Search NASA⌕ Search

SEARCH · Search NASA

Results for “matrix diffusion”

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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Pore-size dependence and characteristics of water diffusion in slitlike micropores

The temperature dependence of the dynamics of water inside microporous activated carbon fibers (ACF) is investigated by means of incoherent elastic and quasielastic neutron-scattering techniques. The aim is to evaluate the effect of increasing pore size on the water dynamics in these primarily hydrophobic slit-shaped channels. Using two different micropore sizes (similar to 12 and 18 angstrom, denoted, respectively, ACF-10 and ACF-20), a clear suppression of the mobility of the water molecules is observed as the pore gap or temperature decreases. Suppression, we found, is accompanied by a systematic dependence of the average translational diffusion coefficient D-r and relaxation time [tau(0)] of the restricted water on pore size and temperature. We observed D-r values and tested against a proposed scaling law, in which the translational diffusion coefficient D-r of water within a porous matrix was found to depend solely on two single parameters, a temperature-independent translational diffusion coefficient D-c associated with the water bound to the pore walls and the ratio theta of this strictly confined water to the total water inside the pore, yielding unique characteristic parameters for water transport in these carbon channels across the investigated temperature range.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Solutions of the two-dimensional Hubbard model: Benchmarks and results from a wide range of numerical algorithms

Numerical results for ground-state and excited-state properties (energies, double occupancies, and Matsubara-axis self-energies) of the single-orbital Hubbard model on a two-dimensional square lattice are presented, in order to provide an assessment of our ability to compute accurate results in the thermodynamic limit. Many methods are employed, including auxiliary-field quantum Monte Carlo, bare and bold-line diagrammatic Monte Carlo, method of dual fermions, density matrix embedding theory, density matrix renormalization group, dynamical cluster approximation, diffusion Monte Carlo within a fixed-node approximation, unrestricted coupled cluster theory, and multireference projected Hartree-Fock methods. Comparison of results obtained by different methods allows for the identification of uncertainties and systematic errors. The importance of extrapolation to converged thermodynamic-limit values is emphasized. Furthermore, cases where agreement between different methods is obtained establish benchmark results that may be useful in the validation of new approaches and the improvement of existing methods.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

A Literature Review on the Study of Moisture in Polymers

This literature review covers the main chemical and physical interactions between moisture and the polymer matrix. Fickian versus Non-Fickian diffusion behaviors are discussed in approximating the characteristics of moisture sorption. Also, bound water and free water sorbed in polymers are distinguished. Methods to distinguish between bound and free water include differential scanning calorimetry, infrared spectroscopy, and time-domain nuclear magnetic resonance spectroscopy. The difference between moisture sorption and water sorption is considered, as well as the difficulties associated with preventing moisture sorption. Furthermore, specific examples of how moisture sorption influences polymers include natural fiber-polymer composites, starch-based biodegradable thermoplastics, and thermoset polyurethane and epoxies.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a “fuel body”, in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 8-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both to be focused on obtaining fission product transport data to support improvements in modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and postirradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These goals were combined in the AGR-3/4 irradiation experiment, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body, in which the outer graphite ring included a floor and cap to fully enclose the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to obtain data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compacts 1-3, 4-3, 10-1, and 10-2

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined into one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011 and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support modeling improvement. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite (Petti et al. 2005). These two goals were combined in the AGR-3/4 irradiation, which consisted of twelve independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. There were two capsule types: a standard capsule and a fuel body in which the outer graphite ring included a floor and cap that fully enclosed the fuel (Stempien et al. 2018a). The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from matrix and graphite at accident temperatures (Demkowicz 2017)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Determination of Average Burnup in AGR-3/4 Compacts 1-4 and 7-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The irradiation test began on December 14, 2011, and ended on April 12, 2014 (Collin 2016). The originally planned AGR-3 and AGR-4 irradiation experiments were both focused on obtaining data on fission product transport to support the improvement of modeling. The AGR-3 experimental plan was focused on gaseous and metallic fission product release from the kernels and diffusion in the coatings during irradiation and post-irradiation safety testing. The AGR-4 experimental plan was focused on diffusivities and sorptivities in the compact matrix and reactor graphite. These two goals were combined in the AGR-3/4 irradiation, which consisted of 12 independently monitored capsules that each contained four AGR-3/4 compacts in a single stack surrounded by an inner ring of matrix or graphite and an outer ring of graphite. Two capsule types were used: a standard capsule and a so-called fuel body, in which the outer graphite ring included a floor and cap that fully enclosed the fuel. The fuel body design supported post-irradiation safety testing of the intact fuel and ring assembly to provide data on fission product transport and release from the matrix and graphite at accident temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Modeling the Long-Term Behavior of Cementitious Barriers - 18188

Cementitious materials are commonly included as backfill/barrier materials in geological repositories for radioactive waste. An understanding of the long-term behavior of cement-based barriers is therefore required to provide confidence in their safety functions within an environmental safety case. Geochemical modelling has been undertaken on several different radioactive waste repository concepts (and their natural/industrial analogues), to explore the nature of cement degradation over timescales relevant to postclosure safety (many tens of thousands of years to c. 1 Ma). In radionuclide transport calculations undertaken as part of performance assessments, it is typically considered that the properties of cementitious engineered barrier materials change as they degrade, potentially resulting in an increase in porosity hydraulic conductivity, and effective diffusion coefficients. However, reactive transport modelling (supported by some analogue data) suggests that, in some repository systems, a degree of clogging of the cement pore space could occur and that some advective flow paths (cracks in the barriers, fractures in the host rock) could become (to some extent) infilled by cement and rock alteration products. This could result in a reduction in water flow and contaminant transport in the vicinity of the cement-rock interface, but some uncertainties remain (especially process couplings and the nature of water flow in complex fracture networks that are becoming clogged). In addition, the armoring of fracture surfaces could result in decreased potential for diffusion of solutes into the rock matrix. It is envisaged that ongoing work will continue to improve reactive-transport model capabilities, such that the implications cement degradation has for contaminant transport can be predicted with greater confidence. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Simulating the Effect of Fracture Connectivity on Repository Performance with GDSA Framework - 18589

This work examines the possible migration of radionuclides from a deep geologic nuclear waste repository sited in fractured crystalline host rock. The key safety concern is the potential effect on waste isolation of the inter-connectivity of the fracture network, which is primarily established by the temporal evolution of the temperature and stress fields at the time of original rock intrusion and cooling. Two end members are considered in these simulations, one with a high degree of connectivity to the biosphere, such that advective transport through the fracture and fault network controls radionuclide migration, and the other with a low degree of connectivity, such that slow diffusive transport through the crystalline rock matrix is the controlling process for radionuclide migration to the biosphere. Both end members have several areally extensive, high-transmissivity deformation zones, whose distance to the repository is an important factor for waste isolation capability. Uncertainties in fracture properties (transmissivity, orientation, radius) and fracture distribution also give rise to uncertainty in waste isolation capability (i.e., the overall connectivity of the repository with the surface biosphere). These types of natural system uncertainties are generally present in any geologic site-characterization program and thus represent an important factor in assessing repository performance in hard rock environments. The waste considered in these simulations are spent fuel rod assemblies from the U.S. commercial nuclear reactor fleet. Thermal output of the waste must be considered in the simulations and can influence the rate and timing of waste package failure, waste form degradation, and fluid flux due to thermal expansion around the repository horizon. To achieve a high-fidelity representation of radionuclide transport in fractures and rock matrix, combined with thermal energy transport and fluid flow in fractures and matrix, the mathematical model is solved numerically in a parallel high-performance computing (HPC) environment on a finite volume unstructured grid consisting of approximately 5 million cells, using Geologic Disposal Safety Assessment (GDSA) Framework (https://pa.sandia.gov), an open-source performance assessment tool for deep underground disposal of nuclear waste. GDSA Framework uses PFLOTRAN to solve the balance equations on a three-dimensional grid with heterogeneous properties, using multiple processors in a parallel configuration based on domain decomposition. The fracture networks in these simulations are originally generated as discrete fracture networks (DFNs), which are sets of two-dimensional planes distributed in a three-dimensional domain. The method used in GDSA Framework maps the stochastically generated DFN to an equivalent continuous porous medium (ECPM) domain that allows for the simulation of coupled heat flow, fluid flow, and radionuclide transport, including heat conduction through the matrix of the fractured rock, which is a process not easily modeled in a DFN representation. Computational efficiency is also greatly enhanced using the ECPM method, allowing for a realistic representation and analysis of uncertainties in a multi-realization performance assessment of a deep geologic repository. The effect of fracture connectivity on the waste isolation safety function, as brought to light by these GDSA Framework simulations, points to the importance of including a realistic representation of uncertainties in fracture properties and distribution (effectively, uncertainty in spatial heterogeneity) in repository safety assessment simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Towards High Performance Oxide-Dispersion-Strengthened Alloys

The current first-generation candidate swelling-resistant alloys envisioned to replace austenitics are ferritic and ferritic-martensitic steels, with second-generation alloys being yttria dispersion-hardened variants of these steels. In ferritic-martensitic alloys, ferrite grains always swell earlier than martensite. Currently, most oxide dispersion-strengthened (ODS) alloys have only a ferrite matrix, thus giving up the first line of defense against swelling provided by tempered martensite. Towards development of better swelling-resistant ODS alloys, dual phase (ferrite + martensite) ODS alloys with oxide particles uniformly distributed in both ferrite and martensite phases is needed. Furthermore, fundamental understanding on mechanisms governing dispersoid stability and its correlation with void swelling is greatly needed. We reported here further alloy improvement by exploring new 9Cr ODS alloy with majority of oxide dispersoids uniformly distributed in tempered martensite grains. The alloy was irradiated by 3.5 MeV Fe self ions up to 600 dpa (displacements per atom) and at temperatures ranging from 325 to 625 deg. C. According to SRIM code, the dpa profile peaks at about 1000 nm. First of all, ion irradiation reduces the dispersoid sizes. The dispersoid sizes begin to increase beyond ∼1000 nm. Second, higher irradiation temperatures result in larger dispersoids on average. Third, there is no significant dispersoid size difference between 100 and 200 dpa irradiation. The equilibrium sizes are quickly saturated. Size and coherency of a dispersoid is closely related. In martensite grains, fine dispersoids tend to be coherent while coarse dispersoids tend to incoherent. We found that at a low irradiation temperature, coherent small dispersoids become smaller and their interface coherency remains. In a comparison, incoherent large dispersoids disappear after irradiation. At an intermediate irradiation temperature, the final equilibrium dispersoid sizes increase. At a high irradiation temperature, the final dispersoid sizes further increase and some large semi-coherent dispersoids survive. In ferrite grains, as-received dispersoids are relatively smaller in comparison with martensite grains. Their equilibrium size changes after irradiation follow a similar trend but large dispersoids completely disappear even under high temperature ion irradiation. We propose a model to explain the above dispersoid size changes under various irradiation conditions. The model considers the competing effects between ballistic-collision- assisted dispersoid dissolution into matrix and thermally driven back-diffusion for recovery. When the competing reaches a balance, an equilibrium size distribution curve is determined. For dispersoids smaller than the equilibrium size, they grow upon ion irradiation. Large dispersoids may shrink and reach the equilibrium size but ultra-large ones will completely disappear if they shrinkage paths are far away from the curve. The model further considers the complexity from the dpa rates, and damage cascade morphology.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Hydrogen effects on thermal diffusivity and electrical resistivity of zircaloy cladding

Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. Here, in this work, we first used laser flash analysis method and four-probe configuration method to measure the thermal diffusivity and electrical resistivity of Zircaloy tubes respectively, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the laser flash analysis setup and four-probe configuration method for investigating thermal diffusivity and electrical resistivity of Zircaloy tubes. We further investigated the hydrogen effect on thermal diffusivity and electrical resistivity of Zircaloy. Hydrogen plays significant roles in the thermal diffusivity of Zircaloy, which depends on the hydrogen concentration. For higher hydrogen contents (1130 and 1820 wppm in this work), where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature at the α-Zr + δ-hydride phase regime, while increase at higher temperature at the α-Zr + β-Zr phase regime. For low hydrogen concentration, where hydride dissolved into α-Zr matrix phase at higher temperature, the thermal diffusivity is lower than non-hydrided Zircaloy-4 with similar temperature-dependent trends. Such observation demonstrates the hydrogen effects on reducing thermal diffusivity of α-Zr phase. Hydrogen increases the electrical resistivity of Zircaloy. Similar to thermal diffusivity results, the phase transformation causes a reversal of temperature-dependent trends in electrical resistivity results in Zircaloy-4 with higher hydrogen contents.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Relative Size of the Polymer and Nanoparticle Controls Polymer Diffusion in All-Polymer Nanocomposites

The dynamics of polymers in an all-polymer nanocomposite that are composed of soft cross-linked polystyrene nanoparticles and linear polystyrene have been examined. In this article, we describe how the relative size of the nanoparticle to that of the polymer chain and its rigidity impact the linear polymer chain diffusion. The results of the in situ neutron reflectivity experiments show three distinct regimes in the linear polymer diffusion. The results indicate that the inclusion of soft nanoparticles increases the amount of topological constraints and confinement effects for low matrix molecular weight polymer. At modest molecular weights where the size of the nanoparticle and polymer chain are similar, the soft nanoparticles neither inhibit nor enhance the linear polymer diffusion, while at the highest polymer matrix molecular weight, the linear polymer diffusion increases due to an increase in the constraint release mechanism by the soft nanoparticles. Thermal analysis demonstrates that the nanoparticles do not increase the free volume of the system nor do they behave as a plasticizing agent. These results are interpreted to indicate that the competition between a topological barrier effect and enhancing constraint release defines the behavior of a given all-polymer nanocomposite.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Interaction between Al and atomic layer deposited (ALD) ZrN under high-energy heavy ion irradiation

Uranium-molybdenum (U-Mo) particles dispersed in an aluminum matrix is the most promising candidate fuel to convert high-power research and test reactors in Europe from using high-enriched to using low-enriched fuel. However, chemical interaction between the U-Mo and the Al matrix leads to undesirable fuel behavior. Zirconium nitride (ZrN) is used as a diffusion barrier between the U-Mo fuel particles and the Al matrix. To understand the potential microstructural evolution of ZrN during irradiation, a high-energy heavy ion (84 MeV Xe) irradiation experiment was performed on atomic layer deposited (ALD) nanocrystalline ZrN deposited on an Al plate. A fluence of 1.86 x 10 17 ions/cm 2 , or 90.3 dpa was reached during this experiment. Both analytic transmission electron microscopy (TEM) and synchrotron microbeam X-ray diffraction (ARD) techniques were utilized to investigate the kinetics of radiation-induced grain growth of ZrN at various radiation doses based on the Williamson-Hall analyses. The grain growth kinetics can be described by a power law expression, D n - D$n\atop{0}$ = K phi, with n = 5.1. The Al-ZrN interaction products (Al 3 Zr and AIN) created by radiation-induced ballistic mixing/radiation-enhanced diffusion and their corresponding formation mechanism were determined from electron diffraction and elemental composition analysis. These experimental results were confirmed by first principle thermodynamic density functional theory (DFT) calculations. The results from this ion irradiation study were also compared to in-pile irradiation data from physical vapor deposited (PVD) ZrN samples for a comprehensive evaluation of the interaction between Al and ZrN and its influence on diffusion barrier performance.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Solute diffusion behavior during heat treatment and its impact in Sc-microalloyed Al-Cu system

Scandium (Sc) is a promising microalloying element that enhances the strength and thermal stability of aluminum (Al) alloys. However, these benefits are not fully realized in aluminum-copper (Al-Cu) systems, and corrosion resistance often declines due to the complex phase evolution and diffusion behavior of Cu and Sc. Here, to clarify this, solute diffusion behavior and Cu-Sc interaction during heat treatment (HT) were investigated using in situ synchrotron-based transmission x-ray microscopy (TXM), wide-angle x-ray scattering (WAXS), and x-ray absorption near-edge structure (XANES) spectroscopy. The results, supported by electron microscopy, reveal strong Cu-Sc bonding that significantly impedes solute diffusion, leading to inhomogeneous solute distribution and non-uniform precipitation. Moreover, the Al-Cu-Sc eutectic phase exhibits high thermal stability, resisting dissolution even near the matrix liquidus. These findings quantitatively elucidate the sluggish diffusion kinetics of Cu and Sc, which can help redesign HT schedules to improve both mechanical properties and corrosion resistance in Sc-microalloyed Al-Cu alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Ultem ® /ZIF-8 mixed matrix membranes for gas separation: Transport and physical properties

Mixed matrix membranes are promising options for improving gas separation processes. Zeolitic imidazolate frameworks (ZIFs) have a porous structure similar to conventional zeolites, being capable in principle of separating gases based on their differences in kinetic diameter while offering the advantage of having a partial organic character. This partial organic nature improves the compatibility between the sieve and the polymer, and a combination of the mentioned characteristics makes these hybrid materials interesting for the preparation of mixed matrix gas separation membranes. In this context the present work reports the preparation of Ultem ® /ZIF-8 mixed matrix membranes and their permeabilities to pure CO 2 , N 2 and CH 4 gases. A significant increase in permeability with increase in CO 2 /N 2 selectivity was observed for the mixed matrix systems as compared to the properties of the neat Ultem ® . Sorption results allowed to speculate that the ZIF-8 framework is not completely stable dimensionally, what influences the separation process by allowing gases with higher kinetic diameter than its nominal aperture to be sorbed and to diffuse through the crystal. Lastly, sorption and diffusion selectivities indicate that the higher separation performance of the mixed matrix membranes is governed by the diffusion process associated with the influence of gas molecule´s geometry.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A Micro-Mechanical Constitutive Model to Predict Hygrothermal Aging of Cross-Linked Polymers

Abstract A multi-physics material model is presented to describe the effects of temperature, oxygen, and humidity on the constitutive response of cross-linked polymers. The effect of hygrothermal damage on the mechanical integrity of the polymer matrix can be considered as the result of damage accumulation of two independent aging mechanism namely, i) thermo-oxidative, and ii) hydrolytic aging. In order to capture the mutual effects of thermo-oxidative and hydrolytic aging, an assumption has been made that each of the aging phenomenon can be superposed to each other. In fact, each of them works independently and as a result, they can compete with each other. Utilizing the theory of network decomposition, all phenomena and their correlation were modeled and thus, the strain energy function of the polymer matrix is written with respect to four independent mechanisms, i) the shrinking original matrix that has neither been attacked by water nor oxygen, ii) conversion of the first network to two new network due to the reduction and formation of cross-links, and iii) energy loss from network degradation due to attack of the water molecules to polymer active agents. Moreover, the proposed model is micro-mechanically based and is mainly relevant on thin samples due to our underlying assumption of homogeneous diffusion of oxygen and water throughout the matrix. The model has been validated against extensive data-sets obtained from experiments we specifically designed for concept validation.

Bahrololoumi, Amir↗
DOE OSTI2019

Investigation of Stress Field and Fracture Development During Shale Maturation Using Analog Rock Systems

The emergence of hydrocarbons within shale as a major recoverable resource has sparked interest in fluid transport through these tight mudstones. Recent studies suggest the importance to recovery of microfracture networks that connect localized zones with large organic content to the inorganic matrix. This paper presents a joint modeling and experimental study to examine the onset, formation, and evolution of microfracture networks as shale matures. Both the stress field and fractures are simulated and imaged. A novel laboratory-scale, phase-field fracture propagation model was developed to characterize the material failure mechanisms that play a significant role during the shale maturation process. The numerical model developed consists of coupled solid deformation, pore pressure, and fracture propagation mechanisms. Benchmark tests were conducted to validate model accuracy. Laboratory-grade gelatins with varying Young’s modulus were used as scaled-rock analogs in two dimensional Hele-Shaw cell apparatus. Yeast within the gelatin generates gas in a fashion analogous to hydrocarbon formation as shale matures. These setups allow study and visualization of host rock elastic-brittle fracture and fracture network propagation mechanisms. The experimental setup was fitted to utilize photoelasticity principles coupled with birefringence properties of gelatin to explore visually the stress field of the gelatin as the fracture network developed. Stress optics image analysis and Linear Elastic Fracture Mechanics (LEFM) principles for crack propagation were used to monitor fracture growth for each gelatin type. Observed and simulated responses suggest gas diffusion within and deformation of the gelatin matrix as predominant mechanisms for energy dissipation depending on gelatin strength. LEFM, an experimental estimation of principal stress development with fracture growth, at different stages was determined for each gelatin rheology. Here, the interplay of gas diffusion and material deformation determines the resulting frequency and pattern of fractures. Results correlate with Young’s modulus. Experimental and computed stress fields reveal that fractures resulting from internal gas generation are similar to, but not identical to, type 1 opening mode.

58 GEOSCIENCES↗
DOE OSTI2018

Formation and Toughening Mechanisms of Dispersions in Interfacial Intermetallics of Dissimilar Laser Al/Steel Joints

The key to improve the strength of Al/steel dissimilar joint is to toughen the hard and brittle interfacial intermetallics. Introducing soft and tough dispersions into the interfacial intermetallics would be a promising method. In the present study, Zn-Al alloy (Zn-22Al) was used as the filler metal to generate dispersions in the interfacial intermetallics matrix of dissimilar laser Al/steel joint. The results have shown that soft and tough dispersions, i.e., FeZn{sub 10,} Al-rich amorphous phase, and Zn solid solution, were successfully introduced into the hard and brittle Fe{sub 2}Al{sub 5−x}Zn{sub x} matrix. The formation of these dispersions was resulted from diffusion of elements Fe and Al and infiltration of liquid Zn along the grain boundaries of Fe{sub 2}Al{sub 5−x}Zn{sub x} matrix and the subsequent chemical reaction and solidification. With this kind of interfacial microstructure, the joint exhibited a stable peak fracture load of 1200 N even though the interfacial layer thickness increased from 20 to 45 μm. This was because the interfacial layer was toughened by the soft and tough dispersions. Using this example, introduction of soft and tough dispersions into hard and brittle intermetallics matrix to toughen interfacial layer and then strengthening of dissimilar joints will be highlighted.

36 MATERIALS SCIENCE↗