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Results for “grain boundary 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.

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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS2015

Alumina-Forming MAX Phases in Turbine Material Systems

Coatings for high temperature turbine components are based on low conductivity YSZ thermal barriers and protective NiAl, NiCoCrAlY bond coats. Good oxidation hot corrosion resistance, intermediate CTE, and strain tolerance of Ti2AlC and Cr2AlC MAX phases are thus of special interest. Their alumina scale growth follows a cubic law in accord with FeCrAlY alloys, with oxygen grain boundary diffusivity: Dgb 1.8 x 10-10 exp(-375 kJmole) m3s. Protective cubic kinetics are also found in high pressure burner rig (6 atm., 25 ms) and TGA tests of MAXthal 211Ti2AlC. The initial portion (0.1 hr) is dominated by fast TiO2 growth (with little evidence of scale volatility in high pressure water vapor, as found for SiO2 scales). Bulk Ti2AlC and Cr2AlC substrates show promise as potential bond coats for YSZ TBCs in 1000-1200 C furnace life (500 h) tests. Cr2AlC is proving to be very resistant to 700-900 C Na2SO4 hot corrosion and is of interest for disk alloys. Preliminary diffusion bonded Cr2AlC-superalloy hybrid couples have survived 1000 hr interrupted furnace tests at 800C with no indication of cracking or debonding. Diffusion zones of -NiAl+Cr7C3 were produced in these above 1000 C, but did not grow to any great extent after 1000 hr at 800 C. Processing as coatings presents challenges, however the basic properties of MAX phases provide novel opportunities for high temperature turbine components.

MAX phase↗
NASA NTRS1993

The role of grain boundaries in hydrogen diffusion in metals at 25 C

The effect of grain size on hydrogen diffusion at 25 C was examined for 4340 steel (body-centered cubic) and for Inconel 718 (face-centered cubic). It was found that the effect of grain size is important for body-centered cubic structures, but plays a much less important role in face centered cubic structures. Accurate measurements of hydrogen desorption coefficients during hydrogen desorption show that these are not greatly different for both types of structures.

Danford, M. D.↗
DOE OSTI2025

Speciation and diffusive dynamics in hydrated grain boundaries of complex oxide Gd2Ti2O7

Abstract Grain boundaries in polycrystalline materials significantly affect their properties, such as ionic transport, corrosion, and chemical durability. The pyrochlore compound (Gd 2 Ti 2 O 7 ) is employed as a model for complex oxides and is known for its diverse applications, including nuclear waste immobilization. Density functional theory-based first-principles molecular dynamics simulations were performed at different temperatures on the hydrated grain boundary system. The results show extensive transformations within the grain boundaries among hydrous water species (OH − , H 2 O, and H 3 O + ). The temperature dependence of self-diffusion coefficients follows Arrhenius behavior, with an activation energy of 35.9 kJ/mol for hydrogen and 46.3 kJ/mol for oxygen. The lifetime of OH − is about three to four times longer than that of H 2 O at temperatures from 800 to 2100 K, suggesting the greater stability of OH − over H 2 O, a unique characteristic of the grain boundaries. The estimated lifetime of the hydrous species decreases as the temperature increases, with an activation energy of 9.9 kJ/mol for OH − and 13.4 kJ/mol for H 2 O. While Gd 3 + is more mobile than Ti 4+ , both the Gd 3 + and Ti 4+ cations are orders of magnitude less mobile than the water species. The results suggest that water species are much more mobile within grain boundaries than in the bulk crystal and have the potential to penetrate deep into polycrystalline materials through grain boundaries, leading to grain boundary degradation and dissolution. The different mobilities of cations in complex oxides can lead to leaching of certain cations and incongruent dissolution during the chemical weathering of Earth and industrial materials.

B. Ghosh, Dipta↗
DOE OSTI2025

Mechanistic nuclear fuel performance modeling of uranium nitride

Uranium mononitride (UN) is a nuclear fuel candidate for advanced reactor designs and an alternative being considered for light water reactors due to its higher thermal conductivity and uranium density than UO 2 . As with any nuclear fuel, swelling and fission gas release are important factors for safety, while also being some of the hardest phenomena to predict with a high degree of confidence. Getting a grasp on the gas swelling behavior and release is crucial to lower the barrier for UN utilization. An accelerated swelling rate at high temperatures observed experimentally, sometimes referred to as “breakaway swelling,” further complicates the prediction of fuel performance of UN. A mechanistic model has been developed using a multiscale approach to describe the intragranular and intergranular fission gas behavior. Lower-length-scale calculations have been employed to inform models of the gas and self-diffusion behavior, resolution rate, and bubble shape. Leveraging previous work on high burnup UO 2 , two populations of intragranular bubbles are considered; small bulk bubbles and larger bubbles located along dislocations. The dislocation bubbles were found to be crucial to the overall swelling behavior, and the breakaway swelling transition was associated with the transition in the gas atom diffusion mechanism from an irradiation-induced athermal diffusion regime at lower temperatures to an intrinsic thermal equilibrium regime at higher temperatures, accelerating the growth of the dislocation bubbles. Similarly, the threshold for fission gas release was associated with the grain boundary vacancy diffusivity surpassing the gas atom diffusivity at sufficiently high temperatures, allowing the over-pressurized grain boundary bubble to grow in size and interconnect. Using thermo-mechanical models with the fission gas model, two integral fuel pin assessment cases were simulated. Finally, this work demonstrates the ability of a multiscale approach to accelerate the understanding of advanced fuel forms when experimental data is limited.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

A Finite Difference informed Random Walk solver for simulating radiation defect evolution in polycrystalline structures with strongly inhomogeneous diffusivity

Diffusivity of species and defects on grain boundaries is usually several orders of magnitude larger than that inside grains. Such strongly inhomogeneous diffusivity requires prohibitively high computational demands for modeling microstructural evolution. Here, this paper presents a highly-efficient numerical solver, combining the Finite Difference method and Random Walk model, designed for accurately modeling strongly inhomogeneous diffusion within polycrystalline structures. The proposed solver, termed Finite Difference informed Random Walk (FDiRW), integrates a customized Finite Difference (cFD) scheme tailored for fast diffusion along thin grain boundaries represented by a single-layer of nodes. Numerical experiments demonstrate that the FDiRW solver achieves an impressive efficiency gain of 1560x compared to traditional Finite Difference methods while maintaining accuracy, making it feasible for personal computer machines to handle diffusional systems with strongly inhomogeneous diffusivity across static polycrystalline microstructures. The model has been successfully applied to simulate radiation defect evolution, showcasing its scalability to engineering scales in both length and time dimensions.

36 MATERIALS SCIENCE↗
NASA NTRS1978

Effect of grain boundaries in silicon on minority-carrier diffusion length and solar-cell efficiency

The spatial variation of minority-carrier diffusion length in the vicinity of a grain boundary for a polycrystalline silicon sheet has been measured by the use of the EBIC technique. The effect of such a variation on solar-cell output has then been computed as a function of grain size. Calculations show that the cell output drops considerably for grain size smaller than three times the bulk diffusion length.

Daud, T.↗
NASA NTRS1990

Characterization of Iridium Coated Rhenium Used in High-Temperature, Radiation-Cooled Rocket Thrusters

Materials used for radiation-cooled rocket thrusters must be capable of surviving under extreme conditions of high-temperatures and oxidizing environments. While combustion efficiency is optimized at high temperatures, many refractory metals are unsuitable for thruster applications due to rapid material loss from the formation of volatile oxides. This process occurs during thruster operation by reaction of the combustion products with the material surface. Aerojet Technical Systems has developed a thruster cone chamber constructed of Re coated with Ir on the inside surface where exposure to the rocket exhaust occurs. Re maintains its structural integrity at high temperature and the Ir coating is applied as an oxidation barrier. Ir also forms volatile oxide species (IrO2 and IrO3) but at a considerably slower rate than Re. In order to understand the performance limits of Ir-coated Re thrusters, we are investigating the interdiffusion and oxidation kinetics of Ir/Re. The formation of iridium and rhenium oxides has been monitored in situ by Raman spectroscopy during high temperature exposure to oxygen. For pure Ir, the growth of oxide films as thin as approximately 200 A could be easily detected and the formation of IrO2 was observed at temperatures as low as 600 C. Ir/Re diffusion test specimens were prepared by magnetron sputtering of Ir on Re substrates. Concentration profiles were determined by sputter Auger depth profiles of the heat treated specimens. Significant interdiffusion was observed at temperatures as low as 1000 C. Measurements of the activation energy suggest that below 1350 C, the dominant diffusion path is along defects, most likely grain boundaries, rather than bulk diffusion through the grains. The phases that form during interdiffusion have been examined by x ray diffraction. Analysis of heated test specimens indicates that the Ir-Re reaction produces a solid solution phase of Ir dissolved in the HCP structure of Re.

Stulen, R. H.↗
NASA NTRS2012

Titanium-Water Thermosyphon Gamma Radiation Exposure and Results

Titanium-water thermosyphons are being considered for use in heat rejection systems for fission power systems. Their proximity to the nuclear reactor will result in some gamma irradiation. Noncondensable gas formation from radiation-induced breakdown of water over time may render portions of the thermosyphon condenser inoperable. A series of developmental thermosyphons were operated at nominal operating temperature under accelerated gamma irradiation, with exposures on the same order of magnitude as that expected in 8 years of heat rejection system operation. Temperature data were obtained during exposure at three locations on each thermosyphon: evaporator, condenser, and condenser end cap. Some noncondensable gas was evident; however, thermosyphon performance was not affected because the noncondensable gas was compressed into the fill tube region at the top of the thermosyphon, away from the heat rejecting fin. The trend appeared to be an increasing amount of noncondensable gas formation with increasing gamma irradiation dose. Hydrogen is thought to be the most likely candidate for the noncondensable gas and hydrogen is known to diffuse through grain boundaries. Post-exposure evaluation of one thermosyphon in a vacuum chamber and at temperature revealed that the noncondensable gas diffused out of the thermosyphon over a relatively short period of time. Further research shows a number of experimental and theoretical examples of radiolysis occurring through gamma radiation alone in pure water.

Sanzi, James, L.A↗
NASA NTRS2012

Titanium-Water Thermosyphon Gamma Radiation Effects and Results

Titanium-water thermosyphons are being considered for use in heat rejection systems for fission power systems. Their proximity to the nuclear reactor will result in some exposure to gamma irradiation. Non-condensable gas formation from radiation may breakdown water over time and render a portion of the thermosyphon condenser inoperable. A series of developmental thermosyphons were operated at nominal operating temperature with accelerated gamma irradiation exposures on the same order of magnitude that is expected in eight years of heat rejection system operation. Temperature data were obtained during exposure at three locations on each thermosyphon; evaporator, condenser, and condenser end cap. Some non-condensable gas was evident, however thermosyphon performance was not affected because the non-condensable gas was compressed into the fill tube region at the top of the thermosyphon, away from the heat rejecting fin. The trend appeared to be an increasing amount of non-condensable gas formation with increasing gamma irradiation dose. Hydrogen is thought to be the most likely candidate for the non-condensable gas and hydrogen is known to diffuse through grain boundaries. Post-exposure evaluation of selected thermosyphons at temperature and in a vacuum chamber revealed that the non-condensable gas likely diffused out of the thermosyphons over a relatively short period of time. Further research shows a number of experimental and theoretical examples of radiolysis occurring through gamma radiation alone in pure water.

Sanzi, James L.↗
DOE OSTI2025

Multiscale, mechanistic modeling of cesium transport in silicon carbide for TRISO fuel performance prediction

Understanding cesium (Cs) transport in TRistructural ISOtropic (TRISO) particle fuel is crucial for predicting fission product release in high-temperature reactors. However, current challenges include significant scatter in diffusivity data and unexplained temperature-dependent diffusion regimes in the silicon carbide layer. This study addresses these challenges by developing a multiscale, mechanistic Cs transport model integrating atomistic simulations and phase field modeling. Our model quantifies temperature and grain size effects on Cs diffusivity, attributing experimentally observed regimes to a transition from bulk-dominated diffusivity at high temperatures to grain boundary-dominated diffusivity at lower temperatures. The model, validated against diffusion measurements and advanced gas reactor (AGR)-1 and AGR-2 post-irradiation fission product release data, enhances the predictive capability of the BISON fuel performance code. This study advances our understanding of Cs release from TRISO particles and its dependence on temperature and silicon carbide grain size, with implications for the safety and efficiency of high-temperature nuclear reactors.

BISON↗
DOE OSTI2024

Demystify radiation-enhanced hydrogen isotope diffusion in Fe-Ni-Cr austenitic stainless steels

Understanding and containing hydrogen isotope diffusion is crucial for many nuclear applications. In situ experiments have consistently shown that radiation significantly enhances isotope diffusion in austenitic stainless steels. Despite extensive research, the mechanism behind this phenomenon remains elusive, as most radiation-induced defects (e.g., vacancies, dislocations, and grain boundaries) typically trap hydrogen, thereby slowing diffusion. While grain boundaries may increase in-plane diffusivity and interstitials may enhance diffusion due to material swelling, these effects are relatively minor. Utilizing an Fe-Ni-Cr-H interatomic potential for stainless steels, we conducted extensive molecular dynamics simulations to investigate the origins of radiation-enhanced diffusion. Here, our findings reveal that when a system is resolidified, mimicking defects created by radiation displacements, the resulting structure contains a mixture of phases, boundaries, and dislocation networks. This defective structure significantly increases hydrogen diffusivity, enhancing it by approximately 1.7 times at 900 K. These results suggest that the complex defect structures formed during radiation displacements are the primary drivers of the observed diffusion enhancement, providing valuable insights into the mechanisms underlying radiation-enhanced diffusion in nuclear materials.

36 MATERIALS SCIENCE↗
NASA NTRS1983

Diffusion processes in Al2O3 scales - Void growth, grain growth, and scale growth

The internal microstructure and growth kinetics of Al2O3 scales on Ni-15Cr-13Al (wt percent) are investigated by TEM and analyzed in relation to models of diffusivity. Polished arc-melted specimens were oxidized in 1-atm air at 1100 C for 0.1, 1.0, and 20 hours and ion-thinned for TEM at 100 kV. The frequency distribution of void size and grain size is determined for different oxidation times and scale depths. The kinetics of microvoid growth and of grain and scale growth are plotted and related via simplified models to lattice and grain-boundary oxygen diffusivity, respectively. Good agreement is found between model predictions and data obtained by Oishi and Kingery (1960) on oxygen diffusion in bulk Al2O3. The further implications and limitations of these findings are discssed.

Smialek, J. L.↗
DOE OSTI2026

Effects of oxygen on the tensile properties and fracture behavior of the Ta-10 W refractory alloy at quasi-static and impact strain rates

This study investigates effects of pre-charged oxygen at levels up to ∼1750 ppm on the tensile properties of the refractory alloy Ta-10 W. Tests were conducted at temperatures to 1100 °C and strain rates from quasi-static (∼10−3 s−1) to impact (∼103 s−1). At all temperatures, ductility decreased with increasing oxygen concentration and could be visually identified as a ductile-to-brittle transition (DBT). Additionally, for a given oxygen concentration, ductility was lower at impact strain rate than at quasi-static. The DBT was accompanied by a change from ductile transgranular to more brittle modes, intergranular and transgranular cleavage. The oxygen concentrations associated with the DBT depend on test temperature and strain rate. However, the transitions themselves are not sharp and cannot be defined as occurring at a specific oxygen concentration. At impact strain rates, ductility is relatively insensitive to test temperatures in the range 300–1100 °C with similar embrittling effects of oxygen; however, oxygen has a more severe embrittling effect at room temperature. This behavior is reversed at quasi-static strain rates where the embrittling effect of oxygen is more severe at elevated temperatures than at room temperature. It appears this difference is related to the kinetics of oxygen diffusion to grain boundaries and crack tips. Specimens that were not pre-charged with oxygen but tensile tested in various partial pressures of oxygen underwent dynamic embrittlement during the test, with the degree of embrittlement increasing as the partial pressure of oxygen in the environment increased. Consistent with such a mechanism, ductility increased with increasing strain rate for a given oxygen partial pressure, which would be the reverse if only intrinsic dislocation mobilities were controlling.

Miller, Roger [ORNL] (ORCID:0000000172155851)↗
DOE OSTI2025

A potential emerging issue concerning repair welding of out-of-core PWR components involving tritium exposure and 3 He retention

This paper identifies a potential but previously unrecognized risk of helium-induced embrittlement and cracking during repair welding of out-of-core pressurized water reactor (PWR) components exposed to tritium-contaminated coolant. While previous weldability concerns centered on 4 He accumulation in neutron-irradiated alloys located within the in-core or near-core regions, new measurements show that 3 He generated by tritium decay can accumulate in out-of-core components. Because hydrogen isotopes readily diffuse to grain boundaries and become trapped there, significant 3 He generation at grain boundaries may lead to cracking during weld repairs. Initial data from far-below-core PWR flux thimble tubes confirm the presence of 3 He levels above known cracking thresholds for repair welds. In conclusion, these findings indicate that out-of-core regions should be considered when defining safe weld repair windows in reactors operating for 60–100 years.

Extended PWR lifetimes↗
DOE OSTI2026

Response to the comments on creep-induced elemental redistribution at grain boundaries of 304L stainless steel – An experimental evidence for diffusional creep mechanisms

Sandström and He reanalyzed our 304L creep data (Acta Materialia 121137, 2025) at 10 MPa/750 °C and proposed that: (i) primary dislocation creep dominates based on the ϕ -model; (ii) measured creep rates deviate from published 304H data and Coble creep predictions using literature grain boundary (GB) diffusivity; and (iii) creep-induced elemental redistribution (CIER) could be explained by dislocation emission/absorption (DEAA) and GB creep (GBC), as opposed to diffusional creep. Here, in this response, we propose that: (i) the creep behavior at 15 MPa/750 °C—for which we demonstrated CIER—is only partially captured by the ϕ -model; (ii) direct comparison between 304L and 304H creep data is inappropriate due to carbides in 304H, and Coble creep predictions estimated from CALPHAD-based bulk diffusivity span experimental rates; and (iii) CIER exhibits distinct fingerprint for diffusional creep characterized by anisotropy and stress orientation dependence whereas neither DEAA nor GBC have been mechanistically substantiated.

36 - MATERIALS SCIENCE↗
NASA NTRS1972

Elevated temperature deformation of TD-nickel base alloys

Sensitivity of the elevated temperature deformation of TD-nickel to grain size and shape was examined in both tension and creep. Elevated temperature strength increased with increasing grain diameter and increasing L/D ratio. Measured activation enthalpies in tension and creep were not the same. In tension, the internal stress was not proportional to the shear modulus. Creep activation enthalpies increased with increasing L/D ratio and increasing grain diameter, to high values compared with that of the self diffusion enthalpy. It has been postulated that two concurrent processes contribute to the elevated temperature deformation of polycrystalline TD-nickel: (1) diffusion controlled grain boundary sliding, and (2) dislocation motion.

Petrovic, J. J.↗
NASA NTRS1974

Elevated temperature deformation of thoria dispersed nickel-chromium

The deformation behavior of thoria nickel-chromium (TD-NiCr) was examined over the temperature range 593 C (1100 F) to 1260 C (2300 F) in tension and compression and at 1093 C (2000 F) in creep. Major emphasis was placed on: (1) the effects of the material and test related variables (grain size, temperature, stress and strain rate) on the deformation process; and (2) the evaluation of single crystal TD-NiCr material produced by a directional recrystallization process. Elevated temperature yield strength levels and creep activation enthalpies were found to increase with increasing grain size reaching maximum values for the single crystal TD-NiCr. Stress exponent of the steady state creep rate was also significantly higher for the single crystal TD-NiCr as compared to that determined for the polycrystalline materials. The elevated temperature deformation of TD-NiCr was analyzed in terms of two concurrent, parallel processes: diffusion controlled grain boundary sliding, and dislocation motion.

Kane, R. D.↗
NASA NTRS1976

Creep deformation of TD-nickel chromium

An investigation was conducted of the mechanical behavioral characteristics of thoria-dispersed (TD) NiCr materials at elevated temperatures. The experimental procedure used is discussed along with the significance of the obtained results. Attention is given to basic creep behavior and creep thermal activation parameter measurements. It is found that the overall creep behavior of TD-NiCr can be explained on the basis of the relative contributions of two parallel-concurrent deformation mechanisms, including diffusion controlled grain boundary sliding and dislocation motion.

Kane, R. D.↗