Carbon and graphite
Mechanical properties, plastic deformation, and tensile creep rate in carbon and graphite
TENSILE CREEP↗
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Effect of diffusional creep on particle morphology of polycrystalline alloys strengthened by second phase particles
Diffusional creep in a polycrystalline alloy containing second-phase particles can disrupt the particle morphology. For alloys which depend on the particle distribution for strength, changes in the particle morphology can affect the mechanical properties. Recent observations of diffusional creep in alloys containing soluble particles (gamma-prime strengthened Ni base alloys) and inert particles have been reexamined in light of the basic mechanisms of diffusional creep, and a generalized model of this effect is proposed. The model indicates that diffusional creep will generally result in particle-free regions in the vicinity of grain boundaries serving as net vacancy sources. The factors which control the changes in second-phase morphology have been identified, and methods of reducing the effects of diffusional creep are suggested.
Wittenberger, J. D.↗
Stiffness and damping characteristics of aluminum in creep
Tensile creep tests conducted at 200 C were performed on annealed commercially pure aluminum specimens in order to measure the dominant elevated temperature dislocation processes. Testing consisted of applying small lateral loads to measure flexural stiffness, and vibrating the specimens laterally in order to measure dynamic modulus and internal damping. It was concluded that (1) the strain hardening increased static stiffness and decreased internal damping during early creep, and (2) the dynamic modulus remained essentially constant at the elastic value during creep. These results imply that primary creep may constitute a mechanism of recovery of dislocatory disorder induced by yielding the material during loading, and that the inelastic modulus utilized as a mathematical concept in several creep buckling theories is not a directly measurable material property.
Berkovits, A.↗
Structural and mechanical effects of interstitial sinks
Creep strength of columbium and tantalum base alloys
Klein, M. J.↗
AGC-2 Specimen Post Irradiation Data Package Report
This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time of flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Creep Resistance of ZrO2 Ceramic Improved by the Addition of a Small Amount of Er2O3
Zirconia (ZrO2) has great technological importance in structural, electrical, and chemical applications. It is the crucial component for state-of-the art thermal barrier coatings and an enabling component as a solid electrolyte for solid-oxide fuel cell systems. Pure ZrO2 is of limited use for industrial applications because of the phase transformations that occur. Upon the addition of stabilizers, cubic (c-ZrO2) and tetragonal (t-ZrO2) forms can be preserved. It is the stabilized and partially stabilized forms of zirconia that function as thermal barrier coatings, solid electrolytes, and oxygen sensors and that have numerous applications in the electrochemical industry. The cubic form of ZrO2 is typically stabilized through Y2O3 additions. However, Y2O3-stabilized zirconia is susceptible to deformation at high temperatures (greater than 900 C) because of the large number of slip systems and the high oxygen diffusion rates, which result in high creep rates at high temperatures. Successful use of ZrO2 at high temperatures requires that new dopant additives be found that will retain or enhance the desirable properties of cubic ZrO2 and yet produce a material with lower creep rates. At the NASA Glenn Research Center, erbium oxide (Er2O3) was identified as a promising dopant for improving the creep resistance of. ZrO2. The selection of Er2O3 was based on the strong interactions of point defects and dislocations. Single crystals of 5 mol% Er2O3- doped ZrO2 rods (4 mm in diameter) and monofilaments (200 to 300 mm in diameter and 30 cm long) were grown using the laser-heated float zone technique, and their creep behavior was measured as a function of temperature. The addition of 5 mol% Er2O3 to single-crystal ZrO2 improved its creep resistance at high temperatures by 2 to 3 orders of magnitude over state-of-the-art Y2O3-doped crystals. Detailed microstructural characterization of ZrO2-Er2O3 single crystals has identified new mechanisms for improving the creep resistance of this class of materials. Adding Er2O3 to ZrO2 results in microstructure of stable and metastable tetragonal precipitates that with thermal treatment evolve to a tweed structure of nanosize tetragonal lamellae. The superior high-temperature creep resistance of Er2O3-doped ZrO2 is attributed to nanoscale precipitation hardening. Doping with Er2O3 will significantly increase the upper-use temperature limit of ZrO2. Potential applications include using Er2O3-doped ZrO2 as a high-temperature fiber for structural applications and adding Er2O3 to reduce the sintering rates of ZrO2 thermal barrier coatings. This work was conducted at Dpto. de F sica de la Materia Condensada, Universidad de Sevilla, Spain, and at NASA Glenn.
Martinez-Fernandez, Julian↗
Damage-mitigating control of a reusable rocket engine for high performance and extended life
The goal of damage mitigating control in reusable rocket engines is to achieve high performance with increased durability of mechanical structures such that functional lives of the critical components are increased. The major benefit is an increase in structural durability with no significant loss of performance. This report investigates the feasibility of damage mitigating control of reusable rocket engines. Phenomenological models of creep and thermo-mechanical fatigue damage have been formulated in the state-variable setting such that these models can be combined with the plant model of a reusable rocket engine, such as the Space Shuttle Main Engine (SSME), for synthesizing an optimal control policy. Specifically, a creep damage model of the main thrust chamber wall is analytically derived based on the theories of sandwich beam and viscoplasticity. This model characterizes progressive bulging-out and incremental thinning of the coolant channel ligament leading to its eventual failure by tensile rupture. The objective is to generate a closed form solution of the wall thin-out phenomenon in real time where the ligament geometry is continuously updated to account for the resulting deformation. The results are in agreement with those obtained from the finite element analyses and experimental observation for both Oxygen Free High Conductivity (OFHC) copper and a copper-zerconium-silver alloy called NARloy-Z. Due to its computational efficiency, this damage model is suitable for on-line applications of life prediction and damage mitigating control, and also permits parametric studies for off-line synthesis of damage mitigating control systems. The results are presented to demonstrate the potential of life extension of reusable rocket engines via damage mitigating control. The control system has also been simulated on a testbed to observe how the damage at different critical points can be traded off without any significant loss of engine performance. The research work reported here is built upon concepts derived from the disciplines of Controls, Thermo-fluids, Structures, and Materials. The concept of damage mitigation, as presented in this report, is not restricted to control of rocket engines. It can be applied to any system where structural durability is an important issue.
Ray, Asok↗
High–Pressure Deformation of Iron–Nickel–Silicon Alloys and Implications for Earth’s Inner Core
Earth’s inner core exhibits strong seismic anisotropy, often attributed to the alignment of hexagonal close-packed iron (hcp-Fe) alloy crystallites with the Earth’s poles. How this alignment developed depends on material properties of the alloy and is important to our understanding of the core’s crystallization history and active geodynamical forcing. Previous studies suggested that hcp-Fe is weak under deep Earth conditions but did not investigate the effects of the lighter elements known to be part of the inner core alloy. Here, we present results from radial X-ray diffraction experiments in a diamond anvil cell that constrain the strength and deformation properties of iron-nickel-silicon (Fe–Ni–Si) alloys up to 60 GPa. We also show the results of laser heating to 1650 K to evaluate the effect of temperature. Observed alloy textures suggest different relative activities of the various hcp deformation mechanisms compared to pure Fe, but these textures could still account for the theorized polar alignment. Fe–Ni–Si alloys are mechanically stronger than Fe and Fe–Ni; extrapolated to inner core conditions, Si-bearing alloys may be more than an order of magnitude stronger. Furthermore, this enhanced strength proportionally reduces the effectivity of dislocation creep as a deformation mechanism, which may suggest that texture developed during crystallization rather than as the result of postsolidification plastic flow.
36 MATERIALS SCIENCE↗
Creep fatigue life prediction for engine hot section materials (isotropic): Fourth year progress review
As gas turbine technology continues to advance, the need for advanced life prediction methods for hot section components is becoming more and more evident. The complex local strain and temperature histories at critical locations must be accurately interpreted to account for the effects of various damage mechanisms (such as fatigue, creep, and oxidation) and their possible interactions. As part of the overall NASA HOST effort, this program is designed to investigate these fundamental damage processes, identify modeling strategies, and develop practical models which can be used to guide the early design and development of new engines and to increase the durability of existing engines.
Nelson, Richard S.↗
Creep behavior of copper at intermediate temperatures. II - Surface microstructural observations. III - A comparison with theory
Three different types of microstructural slip features are noted to occur during the creep of Cu at 0.46-0.72 of absolute melting point. While single slip is associated with higher temperatures and lower stresses, complex wavy slip features are observed at higher temperatures and higher stresses as well as with increasing strains: suggesting the importance of cross-slip mechanisms. At lower temperatures and higher stresses the multiple-slip morphologies observed indicate that cross-slip mechanisms may control the creep of polycrystalline Cu only over a limited stress and temperature range. A phenomenological model is proposed which assumes that (1) cell boundaries within subgrains act as both sources of, and obstacles to, gliding dislocations, and (2) dislocation annihilation occurs at the cell boundaries by climb and cross-slip.
Raj, S. V.↗
Creep/Stress Rupture Behavior and Failure Mechanisms of Full CVI and Full PIP SiC/SiC Composites at Elevated Temperatures in Air
SiC/SiC composites fabricated by melt infiltration are being considered as potential candidate materials for next generation turbine components. However these materials are limited to 2400 F application because of the presence of residual silicon in the SiC matrix. Currently there is an increasing interest in developing and using silicon free SiC/SiC composites for structural aerospace applications above 2400 F. Full PIP or full CVI or CVI + PIP hybrid SiC/SiC composites can be fabricated without excess silicon, but the upper temperature stress capabilities of these materials are not fully known. In this study, the on-axis creep and rupture properties of the state-of-the-art full CVI and full PIP SiC/SiC composites with Sylramic-iBN fibers were measured at temperatures to 2700 F in air and their failure modes examined. In this presentation creep rupture properties, failure mechanisms and upper temperature capabilities of these two systems will be discussed and compared with the literature data.
Creep Durability↗
Accelerated Creep Test Qualification of Creep-Resistance Using the Wilshire–Cano–Stewart Constitutive Model and Stepped Isostress Method
In this paper, a qualification of accelerated creep-resistance of Inconel 718 is assessed using the novel Wilshire–Cano–Stewart (WCS) model and the stepped isostress method (SSM) and predictions are made to conventional creep data. Conventional creep testing is a long-term continuous process; in fact, the ASME B&PV III requires that 10,000+ h of experiments must be conducted to each heat for materials employed in boilers and/or pressure vessel components. This process is costly and not feasible for rapid development of new materials. As an alternative, accelerated creep testing techniques have been developed to reduce the time needed to characterize the creep resistance of materials. Most techniques are based upon the time-temperature-stress superposition principle that predicts minimum-creep-strain-rate (MCSR) and stress-rupture behaviors but lack the ability to predict creep deformation and consider deformation mechanisms that occur for experiments of longer duration. The SSM has been developed, which enables the prediction of creep deformation response as well as reduce the time needed for qualification of materials. The SSM approach has been successful for polymer, polymeric composites, and recently has been introduced for metals. In this study, the WCS constitutive model, calibrated to SSM test data, qualifies the creep resistance of Inconel 718 at 750 °C and predictions are compared to conventional creep testing data. The WCS model has proven to make long-term predictions for stress-rupture, MCSR, creep deformation, and damage in metallic materials. The SSM varies stress levels after time interval adding damage to the material, which can be tracked by the WCS model. The SSM data is calibrated into the model and the WCS model generates realistic predictions of stress-rupture, MSCR, damage, and creep deformation. The calibrated material constants are used to generate predictions of stress-rupture and are postaudit validated using the National Institute of Material Science database. Similarly, the MCSR predictions are compared from previous studies. Finally, the creep deformation predictions are compared with real data and is determined that the results are well in between the expected boundaries. Material characterization and mechanical properties can be determined at a faster rate and with a more cost-effective method. This is beneficial for multiple applications such as in additive manufacturing, composites, spacecraft, and industrial gas turbines.
42 ENGINEERING↗
The Creep of Single Crystals of Aluminum
The creep of single crystals of high-purity aluminum was investigated in the range of temperatures from room temperature to 400 F and at resolved-shear-stress levels of 200, 300, and 400 psi. The tests were designed in an attempt to produce data regarding the relation between the rate of strain and the mechanism of deformation. The creep data are analyzed in terms of shear strain rate and the results are discussed with regard to existing creep theories. Stress-strain curves were determined for the crystals in tinsel and constant-load-rate tests in the same temperature range to supplement the study of plastic deformation by creep with information regarding the part played by crystal orientation, differences in strain markings, and other variables in plastic deformation.
ALUMINUM↗
FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients
Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FAST-1.2.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients
Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation (including creep), fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
ACCELERATED CREEP TEST (ACT) QUALIFICATION OF CREEP RESISTANCE USING THE WCS CONSTITUTIVE MODEL AND STEPPED ISOSTRESS METHOD (SSM)
In this study, a qualification of accelerated creep-resistance of Inconel 718 is assessed using the novel Wilshire-Cano-Stewart (WCS) model and the stepped isostress method (SSM) and predictions are made to conventional creep data. Conventional creep testing (CCT) is a long-term continuous process, in fact, the ASME B&PV III requires that 10,000+ hours of experiments must be conducted to each heat for materials employed in boilers and/or pressure vessel components. This process is costly and not feasible for rapid development of new materials. As an alternative, accelerated creep testing techniques have been developed to reduce the time needed to characterize the creep resistance of materials. Most techniques are based upon the time-temperature-stress superposition principle (TTSSP) that predicts minimum-creep-strain-rate (MCSR) and stress-rupture behaviors but lack the ability to predict creep deformation and consider deformation mechanisms that occur for experiments of longer duration. The stepped isostress method (SSM) has been developed which enables the prediction of creep deformation response as well as reduce the time needed for qualification of materials. The SSM approach has been successful for polymer, polymeric composites, and recently has been introduced for metals. In this study, the WCS constitutive model, calibrated to SSM test data, qualifies the creep resistance of Inconel 718 at 750°C and predictions are compared to CCT data. The WCS model has proven to make long-term predictions for stress-rupture, minimum-creep-strain-rate (MCSR), creep deformation, and damage in metallic materials. The SSM varies stress levels after time interval adding damage to the material, which can be tracked by the WCS model. The SSM data is calibrated into the model and the WCS model generates realistic predictions of stress-rupture, MSCR, damage, and creep deformation. The calibrated material constants are used to generate predictions of stress-rupture and are post-audit validated using the National Institute of Material Science (NIMS) database. Similarly, the MCSR predictions are compared from previous studies. Finally the creep deformation predictions are compared with real data and is determined that the results are well in between the expected boundaries. Material characterization and mechanical properties can be determined at a faster rate and with a more cost-effective method. This is beneficial for multiple applications such as in additive manufacturing, composites, spacecraft, and Industrial Gas Turbines (IGT).
36 MATERIALS SCIENCE↗
FAST-1.0: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients: Developed under NQA-1-2017
Fuel Analysis under Steady-state and Transients (FAST) is the US Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation, including creep, fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure and void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heattransfer correlations, for normal operation through postulated accidents for today’s US-based light water reactor (LWR) fuel designs. FAST-1.0 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.0, which is the first official version of this code. This document is one of a series of documents on FAST; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)
Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗