskite solar cell manufacturing

Perovskite photovoltaics (PV) have achieved rapid development in the past decade in terms of power conversion efficiency of small-area lab-scale devices; however, successful commercialization still requires further development of low-cost, scalable, and high-throughput manufacturing techniques. One of the critical challenges of developing a new fabrication technique is the high-dimensional parameter space for optimization, but machine learning (ML) can readily be used to accelerate perovskite PV scaling. Herein, we present an ML-guided framework of sequential learning for manufacturing process optimization. We apply our methodology to the Rapid Spray Plasma Processing (RSPP) technique for perovskite thin films in ambient conditions. With a limited experimental budget of screening 100 process conditions, we demonstrated an efficiency improvement to 18.5% as the best-in-our-lab device fabricated by RSPP, and we also experimentally found 10 unique process conditions to produce the top-performing devices of more than 17% efficiency, which is 5 times higher rate of success than the control experiments with pseudo-random Latin hypercube sampling. Our model is enabled by three innovations: (a) flexible knowledge transfer between experimental processes by incorporating data from prior experimental data as a probabilistic constraint; (b) incorporation of both subjective human observations and ML insights when selecting next experiments; (c) adaptive strategy of locating the region of interest using Bayesian optimization first, and then conducting local exploration for high-efficiency devices. Furthermore, in virtual benchmarking, our framework achieves faster improvements with limited experimental budgets than traditional design-of-experiments methods (e.g., one-variable-at-a-time sampling). This framework shows the capability of incorporating researchers’ domain knowledge into the ML-guided optimization loop; therefore, it has the potential to facilitate the wider adoption of ML in scaling to perovskite PV manufacturing.

14 SOLAR ENERGY↗
DOE OSTI2025

Exciton fission enhanced silicon solar cell

While silicon solar cells dominate global photovoltaic energy production, their continued improvement is hindered by the single-junction limit. One possible solution is to use molecular singlet exciton fission to generate two electrons from each absorbed high-energy photon. We demonstrate that the long-standing challenge of coupling molecular excited states to silicon solar cells can be overcome using sequential charge transfer. Combining zinc phthalocyanine, aluminum oxide, and a shallow junction crystalline silicon microwire solar cell, the peak charge generation efficiency per photon absorbed in tetracene is (138% ± 6%), comfortably surpassing the quantum efficiency limit for conventional silicon solar cells and establishing a new, scalable approach to low-cost, high-efficiency photovoltaics.

14 SOLAR ENERGY↗
DOE OSTI2023

Distributed approximate minimal Steiner trees with millions of seed vertices on billion-edge graphs

In this report, we present a parallel 2-approximation Steiner minimal tree algorithm and its MPI-based distributed implementation. In place of expensive distance computations between all pairs of seed vertices, the solution we employ exploits a cheaper Voronoi cell computation. Our design leverages asynchronous processing and message prioritization to accelerate convergence of distance computations, and harnesses vertex and edge centric processing to offer fast time-to-solution. We demonstrate scalability and performance using real-world graphs with up to 128 billion edges and 512 compute nodes, and show the ability to find Steiner trees with up to one million seed vertices. Using 12 data instances, we present comparison with the state-of-the-art exact solver, SCIP-Jack, and two sequential 2-approximate algorithms. We empirically show that, on average, the total distance of the Steiner tree identified by our solution is 1.1290 times greater than the Steiner minimal tree – well within the theoretical approximation bound of 2.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Insights into the phase stability window, phase transformation behavior, and anisotropic thermal expansion of rhombohedral bismuth oxide

Rhombohedral bismuth oxide has emerged as a promising SOFC electrolyte material, particularly at low-to-intermediate temperatures. The stability window of this phase was explored using a co-doped system, using yttrium and lanthanum or aluminum, over 7.5–25 % total dopant content and 0.907–1.154 Å weighted average dopant cationic radius. The phase transformation behavior of rhombohedral-rich phase mixtures was also investigated in situ from 360 to 656 °C to assess the effects of minor impurity phases and thermal evolution pathways. Our results reveal that low-dopant, rhombohedral-rich compositions exhibit poor structural stability upon heating. Anisotropic thermal expansion behavior was observed over 450–656 °C, serving as a sensitive indicator of sequential phase changes, including the rhombohedral β2-to-β1 transition (∼450 °C), monoclinic phase formation, and delayed or incomplete formation of the cubic phase (∼ 550 °C). These multiple unfavorable phase transitions compromise the mechanical robustness required for SOFC operation. This study underscores the need for compositional tuning to balance ionic conductivity with thermal phase stability in rhombohedral Bi 2 O 3 -based systems.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Multiscale characterization of an additively manufactured property graded Ni-base alloy for molten-salts\supercritical-CO 2 heat exchangers

The sequential optimization of strength and corrosion resistance in conventional alloy design procedures often results in a tradeoff between environmental degradation and optimal mechanical properties. A concurrent optimization of these two material properties is essential to efficiently develop high temperature alloys that can withstand harsh environments increasingly required for carbon–neutral energy technologies. In this work, we demonstrate the feasibility of directed energy deposition (DED) to manufacture a dual-corrosion resistant Ni-based alloy (Hastelloy N and Haynes 282), that meets the high temperature operation requirements of a molten-salts\supercritical-CO 2 (sCO 2 ) heat exchanger. A combination of multiscale characterization techniques and computational thermodynamics was employed to evaluate the cracking susceptibilities during fabrication and predict the microstructural stability of the material. Very good agreement was achieved between the observed and predicted phases and phase fractions of the as-printed material. A careful characterization of the transition zone between the two terminal alloy chemistries revealed potential precipitation strengthening (γ') on the Hastelloy N side while columnar-shaped M 23 C 6 and γ' precipitates that formed at grain boundaries (GBs) of the transition zone likely minimized the local interfacial energies. Both these mechanisms are believed to increase the interfacial stability but their performance at high temperatures requires further investigation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Nuclear Data Sheets for A=24

Here, evaluated spectroscopic data and level schemes from radioactive decay and nuclear reaction studies are presented for 24 N, 24 O, 24 F, 24 Ne, 24 Na, 24 Mg, 24 Al, and 24 Si. This evaluation for A=24 supersedes the earlier one by R. B. Firestone (2007Fi14). Highlights of this evaluation are the following: In 24 O, the 7.4- and 7.65-MeV groups reported in 1 H( 24 O,p') (2012Ts03) and 9 Be( 26 F,nX) (2011Ho05), respectively, are considered as separate states in this evaluation, since the former one was observed to decay by only one neutron emission to 23 O, while the latter one by two sequential neutrons to 22 O. Based on assigned configurations, the 7.4- and 7.6-MeV groups are tentatively assigned negative and positive parities, respectively. Additional study would be useful to determine if these states are same or different. In 24 Na, 2014Fi01 (n,γ) proposed Jπ=(0,1,2) - for 4048.8 keV level based on weaker population from the capture state (J π =1 + ,2 + ) and mentioned to reexamine the J π =0 + assignment, suggested in earlier evaluations (1990En08, 2007Fi02). This evaluation includes the first observation of E0 transition, strength ρ 2 (E0)=0.380 70, for the 6432 keV transition from 0 + state at 6432 to 0 + g.s. in 24 Mg reported in 2020Do10. Authors proposed the first-excited 0+ state of 24 Mg to be superdeformed. In 24 Al, the excited level energies deduced from ( 3 He,t) measurements reported by different research groups vary significantly and it was not clear if some of those levels were same or different. Additional measurements are needed.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗