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At least 919 records · Page 51
All sourcesNASA NTRSDOE OSTI
NASA NTRS1984

Fatigue behavior of continuous fiber silicon-carbide-aluminum composites

Four lay-ups of continuous fiber silicon carbide (SCS2) fiber/aluminum matrix composites were tested to assess fatigue mechanisms including stiffness loss when cycled below their respective fatigue limits. The lay-ups were 0 (sub 8), 0(sub 2)/ + or - 45 (sub 2s), 0/90 (sub 2s),and 0/ + or 45/90 (subs). The data were compared with predictions from the author's previously published shakedown model which predicts fatigue-induced stiffness loss in metal matrix composites. A fifth lay-up, + or - 45 (sub 2s), was tested to compare shakedown and fatigue limits. The particular batch of silicon-carbide fibers tested in this program had a somewhat lower modulus (340 GPa) than expected and displayed poor bonding to the aluminum matrix. Good agreement was obtained between the stiffness loss model and the test data. The fatigue damage below the fatigue limit was primarily in the form of matrix cracking. The fatigue limit corresponded to the laminate shakedown for the + or - 45 (sub 2s) laminate.

Johnson, W. S.↗
DOE OSTI2022

Transient Phase-Driven Cyclic Deformation in Additively Manufactured 15-5 PH Steel

The present work extends the examination of selective laser melting (SLM)-fabricated 15-5 PH steel with the 8%-transient-austenite-phase towards fully-reversed strain-controlled low-cycle fatigue (LCF) test. The cyclic-deformation response and microstructural evolution were investigated via in-situ neutron-diffraction measurements. The transient-austenite-phase rapidly transformed into the martensite phase in the initial cyclic-hardening stage, followed by an almost complete martensitic transformation in the cyclic-softening and steady stage. The compressive stress was much greater than the tensile stress at the same strain amplitude. The enhanced martensitic transformation associated with lower dislocation densities under compression predominantly governed such a striking tension-compression asymmetry in the SLM-built 15-5 PH.

15-5 PH stainless steel↗
NASA NTRS2015

Experimental and Finite Element Modeling of Near-Threshold Fatigue Crack Growth for the K-Decreasing Test Method

The experimental methods to determine near-threshold fatigue crack growth rate data are prescribed in ASTM standard E647. To produce near-threshold data at a constant stress ratio (R), the applied stress-intensity factor (K) is decreased as the crack grows based on a specified K-gradient. Consequently, as the fatigue crack growth rate threshold is approached and the crack tip opening displacement decreases, remote crack wake contact may occur due to the plastically deformed crack wake surfaces and shield the growing crack tip resulting in a reduced crack tip driving force and non-representative crack growth rate data. If such data are used to life a component, the evaluation could yield highly non-conservative predictions. Although this anomalous behavior has been shown to be affected by K-gradient, starting K level, residual stresses, environmental assisted cracking, specimen geometry, and material type, the specifications within the standard to avoid this effect are limited to a maximum fatigue crack growth rate and a suggestion for the K-gradient value. This paper provides parallel experimental and computational simulations for the K-decreasing method for two materials (an aluminum alloy, AA 2024-T3 and a titanium alloy, Ti 6-2-2-2-2) to aid in establishing clear understanding of appropriate testing requirements. These simulations investigate the effect of K-gradient, the maximum value of stress-intensity factor applied, and material type. A material independent term is developed to guide in the selection of appropriate test conditions for most engineering alloys. With the use of such a term, near-threshold fatigue crack growth rate tests can be performed at accelerated rates, near-threshold data can be acquired in days instead of weeks without having to establish testing criteria through trial and error, and these data can be acquired for most engineering materials, even those that are produced in relatively small product forms.

Smith, Stephen W.↗
NASA NTRS2020

Characterization of SiC/SiC Composites Subjected to Sequential Tensile Creep and Dwell Fatigue in Air under Thermal Gradient Conditions

Uncoated SiC/SiC composites, with 2D and 3D fiber architectures, were investigated under steady state thermal gradient conditions with sequential tensile creep and dwell fatigue. Melt infiltration (MI), chemical vapor infiltration (CVI), polymer impregnation and pyrolysis (PIP), and hybrid (CVI and PIP) techniques were used to fabricate ceramic matrix composites (CMCs) with a 2D balanced fiber architecture. CMCs with 3D fiber weaves were also fabricated using MI and hybrid methods. Steady state through the thickness thermal gradients were generated for the tensile specimens using laser heating on the front (hot side) and forced air cooling on the back (cold side). Hot side and cold side temperatures of 2400 and 2200 °F, respectively, were used for testing CMCs fabricated with the MI process, whereas hot side and cold side temperatures of 2700 and 2400 °F, respectively, were used for CMCs fabricated with CVI, PIP, and hybrid processes. Each CMC specimen was initially tested under tensile creep at 69 MPa (10 ksi) for 80 hours, which was then followed by tensile dwell fatigue testing at a maximum stress of 69 MPa (10 ksi) for another 80 hours for a total of 160 hours. This type of sequential tensile creep followed by tensile dwell fatigue testing was performed to ascertain the integrity of uncoated CMCs under steady state thermal gradient conditions. Tensile creep and dwell fatigue test results, generated under steady state thermal gradient conditions on different types of SiC/SiC composites, are compared with corresponding creep behaviors observed under isothermal conditions to determine the influence of both monotonic creep (static fatigue) and dwell fatigue on accumulated strains and deformation behaviors of the CMCs.

Kalluri, Sreeramesh↗
DOE OSTI2021

The status of the Japanese material properties handbook and the challenge to facilitate structural design criteria for DEMO in-vessel components

This work summarizes the current status of the material properties handbook for a structural design using Japanese reduced-activation ferritic/martensitic steel F82H. Specifically, the key structural parameters, e.g. time-independent/dependent design stresses and fatigue design curves, were determined by following the French structural design code RCC-MRx. Moreover, under the Japan–U.S. collaboration, tensile data were newly added to the benchmark heavy irradiation data up to 80 dpa, as critical input information in the intermediate check and review in Japan. Furthermore, the status of structural material data and the near-term and long-term issues were clarified by the evaluation using the attribute guides. In parallel, the structural design approaches, which were newly introduced and extended to cope with the structural design issues under the complex environmental conditions peculiar to the DEMO reactor, were noted with the initial R&D results. Of the many design issues, the multi-axial loading conditions due to the complexity of the DEMO reactor as well as the coolant compatibility and the irradiation effect are mentioned. For example, in the paper, multi-axial fatigue–creep testing and evaluation using the modified universal slope method and brittle/ductile fracture testing and evaluation using the local approach are explained toward DEMO.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Oak Ridge National Laboratory Technical Input for the Nuclear Regulatory Commission Review of the 2017 Edition of ASME Section III, Division 5, ‘High Temperature Reactors’

To assist the Nuclear Regulatory Commission in its decision making on endorsement of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code Section III, Division 5 (2017 Edition) for development of advanced non-light water reactors, the following Division 5 portions were reviewed: Article HBB-2000 Material; Article HCB-2000 Material; Article HGB-2000 Material; Mandatory Appendix HBB-I-14 Tables and Figures; and, Nonmandatory Appendix HBB-U Guidelines for Restricted Material Specifications to Improve Performance in Certain Service Applications. In addition to the 2017 Edition, the same parts of the 2019 Edition have also been reviewed as indicated in various sections of the report. This review was conducted by a collaboration of national laboratory and private sector participants with significant industrial experience, including some heavy lifting and deep diving from Clarus Consulting, LLC., all intended to achieve an objective, independent, and practical perspective. The report provides recommendations, descriptions of the evaluation methods, and the source references for the data used. To build confidence required for endorsement of the Code, this review was conducted as a verification and validation of the above Code contents. The objective of verification is to ensure that the Code is free of error – direct or implied; contains the information needed for its use, including proper coverage of the Code-specified materials for the intended application, and completeness and adequacy of references to other portions of the Code. The objective of validation is to authenticate that the Code tabulations and graphs represent design inputs consistent with what are determined using rules and methods specified by the Code. The authentication process used data that were assembled and/or generated independent of Code development, while the methods of analysis followed Code-specified methods where appropriate. The designated portions for this review cover the five alloys codified for high temperature reactor applications in Division 5, i.e. 316 SS, 304 SS, 800H, 2¼Cr-1Mo, and 9Cr-1Mo-V, regarding their general requirements, permitted specifications and design stress intensity values for pressure-retaining applications, deterioration in service, fatigue acceptance test, permissible weld materials, tensile and yield strength, expected minimum stress-to-rupture values (including for Alloy 718), weld stress rupture factors, permissible materials for bolting use, and restricted specifications in certain service applications. Additionally, stress intensity values for bolting materials including 316 SS, 304 SS and alloy 718 were reviewed. Analysis and discussion are also provided on contents outside of these designated Code portions where it was deemed relevant and necessary to develop a technically sound understanding of issues relating to the designated portions. Due to unavailability of sufficient test data on welds during the review period, the weld stress rupture factors in Tables HBB-I-10.14A to E, which cover a total of ten tables for the five alloys welded with twenty-eight different weld metals (some with similar properties), have been deferred to a future review effort. The review identified mainly two types of issues. The first type includes instances where the Code is found factually incomplete or incorrect, such as obsolete materials specifications listings, missing tabulation of stresses for bolting. Changes to the Code are recommended in these cases. The second type of issue includes instances where the Code tabulations and graphs are found to be less conservative than the review analysis results. In these cases, recommendations are made for further review and consideration where the difference in conservatism exceeds 10%, which is our threshold for questioning technical adequacy, meriting a risk assessment by the Nuclear Regulatory Commission and/or reactor designers. It is noted that this effort has been executed using all available data and established methods of analysis, including methods and criteria specified and used by the Code. As such, the findings that are presented in quantitative detail, in a format for convenient comparison with the Code, and with identification of where further review is recommended, should provide a sound technical basis for decisions about quantifying the implications of the reduced design margins and technical adequacy/inadequacy to form a basis for conditioning specific Code tabulation values on endorsement. Recommendations for specific changes to the Code, however, entail design conservatism considerations beyond the scope of this review effort, and are not made in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗