Low cycle fatigue properties of type 316 stainless steel in vacuum

Furuya, Kazuo; Nagata, Norio; Watanabe, Ryoji

doi:10.1016/0022-3115(80)90069-0

Cited by 25 publications

(6 citation statements)

References 7 publications

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“…We now discuss the results of the various tests at 300 K or 333 K, As mentioned in Fig.3, the (Ph- K or 333 K, being in agreement with that Nf of the solution-treated 3 1 6~ stainless steel is not sensitive to the test temperatures [7]. It is noted that the heart of the TP-tests is the abrupt change in the specimen temperatures, i.e., gradual changes in the specimen temperatures hardly modify Nf.…”

Section: N Fsupporting

confidence: 78%

Effects of In-Situ Proton-Irradiation and Thermal-Pulse on the High-Cycle Fatigue Properties of Low Carbon 316 Stainless Steels

et al. 1996

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Abstract. For the modified 3 16L stainless steel in which the dislocation interaction governs the fatigue hardening process, the high-cycle fatigue tests were carried out between 300 and 403 K. The in situ irradiation tests and the in situ thermal-pulse tests indicate that dislocation rearrangements due to thermal-pulse give rise to an increase in the areal density of persistent slip bands (PSBs), resulting in elongation of the fatigue life Nfi and an introduction of obstacles to dislocation motions, irradiation induced lattice defects or the fatigue induced precipitates, brings about a decrease in the areal density of PSBs, giving shortening of Nf

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Section: N Fsupporting

confidence: 78%

Effects of In-Situ Proton-Irradiation and Thermal-Pulse on the High-Cycle Fatigue Properties of Low Carbon 316 Stainless Steels

et al. 1996

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“…Using somewhat smaller specimens (3 mm diameter and 5 mm gauge length) Furuya et al [28] examined 316 steel in LCF from ambient temperature to 800 C in a vacuum. Again, an indirect strain measurement technique was employed whereby the total strain of gauge section was derived by subtracting the total elastic displacement of the specimen shoulders and loading system from the actuator displacement.…”

Section: Environmental Chamber-no Direct Extensometer Accessmentioning

confidence: 99%

Direct and indirect strain measurement during low cycle fatigue of metals at elevated temperature

Skelton¹

2011

Materials at High Temperatures

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Current testing practices in low cycle fatigue (LCF) determinations at elevated temperature recommend the use of side-contacting extensometers or other direct methods for the measurement of strain. In many cases (e.g., when an environmental chamber surrounds the test specimen) this is not possible and resort must be had to indirect methods, which require a correction factor depending on the cyclic stress -strain response of the material. By mounting direct and indirect extensometers on the specimen and grip assembly of the testing machine, direct comparisons were made on the output strain signals during LCF tests on alloys 316, Nimonic 90 and IN718 at temperatures in the range 600 -900 C. Agreement for plastic strain range was in general good, the remote method however requiring an independent value of the elastic modulus for determination of total strain range. Typical uncertainties in these values are discussed. All comparisons were made under continuous cycling conditions and a correction factor for use under creep-fatigue conditions is suggested.

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“…For two heats of materials tested at 593 and 625°C, the loss of fatigue life was recovered at longer hold times. Furuya et al (1980) reported the fatigue data of 316 SS at temperatures of 20-800°C in vacuum. Tests were done under push-pull, strain-controlled mode under continuous cycling with strain rates of 5×10 -3 to 5×10 -5 /sec.…”

Section: Austenitic Stainless Steelsmentioning

confidence: 99%

“…A large number of creep-fatigue tests were carried out on N&T and N&T and aged G91 of five different heats in JAPC-USDOE joint study (ORNL/9Cr/90-2, 91-1, 92-2, Gieseke et al 1993). Test temperatures were in the range of 482 to 593°C.…”

Section: Modified 9cr-1mo Steelmentioning

confidence: 99%

Resolution of qualification issues for existing structural materials.

Natesan¹,

Li²,

Majumdar³

et al. 2012

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This report gives a detailed assessment of several key technical issues that needs resolution for the existing structural materials with emphasis on application in liquid metal reactors (LMRs), in particular, sodium cooled fast reactors. The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for the Advanced Fuel Cycle Initiative (AFCI) Reactor Campaign. The report is the second deliverable in FY09 (M2505050201) under the work package "Advanced Materials Code Qualification".The overall objective of the Advanced Materials Code Qualification project is to evaluate the key technical requirements for the qualification of currently available and future advanced materials for application in sodium reactor systems and the resolution of issues that the U.S. Nuclear Regulatory Commission (NRC) has raised in the past on structural materials in support of the design and licensing of the LMR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. However, the development of sufficient database and qualification of these materials for application in LMRs require considerable amount of time and resources. In the meantime, the currently available materials will be used in the early development of fast reactors.Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants licensing. As the LMR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). Assessment of materials performance issues and high temperature design methodology issues pertinent to the LMR were presented in an earlier report (Natesan et al. 2008). In a subsequent report ), we addressed the needs in high temperature methodologies for design of various high temperature components in sodium cooled fast reactor.The present report addresses several key technical issues for the currently available structural materials such as Type 304 and 316 austenitic stainless steels and ferritic steels such as 2.25Cr-1Mo and modified 9Cr-1Mo. The 60-year design life for the LMR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high temperature structural design methodology. The current Subsection NH is applicable to the design life only up to 34 years. No experimental data contain test durations of 525,000 hours, and it is impractical...

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Low cycle fatigue properties of type 316 stainless steel in vacuum

Cited by 25 publications

References 7 publications

Effects of In-Situ Proton-Irradiation and Thermal-Pulse on the High-Cycle Fatigue Properties of Low Carbon 316 Stainless Steels

Effects of In-Situ Proton-Irradiation and Thermal-Pulse on the High-Cycle Fatigue Properties of Low Carbon 316 Stainless Steels

Direct and indirect strain measurement during low cycle fatigue of metals at elevated temperature

Resolution of qualification issues for existing structural materials.

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