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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

INL/JOU-17-41026-Rev000

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700°C and to a peak dose of 7 dpa (displacements per atom). The specimen’s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Temperature dependence of viscoelastic Poisson’s ratio of cement mortar

Abstract Concrete creep research has focused primarily on uniaxial response. However, biaxially prestressed concrete structures are common, resulting in a multiaxial stress state that can complicate the behavior of a viscoelastic material like concrete. Significant creep strains may be induced in directions transverse to each principle stress due to Poisson’s effect. Past research is unclear regarding the viscoelastic or viscoplastic properties of concrete outside of uniaxial response. It has been reported in separate studies that concrete viscoelastic/viscoplastic Poisson’s ratio (VPR) is an increasing, decreasing and constant function with time, with all reported measurements performed at room temperature. In this paper, the 3D basic creep response of mature cement mortar is examined using a confined compression experiment that allows direct determination of the full stress and infinitesimal strain tensors in a single test, which enables the determination of VPR under a multiaxial stress state. For this purpose, a unique, miniature version of the standardized concrete creep frame is designed that is amenable to placing in climate chambers and temperature ovens. The experimental results indicate that the VPR of sealed, mature cement mortar is nearly constant and equal to the elastic at room temperature, while the VPR gradually increases with time when measured at 60 °C.

Construction & Building Technology↗
DOE OSTI2013

Investigation of a Novel NDE Method for Monitoring Thermomechanical Damage and Microstructure Evolution in Ferritic-Martensitic Steels for Generation IV Nuclear Energy Systems

The main goal of the proposed project is the development of validated nondestructive evaluation (NDE) techniques for in situ monitoring of ferritic-martensitic steels like Grade 91 9Cr-1Mo, which are candidate materials for Generation IV nuclear energy structural components operating at temperatures up to ≈650 °C and for steam-generator tubing for sodium-cooled fast reactors. Full assessment of thermomechanical damage requires a clear separation between thermally activated microstructural evolution and creep damage caused by simultaneous mechanical stress. Creep damage can be classified as "negligible" creep without significant plastic strain and "ordinary" creep of the primary, secondary, and tertiary kind that is accompanied by significant plastic deformation and/or cavity nucleation and growth. Under negligible creep conditions of interest in this project, minimal or no plastic strain occurs, and the accumulation of creep damage does not significantly reduce the fatigue life of a structural component so that low-temperature design rules, such as the ASME Section III, Subsection NB, can be applied with confidence. The proposed research project will utilize a multifaceted approach in which the feasibility of electrical conductivity and thermo-electric monitoring methods is researched and coupled with detailed post-thermal/creep exposure characterization of microstructural changes and damage processes using state-of-the-art electron microscopy techniques, with the aim of establishing the most effective nondestructive materials evaluation technique for particular degradation modes in high-temperature alloys that are candidates for use in the Next Generation Nuclear Plant (NGNP) as well as providing the necessary mechanism-based underpinnings for relating the two. Only techniques suitable for practical application in situ will be considered. As the project evolves and results accumulate, we will also study the use of this technique for monitoring other GEN IV materials. Through the results obtained from this integrated materials behavior and NDE study, new insight will be gained into the best nondestructive creep and microstructure monitoring methods for the particular mechanisms identified in these materials. The proposed project includes collaboration with a national laboratory partner and the results will also serve as a foundation to guide the efforts of scientists in the DOE laboratory, university, and industrial communities concerned with the technological challenges of monitoring creep and microstructural evolution in materials planned to be used in Generation IV Nuclear Energy Systems.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Assessment of the Material Performance for TCR Applications (Progress Report)

The objective of Argonne’s materials research activities for the Transformational Challenge Reactor (TCR) program is to improve the understanding of material properties from additive manufacturing with the focus on understanding the creep and fatigue properties of additively manufactured materials. The material work conducted by Argonne provides support in developing and qualifying advanced materials and manufacturing processes to allow for innovative reactor design and licensing for the TCR. In FY 2020, Oak Ridge National Laboratory (ORNL) provided six ASTM standard sized creep specimens of additively-manufactured 316L stainless steel (AM 316L SS) for evaluation of the effect of post-build heat treatment on the creep property. Two different teat treatments were conducted on the as-built specimens, namely 1050°C/1h and 650°C/1h before creep testing. All specimens were creep tested at 550°C and 275 MPa to understand the effect of heat treatments. We have also received AM 316L SS printed in various forms including rods and sheets, wrought 316 L SS (as a reference material), and AM IN718 for tensile, creep and fatigue testing. Both ASTM standard sized specimens and subsized specimens are being fabricated or tested to contribute to a comprehensive mechanical property database of AM materials and to provide location-specific property data to the TCR digital platform. Mechanical testing at Argonne is conducted to ASME NQA-1 or its lab equivalent for quality assurance. This report presents the results of the creep tests at 550°C and microstructural characterization for the assessment of the effect of the post-build heat treatment on the creep behavior of AM 316L SS. It also discusses ongoing efforts on obtaining location-dependent properties of AM 316L SS.

36 MATERIALS SCIENCE↗
DOE OSTI2013

AGC-1 Specimen Post Irradiation Data Report

Here we report the results of the Post Irradiation Examination (PIE) of the creep, control, and piggyback specimens from the irradiation creep capsule AGC-1. This is the first (prototype) of a series of six capsules planned as part of the Advanced Graphite Creep (AGC) experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphites. The data reported include: specimen dimensions and hence the dimensional change upon irradiation (a comparison of these data for specimen matched pairs will yield the creep strain); mass and volume, hence density; elastic constants (Young’s modulus, shear modulus and Poisson’s ratio) from ultrasonic time of flight (TOF) and velocity measurements; Young’s modulus from the fundamental frequency of vibration; electrical resistivity; and thermal expansion from 100-550°C (hence coefficient of thermal expansion). Over 8500 individual physical measurement have been made and the data are reduced and reported here. The AGC-1 capsule was irradiated in the Advanced Test Reactor (ATR) at INL at approximately 700°C and to a peak dose of 7 dpa (displacements per atom). The specimen’s final dose, temperature and stress conditions have been reported by INL and are also tabulated here. Additionally the analysis of the SiC temperature monitors is reported. These temperature monitors were located along the center of the AGC-1 capsule and provide an independent assessment of the capsules final irradiation temperature. While all the PIE testing is reported (with the exception of the glued-end tensile strength specimens) additional analysis of the AGC data will be required to allow: 1. Improved fits to the existing models of graphite irradiation induced creep strain. 2. With the eventual advent of temperature dependency data from the additional AGC series of capsules, and microstructural data from piggy back specimens in AGC capsules, the development of new and improved models for the development of irradiation induced creep strain in graphite are anticipated. 3. An Investigation into the effects of creep strain on the physical properties of irradiated graphite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

FY24 Integrated Results for High-Temperature Mechanical Testing of LPBF 316H Stainless Steel

This report provides the mechanical test data of laser powder bed fusion (LPBF) 316H stainless steel (SS) collected during Fiscal Year 2024 (FY 24) under the US Department of Energy, Office of Nuclear Energy’s Advanced Materials and Manufacturing Technologies program, along with the current state of microstructural-based understandings of the behaviors. Materials with variabilities in manufacturing site, machine, laser parameters, powder chemistry, porosity, specimen geometry and heat treatment were tested in high-temperature tension, creep, fatigue and creep-fatigue. Electron microscopy and optical microscopy were performed on selected materials before and after the tests to provide microstructural-based understandings to the mechanical behavior. It was discovered that the as-built (AB) and stress-relieved (SR) materials exhibit similar behaviors in tension and creep, as do the solution-annealed (SA) and hot-isostatic pressed (HIP) materials. The AB and SR materials are softer but more ductile than the SA and HIP materials in tension. In creep, the LPBF materials have comparable rupture times but lower rupture strains compared to the wrought materials. The AB and SR materials have low creep rupture strains (<10%) when tested at 720°C and 800°C, while the SA and HIP materials are much more ductile. The LPBF materials exhibit large scatter in fatigue and creep-fatigue lives. The cyclic lives of creep-fatigue tests are lower than those of fatigue tests. In fatigue and creep-fatigue, SR materials outperform SA materials. A common observation from the cyclic tests is that, given the same post-build heat treatment, a lower initial peak stress generally results in a longer cyclic life regardless of hold time. It was also discovered that the batch variation can be more impactful than heat treatment, as demonstrated by the overall better performance of one batch of material than another, regardless of the heat treatment. The results provided insights into how different factors impact the behaviors of LPBF 316H SS. An outlook to FY 25 work scope is provided.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Laser Powder Bed Fusion Manufactured Alloys 625 and 282 and Properties Relevant to Advanced Nuclear Components

Under the Advanced Materials and Manufacturing Technologies program, two Ni-based alloys fabricated by laser powder bed fusion (LPBF) have been evaluated: γ′-strengthened Haynes 282 and solution-strengthened Inconel 625. Large printing defects were observed in the LPBF 282 alloy fabricated using a Renishaw AM250 machine, likely due to particle spattering during printing. Annealing at 1,180°C for 1 h followed by 4 h at 800°C in a high density of 50 nm γ′-strengthening precipitates as well as partial recrystallization resulted in a bimodal grain distribution. Creep testing performed at 750°C revealed lower creep life and ductility for the LPBF 282 compared with wrought 282. X-ray computed tomography combined with optical and scanning electron microscopy microstructural characterization revealed crack formation during creep testing, initiated either from printing defects or from creep cavitation at grain boundaries. Printing defects were the likely reason for the lower creep performance of LPBF 282 and could be suppressed by optimizing the build configuration. Printing 625 on an EOS M290 machine using the recommended EOS parameters resulted in a very low defect density. Superior creep strength at 725°C was observed for the as-printed LPBF 625 along the build direction when compared with wrought 625. No defect larger than ~50 µm could be detected by X-ray computed tomography in the as-printed conditions and after creep testing for 1,000 h at 725°C and 150 MPa. Small, needle-like δ-phase precipitates were observed after creep testing for 100 h at 725°C and 200 MPa. The rapid formation of the γ phase is directly related to the initial Nb and Mo segregation in the LPBF 625 cell walls, and its effect on the alloy’s high-temperature performance will be evaluated. Solution annealing was carried out at 1,150°C for 1 h and led to full recrystallization of the alloy. A decrease in strength and increase in ductility were observed after solution annealing, and the alloy recrystallization resulted in isotropic tensile properties at room temperature, contrary to what was observed for the as-printed LPBF 625.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1981

Elevated temperature mechanical properties and residual tensile properties of two cast superalloys and several nickel-base oxide dispersion strengthened alloys

The elevated temperature tensile, stress-rupture and creep properties and residual tensile properties after creep straining have been determined for two cast superalloys and several wrought Ni-16Cr-4Al-yttria oxide dispersion strengthened (ODS) alloys. The creep behavior of the ODS alloys is similar to that of previously studied ODS nickel alloys. In general, the longitudinal direction is stronger than the long transverse direction, and creep is at least partially due to a diffusional creep mechanism as dispersoid-free zones were observed after creep-rupture testing. The tensile properties of the nickel-base superalloy B-1900 and cobalt-base superalloy MAR-M509 are not degraded by prior elevated temperature creep straining (at least up to 1 pct) between 1144 and 1366 K. On the other hand, the room temperature tensile properties of ODS nickel-base alloys can be reduced by prior creep strains of 0.5 pct or less between 1144 and 1477 K, with the long transverse direction being more susceptible to degradation than the longitudinal direction.

Whittenberger, J. D.↗
NASA NTRS1989

Procedures for characterizing an alloy and predicting cyclic life with the total strain version of Strainrange Partitioning

Procedures are presented for characterizing an alloy and predicting cyclic life for isothermal and thermomechanical fatigue conditions by using the total strain version of strainrange partitioning (TS-SRP). Numerical examples are given. Two independent alloy characteristics are deemed important: failure behavior, as reflected by the inelastic strainrange versus cyclic life relations; and flow behavior, as indicated by the cyclic stress-strain-time response (i.e., the constitutive behavior). Failure behavior is characterized by conducting creep-fatigue tests in the strain regime, wherein the testing times are reasonably short and the inelastic strains are large enough to be determined accurately. At large strainranges, stress-hold, strain-limited tests are preferred because a high rate of creep damage per cycle is inherent in this type of test. At small strainranges, strain-hold cycles are more appropriate. Flow behavior is characterized by conducting tests wherein the specimen is usually cycled far short of failure and the wave shape is appropriate for the duty cycle of interest. In characterizing an alloy pure fatigue, or PP, failure tests are conducted first. Then depending on the needs of the analyst a series of creep-fatigue tests are conducted. As many of the three generic SRP cycles are featured as are required to characterize the influence of creep on fatigue life (i.e., CP, PC, and CC cycles, respectively, for tensile creep only, compressive creep only, and both tensile and compressive creep). Any mean stress effects on life also must be determined and accounted for when determining the SRP inelastic strainrange versus life relations for cycles featuring creep. This is particularly true for small strainranges. The life relations thus are established for a theoretical zero mean stress condition.

Saltsman, James F.↗
NASA NTRS2000

Accelerated Testing of Polymeric Composites Using the Dynamic Mechanical Analyzer

Creep properties of IM7/K3B composite material were obtained using three accelerated test methods at elevated temperatures. Results of flexural creep tests using the dynamic mechanical analyzer (DMA) were compared with results of conventional tensile and compression creep tests. The procedures of the three test methods are described and the results are presented. Despite minor differences in the time shift factor of the creep compliance curves, the DMA results compared favorably with the results from the tensile and compressive creep tests. Some insight is given into establishing correlations between creep compliance in flexure and creep compliance in tension and compression. It is shown that with careful consideration of the limitations of flexure creep, a viable and reliable accelerated test procedure can be developed using the DMA to obtain the viscoelastic properties of composites in extreme environments.

Abdel-Magid, Becky M.↗
DOE OSTI2021

Recommendation for Limiting Conditions for ASME BPVC Section III Division 5 Allowable Stress Criteria

Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) specifies rules for elevated temperature nuclear reactors. Time dependent allowable stresses are an important aspect of the high temperature design models and are related to the time to 1% strain, time to rupture, and time to onset of tertiary creep. The time to onset of tertiary creep criterion has an overly strong negative influence on the allowable stresses for some structural materials, likely due to very limited datasets and misidentification of microstructural changes rather than onset of significant creep damage, as tertiary creep is meant to represent. It is recommended that the onset of tertiary creep be determined through an alloy specific ratio between onset of tertiary creep and creep rupture time, which would give access to the significantly larger creep rupture time datasets.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

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↗
DOE OSTI2016

AGC-2 Irradiation Report

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance are required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Dimensional Stability of the Structural Parts of VVER-1000 FA Made of E-635 Zirconium Alloy

The dependence of dimensional changes occurring in the structural parts made of E-635 zirconium alloy on the metallurgical and operational factors as a result of thermal creep, radiation growth, and radiation creep was studied. Tests close to operating conditions showed high resistance to radiation growth and thermal and radiation creep of the structural parts of VVER-1000 fuel assemblies made of E-635 zirconium alloy. Parts with greater resistance to creep without irradiation also possess high resistance to creep in the presence of irradiation. However, in-reactor creep can be almost 10 times greater than creep with no irradiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

AGC-3 Specimen Post-Irradiation Examination Data Package Report

The Advanced Reactor Technologies Graphite research and development program is conducting an extensive graphite irradiation program to provide data to assist in licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR designs. To support the design and licensing of HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan.[ ] Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be a critical parameter in determining the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. A second, much more significant, internal stress is experienced within the graphite components due to irradiation induced dimensional changes. These induced stresses result from neutron irradiation causing physical changes in the graphite structure, first shrinking it and then expanding it, with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear graphite grades is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Precipitate Phase Stability and Mechanical Properties of Alloy 263 and Variants in Wrought or Cast Form

Alloy 263 is a candidate Ni-based superalloy for use in advanced ultra-supercritical (AUSC) power plants that are targeted to operate above 700 °C. Exposure times in this type of environment are considerable with target creep properties specified at 100,000 h making phase stability an important parameter in the alloy selection process. In this investigation, the microstructure of alloy 263 and two modified compositions in both cast and cast/wrought forms were investigated after exposure at 800 °C for various times ranging from 1,000 h to 10,000 h. Tension and creep properties were assessed for all alloys in both forms. Modification to the Ti and Al concentrations successfully slowed down the γ' to η transformation while doubling the γ' fraction. Thin elongated η needles or platelets formed in the wrought products while thicker, plate-like, η precipitates were found in the castings. The effect of the various microstructures on the creep properties were determined using isothermally aged specimens aged at 800 °C for 3,000 h and 10,000 h prior to creep screening. Strong precipitation of the η phase was found to considerably decrease creep life and increase minimum creep rate. By modifying the Ti and Al content, the creep life and ductility of the new formulation tested on specimens isothermally aged for 3,000 and 10,000 h were near that of the nominal alloy in its standard aged condition.

Detrois, Martin↗
DOE OSTI2016

FY16 Status Report on Development of Integrated EPP and SMT Design Methods

The goal of the Elastic-Perfectly Plastic (EPP) combined integrated creep-fatigue damage evaluation approach is to incorporate a Simplified Model Test (SMT) data based approach for creep-fatigue damage evaluation into the EPP methodology to avoid the separate evaluation of creep and fatigue damage and eliminate the requirement for stress classification in current methods; thus greatly simplifying evaluation of elevated temperature cyclic service. The EPP methodology is based on the idea that creep damage and strain accumulation can be bounded by a properly chosen “pseudo” yield strength used in an elastic-perfectly plastic analysis, thus avoiding the need for stress classification. The original SMT approach is based on the use of elastic analysis. The experimental data, cycles to failure, is correlated using the elastically calculated strain range in the test specimen and the corresponding component strain is also calculated elastically. The advantage of this approach is that it is no longer necessary to use the damage interaction, or D-diagram, because the damage due to the combined effects of creep and fatigue are accounted in the test data by means of a specimen that is designed to replicate or bound the stress and strain redistribution that occurs in actual components when loaded in the creep regime. The reference approach to combining the two methodologies and the corresponding uncertainties and validation plans are presented. Results from recent key feature tests are discussed to illustrate the applicability of the EPP methodology and the behavior of materials at elevated temperature when undergoing stress and strain redistribution due to plasticity and creep.

42 ENGINEERING↗